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Sample records for accident codes applications

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

    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. Codes for NPP severe accident simulation: development, validation and applications

    The software tools that describe various safety aspects of NPP with VVER reactor have been developed at the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN). Functionally, the codes can be divided into two groups: the calculation codes that describe separate elements of NPP equipment and/or a group of processes and integrated software systems that allow solving the tasks of the NPP safety assessment in coupled formulation. In particular, IBRAE RAN in cooperation with the nuclear industry organizations has developed the integrated software package SOCRAT designed to analyze the behavior of NPP with VVER at various stages of beyond-design-basis accidents, including the stages of reactor core degradation and long-term melt retention in a core catcher. The general information about development, validation and applications of SOCRAT code is presented and discussed in the paper. (author)

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

    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.

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

    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

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

    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

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

    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.

  7. Application of MELCOR code to the MCCI analysis in Severe Accident Sequences

    This paper provides some of the technical aspects that can be applied to an analysis of the MCCI phenomena in a severe accident scenario using the current MELCOR version. An application methodology of the MELCOR current version to the analysis of MCCI, the phenomena of which are very uncertain and lack specific knowledge during severe accidents, was introduced. Assumptions based on the experimental results are used instead of the phenomenological detail modeling because of the modeling limitations. In the technical aspects of MCCI, code modification itself is not a big deal, because the code modification is needed for just the user flexibility. The concern will be whether the assumptions made for this analysis are acceptable or not. This paper illustrates the application of a severe accident analysis code, MELCOR, to the analysis of molten corium-concrete interaction (MCCI) phenomena in cases of severe accidents in nuclear power plants. In postulated degraded core accidents, followed by the failure of certain engineered safety features of the reactor system, the reactor core may eventually melt owing to the generation of decay heat. If the safety features of the reactor system fail to arrest the accident within the reactor vessel, the corium (molten core debris) will fall into the reactor cavity and attack the concrete walls and floor. Basemat melt-through refers to the process of concrete decomposition and destruction associated with a corium melt interacting with the reactor cavity basemat. The potential hazard of MCCI is the integrity of the containment building owing to the possibility of a basemat melt-through, containment overpressurization by non-condensible gases, or the oxidation of combustible gases. In the meantime, the MCCI still has large uncertainties in several phenomena such as melt spreading area, debris particulation, and heat transfer between the debris and cooling water. In particular, in the case where the water pool exists in the reactor

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

    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.

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

    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

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

    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

  11. Accident consequence assessment code development

    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)

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

    This paper 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 in 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)

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

    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)

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

    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

  15. A review of the Melcor Accident Consequence Code System (MACCS): Capabilities and applications

    MACCS was developed at Sandia National Laboratories (SNL) under U.S. Nuclear Regulatory Commission (NRC) sponsorship to estimate the offsite consequences of potential severe accidents at nuclear power plants (NPPs). MACCS was publicly released in 1990. MACCS was developed to support the NRC's probabilistic safety assessment (PSA) efforts. PSA techniques can provide a measure of the risk of reactor operation. PSAs are generally divided into three levels. Level one efforts identify potential plant damage states that lead to core damage and the associated probabilities, level two models damage progression and containment strength for establishing fission-product release categories, and level three efforts evaluate potential off-site consequences of radiological releases and the probabilities associated with the consequences. MACCS was designed as a tool for level three PSA analysis. MACCS performs probabilistic health and economic consequence assessments of hypothetical accidental releases of radioactive material from NPPs. MACCS includes models for atmospheric dispersion and transport, wet and dry deposition, the probabilistic treatment of meteorology, environmental transfer, countermeasure strategies, dosimetry, health effects, and economic impacts. The computer systems MACCS is designed to run on are the 386/486 PC, VAX/VMS, E3M RISC S/6000, Sun SPARC, and Cray UNICOS. This paper provides an overview of MACCS, reviews some of the applications of MACCS, international collaborations which have involved MACCS, current developmental efforts, and future directions

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

    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

  17. Development of criticality accident analysis code AGNES

    A one-point kinetics code, AGNES2, has been developed for the evaluation of the criticality accident of nuclear solution fuel system. The code has been evaluated through the simulation of TRACY experiments and used for the study of the condition of the JCO criticality accident. A code, AGNES-P, for the criticality accident of nuclear powder system has been developed based on AGNES2. It is expected that these codes be useful for the evaluation of criticality safety for fuel reprocessing and fabrication plants. (author)

  18. MELCOR Accident Consequence Code System (MACCS)

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

  19. MELCOR Accident Consequence Code System (MACCS)

    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

  20. MELCOR Accident Consequence Code System (MACCS)

    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.

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

    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

  2. MELCOR Accident Consequence Code System (MACCS)

    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

  3. MELCOR Accident Consequence Code System (MACCS)

    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.

  4. Application of the WECHSL code to PWR and BWR specific accident scenarios

    The WECHSL Mod3 version is used to perform an accident analysis for a 1300 MW PWR and a BWR. The analysis starts after the melt has penetrated the reactor pressure vessel and is contained in the dry reactor cavity. The initial melt temperature is estimated to be 2673 K. In the initial phase of the melt/concrete interaction, the dominant energy source in the melt is the energy released in the zirconium oxidation reactions with the concrete decomposition products. Hence the concrete composition will determine the Zr-oxidation and the gas release rates as well as the composition of the released gases. Recent experiments and analyses have shown that the solidus temperature of the oxidic melt decreases much more rapidly with addition of concrete oxide than modelled previously. The solidus temperature of the oxide phase drops rapidly as concrete oxides are incorporated into the melt, approaching the concrete solidus at only about 10 to 20 weight percent of concrete oxides. The calculations are performed using the old estimate and the new solidus temperatures for both reactor types in order to study the influence of that oxide solidus temperature. The condensed Zr/SiO2 chemistry is only relevant for the PWR because of the high content of SiO2 in the siliceous concrete basemat. Compared to former analyses for the PWR the much faster zirconium oxidation leads to a higher temperature of about 100 K in the early phase of melt/concrete interaction and therefore the crust formation process starts later than in the former analyses leading to a longer duration of high gas release rates dominated by H2 because of more effective heat transfer to the concrete in this period of time. The concrete basemat of the BWR consists of pure limestone with a decomposition temperature which is higher than the solidus temperature of the metallic melt. This high concrete decomposition temperature prevents a crust formation at the metal-concrete boundary. Hence a very efficient heat transfer leads

  5. DOZIM - evaluation dose code for nuclear accident

    During a nuclear accident an environmentally significant fission products release can happen. In that case it is not possible to determine precisely the air fission products concentration and, consequently, the estimated doses will be affected by certain errors. The stringent requirement to cope with a nuclear accident, even minor, imposes creation of a computation method for emergency dosimetric evaluations needed to compare the measurement data to certain reference levels, previously established. These comparisons will allow a qualified option regarding the necessary actions to diminish the accident effects. DOZIM code estimates the soil contamination and the irradiation doses produced either by radioactive plume or by soil contamination. Irradiations either on whole body or on certain organs, as well as internal contamination doses produced by isotope inhalation during radioactive plume crossing are taken into account. The calculus does not consider neither the internal contamination produced by contaminated food consumption, or that produced by radioactive deposits resuspension. The code is recommended for dose computation on the wind direction, at distances from 102 to 2 x 104 m. The DOZIM code was utilized for three different cases: - In air TRIGA-SSR fuel bundle destruction with different input data for fission products fractions released into the environment; - Chernobyl-like accident doses estimation; - Intervention areas determination for a hypothetical severe accident at Cernavoda Nuclear Power Plant. For the first case input data and results (for a 60 m emission height without iodine retention on active coal filters) are presented. To summarize, the DOZIM code conception allows the dose estimation for any nuclear accident. Fission products inventory, released fractions, emission conditions, atmospherical and geographical parameters are the input data. Dosimetric factors are included in the program. The program is in FORTRAN IV language and was run on a

  6. Base irradiation simulation and its effect on fuel behavior prediction by TRANSURANUS code: Application to reactivity initiated accident condition

    Highlights: • Selection of parameters for analysis. • Base irradiation simulation of rods fabricated by ENUSA. • Boundary condition implementation using restart options. • RIA simulation of CABRI test CIP3-1. • Sensitivity analysis performance. - Abstract: The purpose of the present paper is to investigate the impact of the base irradiation simulation for predicting fuel behavior under Reactivity Initiated Accident (RIA) conditions. A RIA is a scenario challenging the fuel integrity and consequently, devoted experimental campaigns and related code simulations have been extensively performed. In all experiments in which irradiated fuel is tested, the experiment is preceded by in reactor period, i.e. the base irradiation. In the present paper the considered RIA experiment is CIP3-1 performed in CABRI reactor (part of the OECD/NEA WGFS benchmark); a discussion about the relevance of the base irradiation simulation is presented. Such a work is conducted by sensitivities calculation in which a single parameter, among a preselected set, is changed. The range of variation of such parameters is either supplied within the selected RIA test specification or is taken from typical values available in the open literature. All mentioned calculations have been performed developing a specific model in TRANSURANUS code

  7. Base irradiation simulation and its effect on fuel behavior prediction by TRANSURANUS code: Application to reactivity initiated accident condition

    Lisovyy, Oleksandr, E-mail: o.lisovyy@dimnp.unipi.it [GRNSPG-UNIPI, Via Livornese 1291, Pisa 56122 (Italy); Cherubini, Marco, E-mail: m.cherubini@ing.unipi.it [NINE, Via Livornese 1291, Pisa 56122 (Italy); Lazzerini, Davide, E-mail: d.lazzerini@ing.unipi.it [GRNSPG-UNIPI, Via Livornese 1291, Pisa 56122 (Italy); D’Auria, Francesco, E-mail: f.dauria@ing.unipi.it [GRNSPG-UNIPI, Via Livornese 1291, Pisa 56122 (Italy)

    2015-03-15

    Highlights: • Selection of parameters for analysis. • Base irradiation simulation of rods fabricated by ENUSA. • Boundary condition implementation using restart options. • RIA simulation of CABRI test CIP3-1. • Sensitivity analysis performance. - Abstract: The purpose of the present paper is to investigate the impact of the base irradiation simulation for predicting fuel behavior under Reactivity Initiated Accident (RIA) conditions. A RIA is a scenario challenging the fuel integrity and consequently, devoted experimental campaigns and related code simulations have been extensively performed. In all experiments in which irradiated fuel is tested, the experiment is preceded by in reactor period, i.e. the base irradiation. In the present paper the considered RIA experiment is CIP3-1 performed in CABRI reactor (part of the OECD/NEA WGFS benchmark); a discussion about the relevance of the base irradiation simulation is presented. Such a work is conducted by sensitivities calculation in which a single parameter, among a preselected set, is changed. The range of variation of such parameters is either supplied within the selected RIA test specification or is taken from typical values available in the open literature. All mentioned calculations have been performed developing a specific model in TRANSURANUS code.

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

    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

  9. Code strategy for simulating Severe Accident Scenario

    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)

  10. Fire-accident analysis code (FIRAC) verification

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

  11. Fire-accident analysis code (FIRAC) verification

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

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

    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. ...

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

    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.)

  14. Adjoint-based sensitivity analysis for reactor accident codes

    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

  15. Cost per severe accident as an index for severe accident consequence assessment and its applications

    The Fukushima Accident emphasizes the need to integrate the assessments of health effects, economic impacts, social impacts and environmental impacts, in order to perform a comprehensive consequence assessment of severe accidents in nuclear power plants. “Cost per severe accident” is introduced as an index for that purpose. The calculation methodology, including the consequence analysis using level 3 probabilistic risk assessment code OSCAAR and the calculation method of the cost per severe accident, is proposed. This methodology was applied to a virtual 1,100 MWe boiling water reactor. The breakdown of the cost per severe accident was provided. The radiation effect cost, the relocation cost and the decontamination cost were the three largest components. Sensitivity analyses were carried out, and parameters sensitive to cost per severe accident were specified. The cost per severe accident was compared with the amount of source terms, to demonstrate the performance of the cost per severe accident as an index to evaluate severe accident consequences. The ways to use the cost per severe accident for optimization of radiation protection countermeasures and for estimation of the effects of accident management strategies are discussed as its applications. - Highlights: • Cost per severe accident is used for severe accident consequence assessment. • Assessments of health, economic, social and environmental impacts are included. • Radiation effect, relocation and decontamination costs are important cost components. • Cost per severe accident can be used to optimize radiation protection measures. • Effects of accident management can be estimated using the cost per severe accident

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

    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.

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

    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.

  18. A computer code for analysis of severe accidents in LWRs

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  19. A computer code for analysis of severe accidents in LWRs

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  20. 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

    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)

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

    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.

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

    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

  3. Application of PCTRAN-3/U to studying accident management during PWR severe accident

    In order to improve the safety of nuclear power plant, operator action should be taken into account during a severe accident. While it takes a long time to simulate the plant transient behavior under a severe accident in comparison with the design based accident, a transient simulator should have both high speed calculation capability and interactive functions to model the operating procedures. PCTRAN has been developing to be a simple simulator by using a personal computer to simulate plant behavior under an accident condition. While currently available means usually take relatively long time to simulate plant behavior, using a current high-powered personal computer (PC), PCTRAN-3/U code is designed to operate at a speed significantly faster than real-time. The author describes some results of PCTRAN application in studying the efficiency of accident management for a pressurized water reactor (PWR) during an severe accident

  4. An overview of selected severe accident research and applications

    Severe accident research is being conducted world wide by industry organizations, utilities, and regulatory agencies. As this research is disseminated, it is being applied by utilities when they perform their Individual Plant Examinations (IPEs) and consider the preparation of Accident Management programs. The research is associated with phenomenological assessments of containment challenges and associated uncertainties, severe accident codes and analysis tools, systematic evaluation processes, and accident management planning. The continued advancement of this research and its applications will significantly contribute to the enhanced safety and operation of nuclear power plants. (author)

  5. Overview of SAMPSON code development for LWR severe accident analysis

    The Nuclear Power Engineering Corporation (NUPEC) has developed a severe accident analysis code 'SAMPSON'. SAMPSON's distinguishing features include inter-connected hierarchical modules and mechanistic models covering a wide spectrum of scenarios ranging from normal operation to hypothetical severe accident events. Each module included in the SAMPSON also runs independently for analysis of specific phenomena assigned. The OECD International Standard Problems (ISP-45 and 46) were solved by the SAMPSON for code verifications. The analysis results showed fairly good agreement with the test results. Then, severe accident phenomena in typical PWR and BWR plants were analyzed. The PWR analysis result showed 56 hours as the containment vessel failure timing, which was 9 hours later than one calculated by MELCOR code. The BWR analysis result showed no containment vessel failure during whole accident events, whereas the MELCOR result showed 10.8 hours. These differences were mainly due to consideration of heat release from the containment vessel wall to atmosphere in the SAMPSON code. Another PWR analysis with water injection as an accident management was performed. The analysis result showed that earlier water injection before the time when the fuel surface temperature reached 1,750 K was effective to prevent further core melt. Since fuel surface and fluid temperatures had spatial distribution, a careful consideration shall be required to determine the suitable location for temperature measurement as an index for the pump restart for water injection. The SAMPSON code was applied to the accident analysis of the Hamaoka-1 BWR plant, where the pipe ruptured due to hydrogen detonation. The SAMPSON had initially been developed to run on a parallel computer. Considering remarkable progress of computer hardware performance, as another version of the SAMPSON code, it has recently been modified so as to run on a single processor. The improvements of physical models, numerical

  6. Severe accident analysis code Sampson for impact project

    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)

  7. Severe accident analysis code Sampson for impact project

    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. The development of a severe accident analysis code

    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 in an effect to improve existing models and develop analytical tools for the assessment of severe accidents. For hydrogen control, the analysis of hydrogen concentration in the containment and visualization for the concentration in the cell were performed. The computer code to predict combustion flame characteristic was also developed. the analytical model for the expansion phase of vapor explosion was developed and verified with the experimental results. The corium release fraction model from the cavity with the capture volume was developed and applied to the power plants. Pre-test calculation was performed for molten corium concrete interaction study and the crust formation process, heat transfer characteristics of the crust, and the sensitivity study using MELCOR code was carried out. A stress analysis code using finite element method for the reactor vessel lower head failure analysis was developed and the effect by gap formation between molten corium and vessel was analyzed. Through the international program of PHEBUS-FP and participation in the software development, the study on fission products release and transportation in the software development, the study on fission products release and transportation and aerosol deposition were performed. The system for severe accident analysis codes, CONTAIN and MELCOR codes etc., under the cooperation with USNRC were also established by installing in workstation and applying to experimental results and real plants. (author). 116 refs., 31 tabs., 59 figs

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

    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.

  10. Code assessment in context of severe accident phenomenology

    Bratfisch, C.; Agethen, K.; Braehler, T.; Risken, T.; Koppers, V.; Gremme, F.; Hoffmann, M.; Koch, M.K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2014-05-15

    The following paper gives an outline of current research activities in the field of reactor simulation and safety at Ruhr-Universitaet Bochum. Results related to phenomena of core degradation, hydrogen combustion and molten corium concrete interaction will be presented. These deal with the simulation of relevant experiments in order to validate the severe accident codes ASTEC, ATHLET-CD and COCOSYS. Exemplarily, simulation results of the tests QUENCH-16, BMC Ix9 and OECD CCI-2/-3 are discussed. The importance of these phenomena is illustrated by the Three Mile Island and Fukushima Daiichi accidents. (orig.)

  11. Coupled code calculation of rod withdrawal at power accident

    Grgić, Davor, E-mail: davor.grgic@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Benčik, Vesna, E-mail: vesna.bencik@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Šadek, Siniša, E-mail: sinisa.sadek@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia)

    2013-08-15

    Highlights: ► Sensitivity calculations (withdrawal speed, initial power, secondary side influence) were performed for the rod withdrawal at power accident in PWR. ► Best estimate coupled RELAP5-PARCS code calculation was done, using COBRA code to model the core thermal-hydraulics. ► Specific modelling features included reactor vessel split model, explicit model of the RTD bypass and the overtemperature ΔT setpoint function. ► Average whole core values and the local hot spots were predicted. ► Local fuel centerline temperature and local DNBR were calculated using a COBRA-like model. ► Influence of the burnup on the fuel centerline temperature was studied. -- Abstract: The rod withdrawal at power (RWAP) accident is analyzed for NPP Krško as part of activity related to possible resistance temperature detectors (RTDs) bypass removal. The RWAP accident can be departure from nucleate boiling (DNB) or overpower limiting accident depending on initial power level and rate and amount of reactivity addition. In this paper we have analyzed the response of the plant in current configuration to RWAP for different withdrawal speeds and different initial power levels. By demonstrating adequacy of current protection system we can, in the next step, quantify the influence of change in narrow range coolant temperature measurement to available safety margins. The overtemperature ΔT setpoint and its relation to local DNBR values are in center of attention. The coupled RELAP5–PARCS code was used as the calculation tool with the provision to extend the calculation to local pin-by-pin COBRA subchannel calculation for selected state points derived from main coupled code results. In the first part of the calculation methodology, point kinetics calculation is performed using standalone RELAP5 to reproduce USAR results, and in the second part, more demanding coupled code calculation is introduced.

  12. Coupled code calculation of rod withdrawal at power accident

    Highlights: ► Sensitivity calculations (withdrawal speed, initial power, secondary side influence) were performed for the rod withdrawal at power accident in PWR. ► Best estimate coupled RELAP5-PARCS code calculation was done, using COBRA code to model the core thermal-hydraulics. ► Specific modelling features included reactor vessel split model, explicit model of the RTD bypass and the overtemperature ΔT setpoint function. ► Average whole core values and the local hot spots were predicted. ► Local fuel centerline temperature and local DNBR were calculated using a COBRA-like model. ► Influence of the burnup on the fuel centerline temperature was studied. -- Abstract: The rod withdrawal at power (RWAP) accident is analyzed for NPP Krško as part of activity related to possible resistance temperature detectors (RTDs) bypass removal. The RWAP accident can be departure from nucleate boiling (DNB) or overpower limiting accident depending on initial power level and rate and amount of reactivity addition. In this paper we have analyzed the response of the plant in current configuration to RWAP for different withdrawal speeds and different initial power levels. By demonstrating adequacy of current protection system we can, in the next step, quantify the influence of change in narrow range coolant temperature measurement to available safety margins. The overtemperature ΔT setpoint and its relation to local DNBR values are in center of attention. The coupled RELAP5–PARCS code was used as the calculation tool with the provision to extend the calculation to local pin-by-pin COBRA subchannel calculation for selected state points derived from main coupled code results. In the first part of the calculation methodology, point kinetics calculation is performed using standalone RELAP5 to reproduce USAR results, and in the second part, more demanding coupled code calculation is introduced

  13. Test Data for USEPR Severe Accident Code Validation

    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.

  14. 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

    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.

  15. Severe accident analysis code SAMPSON improvement for IMPACT project

    SAMPSON is the integral code for severe accident analysis in detail with modular structure, developed in the IMPACT project. Each module can run independently and communications with multiple analysis modules supervised by the analysis control module makes an integral analysis possible. At the end of Phase 1 (1994-1997), demonstration simulation by combinations of up to 11 analysis modules had been performed and physical models in the code had been verified by separate-effect tests and validated by integral tests. Multi-dimensional mechanistic models and theoretical-based conservation equations have been applied, during Phase 2 (1998 - 2000). New models for Accident Management evaluation have been also developed. Verification and validation have been performed by analysing separate-effect tests and integral tests, while actual plant analyses are also being in progress. (author)

  16. Health effects estimation code development for accident consequence analysis

    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)

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

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

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

    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

  19. Evaluation of nuclear accidents consequences. Risk assessment methodologies, current status and applications

    General description of the structure and process of the probabilistic methods of assessment the external consequences in the event of nuclear accidents is presented. attention is paid in the interface with Probabilistic Safety Analysis level 3 results (source term evaluation) Also are described key issues in accident consequence evaluation as: effects evaluated (early and late health effects and economic effects due to countermeasures), presentation of accident consequences results, computer codes. Briefly are presented some relevant areas for the applications of Accident Consequence Evaluation

  20. Development and application of calculational theoretical methods for analysis of the RBMK reactor severe accidents

    One studied high-improbable reactor emergencies that may result in a high consequence accident. To control these accidents and to mitigate their consequences one should study and analyze similar emergencies via detailed computer simulation. Application of foreign and Russian codes for RBMK type reactor should be associated with their supplementary verification. In that context one elaborated the table list of processes for supplementary verification of thermohydraulic models of codes designed to analyze severe accidents

  1. 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

    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

  2. Application of FFTBM to severe accidents

    In Europe an initiative for the reduction of uncertainties in severe accident safety issues was initiated. Generally, the error made in predicting plant behaviour is called uncertainty, while the discrepancies between measured and calculated trends related to experimental facilities are called the accuracy of the prediction. The purpose of the work is to assess the accuracy of the calculations of the severe accident International Standard Problem ISP-46 (Phebus FPT1), performed with two versions of MELCOR 1.8.5 for validation purposes. For the quantitative assessment of calculations the improved fast Fourier transform based method (FFTBM) was used with the capability to calculate time dependent code accuracy. In addition, a new measure for the indication of the time shift between the experimental and the calculated signal was proposed. The quantitative results obtained with FFTBM confirm the qualitative conclusions made during the Jozef Stefan Institute participation in ISP-46. In general good agreement of thermal-hydraulic variables and satisfactory agreement of total releases for most radionuclide classes was obtained. The quantitative FFTBM results showed that for the Phebus FPT1 severe accident experiment the accuracy of thermal-hydraulic variables calculated with the MELCOR severe accident code is close to the accuracy of thermal-hydraulic variables for design basis accident experiments calculated with best-estimate system codes. (author)

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

    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

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

    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

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

    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

  6. Algebraic geometric codes with applications

    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.

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

    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

  8. Review of models applicable to accident aerosols

    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

  9. Coremelt-2D Code for Analysis of Severe Accidents in a Sodium Fast Reactor

    In the paper there is a description of COREMELT-2D code designed for carrying out coupled two-dimensional analysis of neutronic and thermohydraulic transients, which may occur in the core of sodium cooled fast reactor (SFR), including severe accidents resulting in damage of SFR core and relocation of its components with the change of their aggregative state, namely: boiling and condensation of coolant, damage and melting of fuel element claddings and fuel, relocation of molten core components, thermal interaction of fuel and coolant and freezing of steel and fuel. So, COREMELT-2D code is capable of analyzing all stages of ULOF accident up to expansion phase characterized by the intensive interaction of molten fuel and sodium. Modular structure of COREMELT-2D code consisting of thermohydraulic module COREMELT and neutronic module RADAR is presented. Preservation equations are solved in COREMELT module in two-dimensional cylindrical R-Z geometry in porous body approximation. RADAR module is used for solving multi-group neutron diffusion equation in R-Z and X-Y geometry. Application of the code for solving dynamics tasks with rather rapid changes of neutron constants requires efficient unit for constants preparation. For this purpose, steady state analysis TRIGEX code (HEX-Z geometry) is used, which includes the program of nuclear data preparation CONSYST connected to the ABBN-93 group constants library. In the paper presented are the results of comparative analytical studies on ULOF beyond design severe accident as applied to the BN-1200 reactor design made by COREMELT-2D code and by its previous version based on neutron kinetics point model. The results of analysis make it possible to evaluate the effect of space-time changes of reactor neutronics caused by sodium removal from the core as a result of sodium boiling. (author)

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

    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.

  11. ATHLET-CD and COCOSYS: the mechanistic computer codes of GRS for simulating severe accidents

    Simulating accident sequences within the framework of safety analyses of nuclear power plants requires the use of deterministic computer codes furnishing the most realistic results (best estimates) in the light of the state of the art. This requirement exists for design basis accidents as well as accidents and events exceeding the design basis. For simulations of reactor behavior and of the source term from the nuclear steam supply system, the ATHLET (Analysis of Thermohydraulics of Leaks and Transients) code has been developed and validated for transients and accidents without major core damage, and the ATHLET-CD (Core Degradation) code has been developed and validated for accidents resulting in major core damage, while the COCOSYS (Containment Code System) code has been developed and validated for the behavior of the containment and the source term for the environment. (orig.)

  12. RAVE code system for 3-D core non-LOCA accident analysis

    Full text of publication follows: This paper provides an overview of the application of the Westinghouse updated RAVE three dimensional (3-D) core transient analysis code system for PWR non-LOCA accident analysis. The RAVE code system consists of a linkage of the following USNRC-approved codes: the EPRI RETRAN-02 (RETRAN) system transient analysis code, the Westinghouse SPNOVA (also referred to as ANC-K) reactor core neutron kinetic nodal code, and the EPRI VIPRE-01 (VIPRE) reactor core thermal-hydraulic (T/H) code. The RETRAN code is used for calculating transient conditions in the reactor coolant system (RCS), including reactor vessel, RCS loops, pressurizer and steam generators. RETRAN also models reactor trips, engineering safety feature (ESF) functions, and the control systems. The SPNOVA code is used to perform 3-D core neutronic calculations for core average power and power distributions in the core. Its reactivity feedback calculation is based on transient fluid conditions and fuel temperatures obtained from the VIPRE code. Based on core inlet temperature, inlet flow and core exit pressure from RETRAN, and the nodal nuclear power from SPNOVA, VIPRE provides back to RETRAN transient nodal heat flux in the reactor core region. An effective 3-D analysis requires RETRAN, SPNOVA and VIPRE calculations to be closely linked for the entire reactor core. The linking architecture uses a standard external communication interface protocol for communication among the running programs on the same or different computers. The RAVE code system currently uses the Parallel Virtual Machine (PVM) software for the data transfer. Besides the necessary changes for data transfer, no other changes were made to RETRAN, SPNOVA or VIPRE fundamental code algorithms or solution methods. The RETRAN model in the RAVE system uses the same detailed reactor vessel, RCS loops, pressurizer, and steam generator, and control and protection models as has been licensed for current plant Safety

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

    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

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

    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.

  15. Analysis of verification and validation problems of calculation means (codes) of accident thermohydrodynamic processes for domestic NPPs

    Analysis of known approaches in area of verification and validation of calculation means (codes) modelling the accident / transition processes in nuclear power plant (NPP) equipment is represented in this review article. Needs to develop and realise the generalised calculation means verification / validation methodology taking into account, together with traditional procedures, the codes applicability assessment criteria for decision of specific tasks and for specific equipment, mathematical models and experimental stands adequacy to full-scale conditions are shown

  16. Development status of Severe Accident Analysis Code SAMPSON

    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)

  17. Development status of Severe Accident Analysis Code SAMPSON

    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)

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

    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.

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

    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.

  20. Artificial intelligence applications in accident management

    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)

  1. Modeling of pipe break accident in a district heating system using RELAP5 computer code

    Reliability of a district heat supply system is a very important factor. However, accidents are inevitable and they occur due to various reasons, therefore it is necessary to have possibility to evaluate the consequences of possible accidents. This paper demonstrated the capabilities of developed district heating network model (for RELAP5 code) to analyze dynamic processes taking place in the network. A pipe break in a water supply line accident scenario in Kaunas city (Lithuania) heating network is presented in this paper. The results of this case study were used to demonstrate a possibility of the break location identification by pressure decrease propagation in the network. -- Highlights: ► Nuclear reactor accident analysis code RELAP5 was applied for accident analysis in a district heating network. ► Pipe break accident scenario in Kaunas city (Lithuania) district heating network has been analyzed. ► An innovative method of pipe break location identification by pressure-time data is proposed.

  2. Code Injection in Web applications

    Shrestha, Bikesh

    2016-01-01

    Code injection is the most critical threat for the web applications. The security vulnerabilities have been growing on web applications. With the growth of the importance of web application, preventing the applications from unauthorized usage and maintaining data integrity have been challenging. Especially those applications which an interface with back-end database components like mainframes and product databases that contain sensitive data can be addressed as the attacker’s main target. ...

  3. A thermo mechanical benchmark calculation of a hexagonal can in the BTI accident with INCA code

    The thermomechanical behaviour of an hexagonal can in a benchmark problem (simulating the conditions of a BTI accident in a fuel assembly) is examined by means of the INCA code and the results systematically compared with those of ADINA

  4. Recent SCDAP/RELAP5 code applications and improvements

    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

  5. Recent SCDAP/RELAP5 code applications and improvements

    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.

  6. Study on severe accidents and countermeasures for WWER-1000 reactors using the integral code ASTEC

    The research field focussing on the investigations and the analyses of severe accidents is an important part of the nuclear safety. To maintain the safety barriers as long as possible and to retain the radioactivity within the airtight premises or the containment, to avoid or mitigate the consequences of such events and to assess the risk, thorough studies are needed. On the one side, it is the aim of the severe accident research to understand the complex phenomena during the in- and ex-vessel phase, involving reactor-physics, thermal-hydraulics, physicochemical and mechanical processes. On the other side the investigations strive for effective severe accident management measures. This paper is focused on the possibilities for accident management measures in case of severe accidents. The reactor pressure vessel is the last barrier to keep the molten materials inside the reactor, and thus to prevent higher loads to the containment. To assess the behaviour of a nuclear power plant during transient or accident conditions, computer codes are widely used, which have to be validated against experiments or benchmarked against other codes. The analyses performed with the integral code ASTEC cover two accident sequences which could lead to a severe accident: a small break loss of coolant accident and a station blackout. The results have shown that in case of unavailability of major active safety systems the reactor pressure vessel would ultimately fail. The discussed issues concern the main phenomena during the early and late in-vessel phase of the accident, the time to core heat-up, the hydrogen production, the mass of corium in the reactor pressure vessel lower plenum and the failure of the reactor pressure vessel. Additionally, possible operator's actions and countermeasures in the preventive or mitigative domain are addressed. The presented investigations contribute to the validation of the European integral severe accidents code ASTEC for WWER-1000 type of reactors

  7. Development of GRIF-SM: The code for analysis of beyond design basis accidents in sodium cooled reactors

    GRIF-SM code was developed at the IPPE fast reactor department in 1992 for the analysis of transients in sodium cooled fast reactors under severe accident conditions. This code provides solution of transient hydrodynamics and heat transfer equations taking into account possibility of coolant boiling, fuel and steel melting, reactor kinetics and reactivity feedback due to variations of the core components temperature, density and dimensions. As a result of calculation, transient distribution of the coolant velocity and density was determined as well as temperatures of the fuel pins, reactor core and primary circuit as a whole. Development of the code during further 6 years period was aimed at the modification of the models describing thermal hydraulic characteristics of the reactor, and in particular in detailed description of the sodium boiling process. The GRIF-SM code was carefully validated against FZK experimental data on steady state sodium boiling in the electrically heated tube; transient sodium boiling in the 7-pin bundle; transient sodium boiling in the 37-pin bundle under flow redaction simulating ULOF accident. To show the code capabilities some results of code application for beyond design basis accident analysis on BN-800-type reactor are presented. (author)

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

    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)

  9. Modeling of DECL accident in the reactor containment by the CONTAIN 2.0 code

    Abbasi, Molood; Rahgoshay, Mohhamad [Islamic Azad Univ., Teheran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

    2013-11-15

    In this paper, a specific type of the Loss of Coolant Accident (LOCA), the DECL (Double Ended Cold Leg) break, that means totally Guillotine type of break in the cold leg pipe, has been modeled. The accident is simulated with the CONTAIN 2.0 code. In the event of a LOCA accident, coolant mass and energy are released to the containment through the break. This causes an increase of pressure and temperature in the containment. The modeling is performed in the VVER-1000 reactor containment. The analysis includes average pressure in the containment and temperature distribution in sample cells in the long-time. The results are compared with the existing reports on studies that used the ANGAR code. Results show that the CONTAIN 2.0 code is an adaptable tool for the analysis of nuclear events such as DECL accident. (orig.)

  10. Modeling of DECL accident in the reactor containment by the CONTAIN 2.0 code

    In this paper, a specific type of the Loss of Coolant Accident (LOCA), the DECL (Double Ended Cold Leg) break, that means totally Guillotine type of break in the cold leg pipe, has been modeled. The accident is simulated with the CONTAIN 2.0 code. In the event of a LOCA accident, coolant mass and energy are released to the containment through the break. This causes an increase of pressure and temperature in the containment. The modeling is performed in the VVER-1000 reactor containment. The analysis includes average pressure in the containment and temperature distribution in sample cells in the long-time. The results are compared with the existing reports on studies that used the ANGAR code. Results show that the CONTAIN 2.0 code is an adaptable tool for the analysis of nuclear events such as DECL accident. (orig.)

  11. Network Coding Fundamentals and Applications

    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

  12. Codes, methods and approaches for accident analyses of the core and fuel behaviour

    Thermohydraulic and fuel behaviour computer codes developed for WWER reactors by the Nuclear Power Plants Research Institute, Trnava (SK), are described. The features of presently used codes PIN, DEFOS-1A, DEFOS-2A, SICHTA, FEMBUL, CALOPEA and DYN3D/M3, their utilization areas, interconnections and the safety analyses procedures are briefly described. General approach in safety simulation and evaluation is given. The interconnections between the proposed criteria - anticipated transients, postulated accidents and cladding failure - are shown. The acceptance criteria of IAEA are checked by the analyses of the transients using the corresponding codes. For most accident analyses, the transient simulation by means of the codes for system transient analysis (RECAP, DYNAMIKA etc.) is sufficient to provide evaluation of the criteria needed. For some transients more detailed analysis is necessary using DYN3D and SICHTA codes (e.g., reactivity initiated accidents). Parameters defining fuel behaviour are determined having in mind that for most of the typical WWER accidents no or very limited damage of fuel assemblies occurred. It allows, on one hand, the use of conservative criteria, and, on the other, to use approach of bounding accidents for proving some criteria like calculated doses below limits, local clad oxidation not exceeding 17% and hydrogen generation below limit. It limits in the current conditions the necessary use of PIN and DEFOS codes to not very large number of analyses. 1 tab., 8 refs

  13. Modelling of severe accident behaviour using the code ATHLET-CD

    Thermal-hydraulic and core degradation phenomena play a decisive role for the course of severe accidents in light water reactors. Therefore, the simulation of such accidents with computer codes requires comprehensive and detailed modelling of these processes. The code ATHLET-CD is being developed for realistic simulation of accidents with core degradation and for evaluation of accident management measures. It makes use of the detailed and validated models of the thermal-hydraulic code ATHLET in an efficient coupling with models for core degradation and fission product behaviour. The capabilities of the coupled code are demonstrated by means of the calculation of the TMI-2 accident. The first three phases of the accident were successfully simulated in a reasonable computing time. The calculated system pressure and pressurizer level after pump trip, during the pump restart, and until core slump are in acceptable agreement with the measured data. The calculated hydrogen generation before the pump restart is in accordance with the deduced value. Contrary to estimates based on the system behaviour, no significant hydrogen generation was calculated during the quench phase. Further model improvements regarding the quenching of degraded core material, fracture and relocation of solid fuel rods, as well as the simulation of debris bed behaviour are necessary for better simulation. (authors)

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

    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

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

    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'.

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

    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'

  17. Conjugate Codes and Applications to Cryptography

    Hamada, M

    2006-01-01

    A conjugate code pair is defined as a pair of linear codes such that one contains the dual of the other. The conjugate code pair represents the essential structure of the corresponding Calderbank-Shor-Steane (CSS) quantum code. It is argued that conjugate code pairs are applicable to quantum cryptography in order to motivate studies on conjugate code pairs.

  18. Evaluation of the General Atomic codes TAP and RECA for HTGR accident analyses

    Ball, S.J.; Cleveland, J.C.; Sanders, J.P.

    1978-04-04

    The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, and the susceptibility of the codes to misuse and input errors is discussed. A critique of code verification plans is also included. The several cases where direct comparisons could be made between TAP/RECA calculations and those based on other independently developed codes indicated generally good agreement, thus contributing to the credibility of the codes.

  19. Evaluation of the General Atomic codes TAP and RECA for HTGR accident analyses

    The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, and the susceptibility of the codes to misuse and input errors is discussed. A critique of code verification plans is also included. The several cases where direct comparisons could be made between TAP/RECA calculations and those based on other independently developed codes indicated generally good agreement, thus contributing to the credibility of the codes

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

    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

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

    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)

  2. Fission product release analysis code during accident conditions of HTGR, RACPAC

    Fission product release analysis code, RACPAC (Fission Product Release Analysis Code from Fuel Particle in Accident Condition), was developed to calculate fractional release from the core during accident conditions of High Temperature Gas-cooled Reactor. RACPAC code has following features. (1) Fission product release fraction after the reactor scram is calculated based on the analytical solution with reduced diffusion coefficient. (2) The reduced diffusion coefficient for each nuclide is calculated from the (R/B) value, which is defined as release rate to birth rate of fission product. (3) The temperature transient after the accident can be taken into consideration in fractional release calculation with RACPAC. This paper describes calculation model of fission product release from fuel particle, calculation model of the reduced diffusion coefficient, users' manual and calculation examples. (author)

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

    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

  4. Computer code application programme of TAEA for thermal hydraulic research

    Evaluation of thermal-hydraulic conditions, fuel behavior, and reactor kinetic during various operating and postulated accident conditions results in conclusions that support decision-making process, the review of license application, and the resolution of other technical issues related to nuclear safety. Also these activities increase the understanding and involvement into new technical developments. Thermal-hydraulic research activities at TAEA focus on the application of computer codes that simulate the behavior of the reactor system. The computer codes are used to analyzed loss of coolant accidents, and system transients in light water nuclear reactors and to assess the consequences if imbalance occurs and to determine the effectiveness of mitigating actions. TAEA has used nuclear reactor system codes (RELAP5/Mod3.2 and higher versions, PARCS) and nuclear plant visual analyzer codes (NPA, SNAP, and XNGR5) obtained by in the framework of the CAMP Agreement signed between TAEA and the United States Nuclear Regulatory Commission (US NRC). TAEA performs and documents the code assessments including improvements and error corrections. Moreover, research activities concerning the passive cooling application and simulations of advanced nuclear power plant have been carried out by both experimental and theoretical means. For example, the experiment test facility, which was designed to investigate the effect of noncondensable gases on condensation, was conducted in cooperation with the Mechanical Engineering Department of the Middle East Technical University in Ankara (Turkey) and was finished. The text matrix obtained from this research was also submitted to US NRC data bank. Application of RELAP5 code f system transients include International Standard Problem (ISP) studies (ISP 33, 38, 42, 45, 46), accident analysis for different reactors and special topics in nuclear heat transfer problems (mid-loop operation). Research studies of severe accidents assess the detailed

  5. EUREKA-2: a computer code for the reactivity accident analysis in a water cooled reactor

    EUREKA-2, a computer code for the reactivity accident analysis, has been developed in order to analyze neutronic, thermal and hydrodynamic transient behaviors in a water cooled reactor. EUREKA-2 can analyze the transient response of the core against the reactivity change caused by control rod withdrawal, coolant flow change and/or coolant temperature change. Especially, it can well simulate fast transient behaviors in serious reactivity accidents. This code calculates coupled neutronic and thermal-hydrodynamic responses for multi-regions in the core. EUREKA-2 has been developed by improving the fluid flow model of EUREKA and can analyze the reactivity accidents in which coolant temperature rises quickly and vapor is produced. (author)

  6. Simulation of rod ejection accident byPARCS code

    Matějková, J.

    2015-01-01

    This paper describes reactor core model used for simulating REA. The model was designed in PARCS utilizing graphical interface SNAP. The data for model were given from benchmark NEACPR L-335. The PARCS model used integrated thermal hydraulic block for calculation. The results and solution is shown in the paper. Thermal hydraulic calculation can also be provided by external system code TRACE. The PARCS model is prepared to couple with TRACE model for giving more accurate calculation.

  7. Adaptation of the ASTEC code system to accident scenarios in fusion installations

    Highlights: ► IRSN has a first version of ASTEC able to model an accident in ITER. ► Models are developed to make possible water/air ingress simulations in the vessel. ► Some thermal-hydraulic calculations in agreement with MELCOR are discussed. -- Abstract: ASTEC is a code system aiming to compute severe accident scenarios and their consequences in nuclear fission Pressurized Water Reactors (PWRs). Its capabilities have been recently extended to address the main accident sequences which may occur in the fusion installations, in particular in ITER. The purpose of this paper is to present a synthesis of the work that has been performed on ASTEC as part of its adaptation to fusion ITER facility, in particular concerning the development of some specific models (dust behavior, jet impaction and wall oxidation), the state of validation of the code and some first calculations for accident transients considered in the basis design. Comparisons with the MELCOR code, selected by ITER for their own safety analysis are provided and show a good agreement between both codes

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

    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 H2/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)

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

    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).

  10. Code portability and data management considerations in the SAS3D LMFBR accident-analysis code

    The SAS3D code was produced from a predecessor in order to reduce or eliminate interrelated problems in the areas of code portability, the large size of the code, inflexibility in the use of memory and the size of cases that can be run, code maintenance, and running speed. Many conventional solutions, such as variable dimensioning, disk storage, virtual memory, and existing code-maintenance utilities were not feasible or did not help in this case. A new data management scheme was developed, coding standards and procedures were adopted, special machine-dependent routines were written, and a portable source code processing code was written. The resulting code is quite portable, quite flexible in the use of memory and the size of cases that can be run, much easier to maintain, and faster running. SAS3D is still a large, long running code that only runs well if sufficient main memory is available

  11. Analysis code for large rupture accidents in ATR. SENHOR/FLOOD/HEATUP

    NONE

    1997-08-01

    In the evaluation of thermo-hydraulic transient change, the behavior of core reflooding and the transient change of fuel temperature in the events which are classified in large rupture accidents of reactor coolant loss, that is the safety evaluation event of the ATR, the analysis codes for thermo-hydraulic transient change at the time of large rupture SENHOR, for core reflooding characteristics FLOOD and for fuel temperature HEATUP are used, respectively. The analysis code system for loss of coolant accident comprises the analysis code for thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC in addition to the above three codes. Based on the changes with time lapse of reactor thermal output and steam drum pressure obtained by the SENHOR, average reflooding rate is analyzed by the FLOOD, and the time of starting the turnaround of fuel cladding tube temperature and the heat transfer rate after the turnaround are determined. Based on these data, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The SENHOR code, the FLOOD code and the HEATUP code and various models for these codes are explained. The example of evaluation and the sensitivity analysis of the ATR plant are reported in the Appendix. (K.I.)

  12. Adaptation of the severe accident codes to VVER-440/V213 (V230) reactor unit, their comparison and utilisation of the results

    This paper presents an application and comparison of the computer codes, devoted for severe accident analysis of PWR up to source term evaluation, to a VVER-440/V213 and V230 NPP. The basic results of selected sequences are described and some physical parameters predicted by different codes are compared. The comparison is deliberated mainly on the timing of main primary circuit events and fission products behaviour up to source term evaluation. Utilisation of the results of the severe accident analysis for development of the emergency procedures for rapid assessment of barriers status and source term category is shortly described, too. (author)

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

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

  14. A simulation of steam generator tube rupture accident by safety analysis code RELAP5/MODI

    Steam-generator-tube-rupture accident occurred at Prairie Island unit 1 is simulated using the RELAP5/MOD1 code which has been developed as a best-estimate safety analysis code for light water reactors. The purpose of the simulation is to examine its capacity as a tool of obtaining high-quality and verified data base needed for developing diagnostic techniques of nuclear power plants. The simulation is conducted until 3200 seconds after the tube rupture. The simulation results agrees fairly well with both the plant records and the RETRAN-02 simulation results conducted at Japan Atomic Energy Research Institute, and it is concluded that the RELAP5/MOD1 code is effective to simulate the overall plant behavior during the accident, although several items remain for future improvement. (author)

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

    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

  16. Parameterization of the driving time in the evacuation or fast relocation model of an accident consequence code

    The model of protective measures in the accident consequence code system UFOMOD of the German Risk Study, Phase B, requires the driving times of the population to be evacuated for the evaluation of the dose received during the evacuation. The parameter values are derived from evacuation simulations carried out with the code EVAS for 36 sectors from various sites. The simulations indicated that the driving time strongly depends on the population density, whereas other influences are less important. It was decided to use different driving times in the consequence code for each of four population density classes as well as for each of three or four fractions of the population in a sector. The variability between sectors of a class was estimated from the 36 sectors, in order to derive subjective probability distributions that are to model the uncertainty in the parameter value to be used for any of the fractions in a particular sector for which an EVAS simulation has not yet been performed. To this end also the impact of the uncertainties in the parameters and modelling assumptions of EVAS on the simulated times was quantified using expert judgement. The distributions permit the derivation of a set of driving times to be used as so-called ''best estimate'' or reference values in the accident consequence code. Additionally they are directly applicable in an uncertainty and sensitivity analysis

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

    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

  18. Code Pointer Masking: Hardening applications against code injection attacks

    Philippaerts, Pieter; Younan, Yves; Muylle, Stijn; Piessens, Frank; Lachmund, Sven; Walter, Thomas

    2011-01-01

    In this paper we present an efficient countermeasure against code injection attacks. Our countermeasure does not rely on secret values such as stack canaries and protects against attacks that are not addressed by state-of-the-art countermeasures of similar performance. By enforcing the correct semantics of code pointers, we thwart attacks that modify code pointers to divert the application's control flow. We have implemented a prototype of our solution in a C-compiler for Linux. The evaluatio...

  19. MELCOR code analysis of a severe accident LOCA at Peach Bottom Plant

    A design-basis loss-of-coolant accident (LOCA) concurrent with complete loss of the emergency core cooling systems (ECCSs) has been analyzed for the Peach Bottom atomic station unit 2 using the MELCOR code, version 1.8.1. The purpose of this analysis is to calculate best-estimate times for the important events of this accident sequence and best-estimate source terms. Calculated pressures and temperatures at the beginning of the transient have been compared to results from the Peach Bottom final safety analysis report (FSAR). MELCOR-calculated source terms have been compared to source terms reported in the NUREG-1465 draft

  20. Licensing applications and topical reviews of the RETRAN computer code

    On September 4, 1984, the US Nuclear Regulatory Commission (NRC) approved, contingent upon completion of correction of errors discovered in the review process, use of RETRAN-02/MOD002 for certain utility applications. Because this generic code review had not included review of plant specific applications, the approval was of the code only, and applicants are required to justify use of the code and its models on a plant and transient specific basis. During and following completion of the generic review, a number of utilities have submitted analyses requesting that the NRC approve the submittals for purposes ranging from simply demonstrating that the applicant was able to use the code, to detailed transient and accident analysis in the licensing arena. The NRC guidance for applicants consists primarily of Generic Letter 83-11 in which an applicant is broadly instructed to perform its own code verification and demonstrate its own technical competence. This broad guidance has led to a wide range of technical detail in submittals and therefore to a correspondingly broad spectrum of NRC approvals. This paper discusses the status of those applications, reviews the breadth of requests and, finally discusses issues that have arisen with respect to these applications

  1. Thermal hydraulic studies of undercooling accidents in LMFBR safety analysis: Codes and validation

    This communication is related to the LMFBR safety analysis of undercooling accidents such as pump run down or total inlet blockage of a subassembly. The authors present the physical models developed for sodium boiling propagation and clad motion and their application to SCARABEE in pile experiments simulating loss of flow accidents in bundle geometry. These studies showed the validity of our description of boiling propagation and improved our understanding of the clad relocation phenomena

  2. Development of Parameter Network for Accident Management Applications

    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

  3. Development of Parameter Network for Accident Management Applications

    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.

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

    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

  5. User's handbook for the iodine severe accident behavior code IMPAIR 2.2

    This publication describes the second version of the Iodine Severe Accident Code (IMPAIR 2.2). This code aims to model postulated conditions of iodine chemistry present in a containment (sump, deposition and atmosphere) during a postulated severe accident in a LWR by using 29 differential equations and 69 rate constants. These equations model the behavior of various iodine species in the sump and the gas phase. Apart from purely chemical equilibria, the mass transport of aerosols, elemental iodine, organoiodine species and droplet carry-over during pressure release while venting are also described. Various improvements and extensions have been made since the original publication. Meanwhile the multi-compartment version, IMPAIR 2/M has become available. It will be updated with the revised models described here as well as other changes. The updated multi-compartment code will be designated IMPAIR 3 and will replace the single-compartment code IMPAIR 2.2. It should be available in 1992. The revised code IMPAIR 2.2, although it has not reached the maturity of an 'assessed code', is significantly improved in its function to model the iodine chemistry of severe accident scenarios in a containment using a mainly phenomenological approach. This improvement was achieved largely through data available from ACE/RTF tests. Two sets of test data with drastic pH difference were used to validate the updated code. The calculated results show a very good correlation for all iodine species in the gas and water phases and in deposition to within an order of magnitude. These validation results were presented at the Third CSNI Workshop on Iodine Chemistry, Tokai-Mura, Japan. (author) 15 tabs., 19 refs

  6. Dispersion of radioactive materials from JRTR following a postulated accident using HOTSPOT code

    Jordan Research and Training Reactor (JRTR) is the first nuclear facility in Jordan. The JRTR is 5 MW, light water moderated and open type pool reactor. In case of an accident, the radioactive materials will be released to the surrounding environment and endanger the people living in the vicinity of the reactor. However, up to now, no study has been published about the dispersion of radioactive materials from JRTR in case of an accident. As preliminary stage for the construction of the JRTR, the dispersion of the radioactive materials from JRTR in case of an accident was studied using HOTSOT code. The result of the report indicates that for ground level release with an average speed of 3.6 m/s of hourly averaged meteorological data for one year with a dominant direction from the west a person located at distance .062 km from the reactor site will receive .25 Sv

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

    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

  8. Preliminary Analysis of a Loss of Condenser Vacuum Accident Using the MARS-KS Code

    In accordance with revision of NUREG-0800 of USNRC, the area of review for loss of condenser vacuum(LOCV) accident has been expanded to analyze both peak pressures of primary and secondary system separately. Currently, the analysis of LOCV accident, which is caused by malfunction of condenser, has been focused to fuel cladding integrity and peak pressure in the primary system. In this paper, accident analysis for LOCV using MARS-KS code were conducted to support the licensing review on transient behavior of secondary system pressure of APR1400 plant. The accident analysis for the loss of condenser vacuum (LOCV) of APR1400 was conducted with the MARS-KS code to support the review on the pressure behavior of primary and secondary system. Total four cases which have different combination of availability of offsite power and the pressurizer spray are considered. The preliminary analysis results shows that the initial conditions or assumptions which concludes the severe consequence are different for each viewpoint, and in some cases, it could be confront with each viewpoint. Therefore, with regard to the each acceptance criteria, figuring out and sensitivity analysis of the initial conditions and assumptions for system pressure would be necessary

  9. Preliminary Analysis of a Loss of Condenser Vacuum Accident Using the MARS-KS Code

    Kim, Jieun Kim; Bang, Young Seok; Oh, Deog Yeon; Kim, Kap; Woo, Sweng-Wong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    In accordance with revision of NUREG-0800 of USNRC, the area of review for loss of condenser vacuum(LOCV) accident has been expanded to analyze both peak pressures of primary and secondary system separately. Currently, the analysis of LOCV accident, which is caused by malfunction of condenser, has been focused to fuel cladding integrity and peak pressure in the primary system. In this paper, accident analysis for LOCV using MARS-KS code were conducted to support the licensing review on transient behavior of secondary system pressure of APR1400 plant. The accident analysis for the loss of condenser vacuum (LOCV) of APR1400 was conducted with the MARS-KS code to support the review on the pressure behavior of primary and secondary system. Total four cases which have different combination of availability of offsite power and the pressurizer spray are considered. The preliminary analysis results shows that the initial conditions or assumptions which concludes the severe consequence are different for each viewpoint, and in some cases, it could be confront with each viewpoint. Therefore, with regard to the each acceptance criteria, figuring out and sensitivity analysis of the initial conditions and assumptions for system pressure would be necessary.

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

    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)

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

    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.

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

    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.

  13. Analysis of TRACY experiment and JCO criticality accident by using AGNES code

    A one-point kinetics code, AGNES, has been developed in JAERI for the purpose of the analysis of TRACY experiment. Four of the experiments performed in ramp feed mode were simulated by AGNES code, and the power, temperature and total fission number were evaluated. The calculated values of them were in agreement with the experimental values with ±15% error. In the analysis of JCO criticality accident, three supposed cases were considered, and the total fission number was evaluated at 4 - 6x1017 by insertion of 1.5 - 3.0$ excess reactivity. (author)

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

    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

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

    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

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

    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.)

  17. A Mobile Application Prototype using Network Coding

    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...

  18. A Mobile Application Prototype using Network Coding

    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...

  19. REACT/THERMIX - a computer code to calculate graphite corrosion due to accidents in pebble-bed reactors

    his report presents the description of the computer code REACT/THERMIX, which was developed for calculations of the graphite corrosion phenomena and accident transients in gas cooled High Temperature Reactors (HTR) under air and/or water ingress accident conditions. The two-dimensional code is characterized by direct coupling of thermodynamic, fluiddynamic and chemical processes with a separate handling of heterogeneous chemical reactions. (orig.)

  20. Mathematical simulation of the RBMK reactor pressure tubes ruptures during accidents: Computer code and verification

    The multiple rupture of the pressure tubes is the most dangerous accident of the channel reactors. There are about 2,000 channels in the RBMK. There exist two potential scenarios: (1) the case of accident when a group of channels becomes overheated; and (2) the case of accident with a rupture of one tube and shock loads on several adjacent channels. The described model considers the prediction technique for potential ruptures according to the first scenario. The probabilistic approach was applied due to existing of substantial scatter and uncertainties in parameters determining pressure tubes deformations and failure in accidents. It was founded on the randomization of the deterministic solution for pressure tube-graphite system deformation and rupture for varied values of chosen chance characters. The mathematical model for the deterministic solution considers the deformation of the system consisting of the pressure tube from the zirconium alloy containing 2.5% of niobium, graphite hard contact rings and graphite blocks. It was solved the common plane strain boundary task. Tube deformation includes three stages: tube deformation until the radial clearance between the tube and graphite disappears; tube deformation with metal flow into the vertical clearance in hard contact rings slits after disappearing of the radial clearance; deformation of the pressure tube-graphite system after closure of the radial clearance up to graphite failure. The mathematical model for the 1st scenario is described. The approach for code verification is also described

  1. Development of Evaluation Technology for Hydrogen Combustion in containment and Accident Management Code for CANDU

    Kim, S. B.; Kim, D. H.; Song, Y. M.; and others

    2011-08-15

    For a licensing of nuclear power plant(NPP) construction and operation, the hydrogen combustion and hydrogen mitigation system in the containment is one of the important safety issues. Hydrogen safety and its control for the new NPPs(Shin-Wolsong 1 and 2, Shin-Ulchin 1 and 2) have been evaluated in detail by using the 3-dimensional analysis code GASFLOW. The experimental and computational studies on the hydrogen combustion, and participations of the OEDE/NEA programs such as THAI and ISP-49 secures the resolving capabilities of the hydrogen safety and its control for the domestic nuclear power plants. ISAAC4.0, which has been developed for the assessment of severe accident management at CANDU plants, was already delivered to the regulatory body (KINS) for the assessment of the severe accident management guidelines (SAMG) for Wolsong units 1 to 4, which are scheduled to be submitted to KINS. The models for severe accident management strategy were newly added and the graphic simulator, CAVIAR, was coupled to addition, the ISAAC computer code is anticipated as a platform for the development and maintenance of Wolsong plant risk monitor and Wolsong-specific SAMG.

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

    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...

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

    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)

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

    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.

  5. Rod ejection accident by the coupled-code system Athlet-Quabox/Cubbox

    The Rod Ejection Accident (REA) is the most limiting case among Reactivity Induced Accident (RIA). Due to the fast reactivity insertion which can lead to prompt criticality and thus to a sharp fuel enthalpy increase in the affected part of the core, REA can cause severe fuel damage. The REA is usually an asymmetric transient where neutron kinetics and the thermohydraulics are strongly coupled (through Doppler feedback). This poster shows results of simulations that have been performed on a generic PWR core with UOX/MOX loading with the coupled code Athlet-Quabox/Cubbox. It is shown the importance of different parameters like the delayed neutron fraction, the initial power level and the nuclear data uncertainties. (A.C.)

  6. JERICHO computer code: PWR containment response during severe accidents description and sensitivity analysis

    The JERICHO code has been developed in order to study the thermodynamic behaviour inside the reactor containment building for the complete spectrum of accident sequences likely to occur in such a reactor, including models for the various mass and energy transfer phenomena, for water spray, for hydrogen and carbon monoxide flammability limits and combustion, as well as for containment venting. Sensitivity analyses have been performed on a severe accident sequence, (namely, small LOCA with failure of the emergency core cooling and containment spray systems), involving core melting and subsequent concrete containment basemat erosion. The effect of various models, such as mass and energy transfer to the structures, has been studied. The influence of the concrete composition, of the fission product deposition and of the thermal degradation of the reactor cavity concrete walls on long term thermodynamic behaviour has also been investigated

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

    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...

  8. Application of simulation techniques for accident management training in nuclear power plants

    core. These capabilities include the optimized use of design margins as well as complementary measures for the prevention of accident progression, its monitoring, and the mitigation of severe accidents. Finally, level 5 includes off-site emergency response measures, the objective of which is to mitigate the radiological consequences of significant releases of radioactive material. Accident management is defined in the IAEA Safety Report on Development and Implementation of Accident Management Programmes in Nuclear Power Plants. The IAEA definitions are in line with the definitions of severe accident management in OECD/NEA documents as given, for example. This report describes simulation techniques used in the training of personnel involved in accident management of NPPs. This concerns both the plant personnel and the persons involved in the management of off-site releases. The report pertains to light water reactors (LWRs) and pressurized heavy water reactors (PHWRs), but it can equally be applied to power reactors of other types. The report is intended for use by experts responsible for planning, developing, executing or supervising the training of personnel involved in the implementation of AMPs in NPPs. It concentrates on existing techniques, but future prospects are also discussed. Various simulation techniques are considered, from incorporating graphical interfaces into existing severe accident codes to full-scope replica simulators. Both preventive and mitigative accident management measures, different training levels and different target personnel groups are taken into account. Based on the available information compiled worldwide, present views on the applicability of simulation techniques for the training of personnel involved in accident management are provided in this report. Apart from the introduction, this report consists of four sections and three appendices. In Section 2, specific aspects of accident management are summarized. Basic approaches in the

  9. SOPHAEROS code development and its application to falcon tests

    One of the key issues in source-term evaluation in nuclear reactor severe accidents is determination of the transport behavior of fission products released from the degrading core. The SOPHAEROS computer code is being developed to predict fission product transport in a mechanistic way in light water reactor circuits. These applications of the SOPHAEROS code to the Falcon experiments, among others not presented here, indicate that the numerical scheme of the code is robust, and no convergence problems are encountered. The calculation is also very fast being three times longer on a Sun SPARC 5 workstation than real time and typically ∼ 10 times faster than an identical calculation with the VICTORIA code. The study demonstrates that the SOPHAEROS 1.3 code is a suitable tool for prediction of the vapor chemistry and fission product transport with a reasonable level of accuracy. Furthermore, the fexibility of the code material data bank allows improvement of understanding of fission product transport and deposition in the circuit. Performing sensitivity studies with different chemical species or with different properties (saturation pressure, chemical equilibrium constants) is very straightforward

  10. International Code Assessment and Applications Program: Annual report

    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

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

    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)

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

    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)

  13. Model verification of the debris coolability analysis module in the severe accident analysis code 'SAMPSON'

    The debris coolability analysis module in the severe accident analysis code 'SAMPSON' has been enhanced to predict more mechanistically the safety margin of present reactor pressure vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris cooling after spreading is solved on the basis of natural convection analysis with melting and solidification on three-dimensional Cartesian co-ordinates. The calculated results for the cooling model are compared with the results from a three-dimensional natural convection experiment. The comparisons show the module capability for predictions of the debris temperature in the cooling process. Furthermore, it is seen that the prediction capability in the thermal load of the vessel wall is improved, since the penetration nozzles melting is modeled and combined with the cooling model. The module provides a good tool for the prediction of the reactor safety margin in a severe accident through the three-dimensional analysis of debris cooling. (author)

  14. Modelling of Core Degradation and Progression of Severe Accident by Using MELCOR Code

    After Fukushima Daiichi Nuclear Accident, every single nuclear-field organization in the world focused in the analysis and study of scenarios that leads to core damage and hydrogen releases, in this way the integrated code MELCOR is used by the Mexican Regulatory Body as a tool in the analysis of severe accident progression, core melting and degradation. Scenarios related to core melting could provide information that show important parameters such as: time to reach the core damage, time window for level recovery, etc. This information is useful in the analysis of progression for this kind of events. In this work, Mexican Regulatory Body presents two simulations for different scenarios: a) Station Blackout with no cooling water injection and b) Station Blackout with late cooling water injection. Those two scenarios enclose the response of the fuel under Severe Accident conditions (progression of melting, relocation, temperature profile), plots in this document are qualitative items that allow to analyze the behavior for fuel/core elements. (author)

  15. Evaluation of finite element codes for demonstrating the performance of radioactive material packages in hypothetical accident drop scenarios

    Drop testing and analysis are the two methods for demonstrating the performance of packages in hypothetical drop accident scenarios. The exact purpose of the tests and the analyses, and the relative prominence of the two in the license application, may depend on the Competent Authority and will vary between countries. The Finite Element Method (FEM) is a powerful analysis tool. A reliable finite element (FE) code when used correctly and appropriately, will allow a package's behaviour to be simulated reliably. With improvements in computing power, and in sophistication and reliability of FE codes, it is likely that FEM calculations will increasingly be used as evidence of drop test performance when seeking Competent Authority approval. What is lacking at the moment, however, is a standardised method of assessing a FE code in order to determine whether it is sufficiently reliable or pessimistic. To this end, the project Evaluation of Codes for Analysing the Drop Test Performance of Radioactive Material Transport Containers, funded by the European Commission Directorate-General XVII (now Directorate-General for Energy and Transport) and jointly performed by Arup and Gesellschaft fuer Nuklear-Behaelter mbH, was carried out in 1998. The work consisted of three components: Survey of existing finite element software, with a view to finding codes that may be capable of analysing drop test performance of radioactive material packages, and to produce an inventory of them. Develop a set of benchmark problems to evaluate software used for analysing the drop test performance of packages. Evaluate the finite element codes by testing them against the benchmarks This paper presents a summary of this work

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

    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

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

    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

    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. Source term code package: modifications and applications

    The Source Term Code Package (STCP) and its development have served a pivotal role in the advancement of source term research by helping to focus both modeling and experimental efforts on issues that are of controlling importance in their ability to predict accident consequences. Currently efforts are continuing with the goal of upgrading various STCP code features. The need for these changes results from the continuing research and growing data base in the area of fission product release and transport. Among the models which are being upgraded are those dealing with direct containment heating, fission product release from the fuel, combined analyses of containment transport, and pool/ice condenser removal, and more direct and adequate coupling of heat transfer and fission product transport in the reactor coolant system

  20. Comparison of Severe Accident Results Among SCDAP/RELAP5, MAAP, and MELCOR Codes

    This paper demonstrates a large-break loss-of-coolant accident (LOCA) sequence of the Kuosheng nuclear power plant (NPP) and station blackout sequence of the Maanshan NPP with the SCDAP/RELAP5 (SR5), Modular Accident Analysis Program (MAAP), and MELCOR codes. The large-break sequence initiated with double-ended rupture of a recirculation loop. The main steam isolation valves (MSIVs) closed, the feedwater pump tripped, the reactor scrammed, and the assumed high-pressure and low-pressure spray systems of the emergency core cooling system (ECCS) were not functional. Therefore, all coolant systems to quench the core were lost. MAAP predicts a longer vessel failure time, and MELCOR predicts a shorter vessel failure time for the large-break LOCA sequence. The station blackout sequence initiated with a loss of all alternating-current (ac) power. The MSIVs closed, the feedwater pump tripped, and the reactor scrammed. The motor-driven auxiliary feedwater system and the high-pressure and low-pressure injection systems of the ECCS were lost because of the loss of all ac power. It was also assumed that the turbine-driven auxiliary feedwater pump was not functional. Therefore, the coolant system to quench the core was also lost. MAAP predicts a longer time of steam generator dryout, time interval between top of active fuel and bottom of active fuel, and vessel failure time than those of the SR5 and MELCOR predictions for the station blackout sequence. The three codes give similar results for important phenomena during the accidents, including SG dryout, core uncovery, cladding oxidation, cladding failure, molten pool formulation, debris relocation to the lower plenum, and vessel head failure. This paper successfully demonstrates the large-break LOCA sequence of the Kuosheng NPP and the station blackout sequence of the Maanshan NPP

  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

    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. Analysis code for medium and small rupture accidents in ATR. LOTRAC/HEATUP

    NONE

    1997-08-01

    In the evaluation of thermo-hydraulic and fuel temperature transient changes in the events which are classified in medium and small rupture accidents of reactor coolant loss that is the safety evaluation event of the ATR, the analysis code for synthetic thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC and the detailed analysis code for fuel temperature HEATUP are used, respectively. By using the LOTAC, the thermo-hydraulic behavior of reactor cooling facility and the temperature behavior of fuel at the time of blow-down are analyzed, and also the characteristics of changing reactor thermal output is analyzed, considering the functioning characteristics of emergency core cooling system. Based on the data of thermo-hydraulic behavior obtained by the LOTRAC, the time of beginning the turn-around of fuel cladding tube temperature obtained by the data of ECCS pouring characteristics, the heat transfer rate after the turn-around and so on, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The LOTRAC code, the HEATUP code, various analysis models, and rupture simulation experiment are reported. (K.I.)

  3. Model Development of Light Water Reactor Fuel Analysis Code RANNS for Reactivity-initiated Accident Conditions

    A light water reactor fuel analysis code RANNS has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly Reactivity-Initiated Accident (RIA) conditions, based on the light water reactor fuel analysis code FEMAXI-7, which has been developed for normal operation conditions and anticipated transient conditions. The recent model development for the RANNS code has been focused on improving predictability of stress, strain, and temperature inside a fuel rod during pellet cladding mechanical interaction (PCMI), which is one of the most important behaviors of high-burnup fuels under RIA conditions. This report provides descriptions of the models developed and/or validated recently via experimental analyses using the RANNS code on the RIA-simulating experiments conducted in the Nuclear Safety Research Reactor (NSRR): models for mechanical behaviors as relocation of fuel pellets, pellet yielding, pellet-cladding mechanical bonding, and PCMI failure limit of fuel cladding, and thermal behaviors as pellet-cladding gap conductance and heat transfer from fuel rod surface to coolant water. (author)

  4. Status on development and verification of reactivity initiated accident analysis code for PWR (NODAL3)

    A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the nodal few-group neutron diffusion theory in 3-dimensional Cartesian geometry for a typical pressurized water reactor (PWR) static and transient analyses, especially for reactivity initiated accidents (RIA). The spatial variables are treated by using a polynomial nodal method (PNM) while for the neutron dynamic solver the adiabatic and improved quasi-static methods are adopted. A simple single channel thermal-hydraulics module and its steam table is implemented into the code. Verification works on static and transient benchmarks are being conducting to assess the accuracy of the code. For the static benchmark verification, the IAEA-2D, IAEA-3D, BIBLIS and KOEBERG light water reactor (LWR) benchmark problems were selected, while for the transient benchmark verification, the OECD NEACRP 3-D LWR Core Transient Benchmark and NEA-NSC 3-D/1-D PWR Core Transient Benchmark (Uncontrolled Withdrawal of Control Rods at Zero Power). Excellent agreement of the NODAL3 results with the reference solutions and other validated nodal codes was confirmed. (author)

  5. Quantification of severe accidents source terms of BWR 4 reactor with Mark I containment using source term code package

    Severe accident source terms of a nuclear power plant which employs a BWR4 reactor with a Mark I containment are quantified with the Source Term Code Package (STCP). Accident scenarios selected for source terms analyses are defined based on the Probabilistic Risk Assessment (PRA) results of accident sequence grouping, containment responses, containment phenomenological event trees, and release category analyses of studies. Included in the paper is a brief description of the structure and major features of STCP together with the modifications made to the code package for the present analysis, the plant model adopted for the STCP source terms quantifications; a presentation and discussion of the source terms as predicted by the STCP for the ten accident sequences analyzed. (orig.)

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

    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

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

    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

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

    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.

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

    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

  10. Validation of the thermal hydraulic computer code S-RELAP5 for performing loss-of-coolant accident analysis (LOCA) in Pressurized Water Reactors (PWRs)

    Siemens Power Corporation (SPC) has developed S-RELAP5, a RELAP5/MOD2 based thermal hydraulic system code with main modifications and improvements relative to RELAP5/MOD2 concerning Multi-Dimensional Capability, Energy Equations, Numerical Solution of Hydrodynamic, Constitutive Models, Heat Transfer Models, Chocked Flow, and Counter-Current Flow Limiting. S-RELAP5 was exercised over a range of integral and separate effects tests in order to demonstrate that the code could predict the important phenomena associated with PWR LBLOCA. A methodology for calculation of statistical uncertainties has been developed and applied to analyses of hypothetical large break loss-of-coolant accidents (LBLOCA). To extend the application capability of S-RELAP5 to small break loss-of-coolant accidents problems (SBLOCA) an investigation program for appropriate experiments was launched and largely carried out. (author)

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

    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

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

    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.

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

    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.

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

    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.

  15. The OECD/CSNI/WGFS Benchmark on Reactivity Initiated Accident Fuel Codes

    Reactivity-initiated accident (RIA) fuel rod codes have been developed for a significant period of time and they all have shown their ability to reproduce some experimental results with a certain degree of adequacy. However, they sometimes rely on different specific modeling assumptions the influence of which on the final results of the calculations is difficult to evaluate. In order to contribute to the assessment of these codes, the Working Group on Fuel Safety (WGFS) of the OECD/NEA organized a benchmark. This exercise was based on a consistent set of four experiments on very similar highly irradiated fuel rods tested under different experimental conditions in the NSRR and CABRI test reactors. The participation to the benchmark has been very important: 17 organizations representing 14 countries provided solutions for some or all the cases that were defined. In terms of computer codes used, the spectrum was also large as solutions were provided with FALCON, FEMAXI, FRAPTRAN, RANNS, RAPTA, SCANAIR, TESPAROD and TRANSURANUS. This paper describes the main conclusions drawn from this benchmark. (author)

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

    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.)

  17. Analysis of Fukushima Daiichi Nuclear Power Plant by SAMPSON severe accident code - Unit 3

    On March 11th 2011 an extremely high magnitude earthquake and following tsunami struck the East coast of Japan, resulting in a nuclear accident unprecedented in time and extent. After scram started at all power stations, diesel generators began operation until tsunami waves reached the power plants. Flooding by tsunami had a great impact on the plant safety systems availability, leading to the condition of station black out at Fukushima Daiichi from unit 1 to 3. In the present work the severe accident code SAMPSON is employed for the analysis of the first part of the transient in Fukushima Daiichi Unit 3. In this unit DC batteries remained available for about 40 hours after scram and influenced the time of melting onset, hydrogen release in the reactor building and further explosion. Models of high pressure safety systems were improved in SAMPSON, considering a more realistic pump-turbine unit operation and communication between reactor and containment. Moreover new suppression pool stratification and spray models were developed and implemented in the containment module, showing a great impact in the drywell pressure estimation. It has been shown how the computed results of pressure (in reactor vessel and drywell) and core water level show a fair agreement with the measurement data and notably improvements compared to the previous analyses. (author)

  18. Application of RS Codes in Decoding QR Code

    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.

  19. Applicability of health physics lessons learned from the Three Mile Island Unit 2 accident to the Fukushima Daiichi accident

    The TMI-2 and Fukushima Daiichi accidents appear to be dissimilar because they involve different reactor types. However, the health physics related lessons learned from TMI-2 are applicable, and can enhance the Fukushima Daiichi recovery effort. - Highlights: ► TMI-2 health physics lessons learned are applicable to Fukushima Daiichi Accident. ► Fuel damage sequence of these accidents are similar. ► On-site recovery actions will be similar, but Fukushima Daiichi is more demanding. ► Offsite recovery actions are significantly more challenging at Fukushima Daiichi.

  20. MARS code developments, benchmarking and applications

    Recent developments of the MARS Monte Carlo code system for simulation of hadronic and electromagnetic cascades in shielding, accelerator and detector components in the energy range from a fraction of an electron volt up to 100 TeV are described. The physical model of hadron and lepton interactions with nuclei and atoms has undergone substantial improvements. These include a new nuclear cross section library, a model for soft prior production, a cascade-exciton model, a dual parton model, deuteron-nucleus and neutrino-nucleus interaction models, a detailed description of negative hadron and muon absorption, and a unified treatment of muon and charged hadron electro-magnetic interactions with matter. New algorithms have been implemented into the code and benchmarked against experimental data. A new Graphical-User Interface has been developed. The code capabilities to simulate cascades and generate a variety of results in complex systems have been enhanced. The MARS system includes links to the MCNP code for neutron and photon transport below 20 MeV, to the ANSYS code for thermal and stress analyses and to the STRUCT code for multi-turn particle tracking in large synchrotrons and collider rings. Results of recent benchmarking of the MARS code are presented. Examples of non-trivial code applications are given for the Fermilab Booster and Main Injector, for a 1.5 MW target station and a muon storage ring

  1. MARS code developments, benchmarking and applications

    Recent developments of the MARS Monte Carlo code system for simulation of hadronic and electromagnetic cascades in shielding, accelerator and detector components in the energy range from a fraction of an electronvolt up to 100 TeV are described. The physical model of hadron and lepton interactions with nuclei and atoms has undergone substantial improvements. These include a new nuclear cross section library, a model for soft pion production, a cascade-exciton model, a dual parton model, deuteron-nucleus and neutrino-nucleus interaction models, a detailed description of negative hadron and muon absorption, and a unified treatment of muon and charged hadron electromagnetic interactions with matter. New algorithms have been implemented into the code and benchmarked against experimental data. A new Graphical-User Interface has been developed. The code capabilities to simulate cascades and generate a variety of results in complex systems have been enhanced. The MARS system includes links to the MCNP code for neutron and photon transport below 20 MeV, to the ANSYS code for thermal and stress analyses and to the STRUCT code for multi-turn particle tracking in large synchrotrons and collider rings. Results of recent benchmarking of the MARS code are presented. Examples of non-trivial code applications are given for the Fermilab Booster and Main Injector, for a 1.5 MW target station and a muon storage ring. (author)

  2. RAPTA-5 code: Modelling behaviour of WWER-type fuel rods in design basis accidents verification calculations

    RAPTA-5 code used for licensing calculations to validate the compliance with the requirements for WWER fuel safety in design basis accidents. The characteristic results are given of design modelling experiments simulating thermomechanical and corrosion behaviour of WWER and PWR fuel rods in LOCA. The results corroborate the adequate predictability of both individual design models and the code as a whole. (author). 14 refs, 12 figs

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

    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

  4. Modification and validation of ATHLET code for sodium-cooled fast reactor application

    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 fur 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. (authors)

  5. Applications of probabilistic accident consequence evaluation in Cuba

    Are presented the approaches and results of the application of Accident Consequence Evaluation methodologies in on emergency in the Juragua Nuclear Power Plant site and a population evaluation of a planned NPP site in the east of the country Findings on population sector weighing and assessment of effectiveness of primary countermeasures in the event of sever accidents (SST1 and PWR4 source terms) in Juragua NPP site are discussed Results on comparative risk-based evaluation of the population predicted evolution (in 3 temporal horizons: base year, 2005 year and 2050 year) for the planned site are described. Evaluation also included sector risk weighing, risk importance of small towns in the nearby of the effects on risk of population freezing and relocation of these villages

  6. ETF system code: composition and applications

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

  7. Application of ITER Safety Analysis for KSTAR : Tritium Leakage from Fusion Power Termination System Failure Accident with MELCOR

    This extreme reactor condition makes serious material limitation and emphasizes the importance of safety analysis. To get permission of construction license, previous researches like preliminary safety research have been analyzed risk assessments of fusion reactors. To simulate the severe accidents in fusion reactor, a number of thermal hydraulic simulation codes were used(ECART, INTRA, ATHENA/RELAP and so on). Before construction, to obtain ITER license about safety issue, MELCOR is chosen as the thermal hydraulic code to be used to simulate radioactive material release from severe accidents. Capability of the simulation code in severe accident analysis is to simulate the cooling system in ITER, the transport of radionuclides during design basis accidents (DBAs) including beyond design basis accidents (BDBAs). MELCOR is fully integrated code that models the accidents in Light Water Reactor (LWR). To analyze the accidents in ITER, MELCOR 1.8.2 version is modified. The amount of release radioactive material is safety acceptance criteria in the nuclear fusion system. There are three kinds of radioactive materials in fusion reactor; tritium (or Tiritiated water: HTO), activation products from divertor or first-wall(AP) and activated corrosion products(ACP). In generic Site Safety Report (GSSR), table I lists the release guidelines for tritium and activation products for normal operation, incidents, and accidents. This small scale facility makes the experimental flexibility to develop fusion technology. Fusion source difference between KSTAR and ITER is D-D(Deuterium- Deuterium reaction) fusion and D-T(Deuterium- Tritium reaction) fusion. This D-D fusion makes Tritium in the 50 percent chance. The radioactivity of tritium is small to consider, but, the accident analysis is indispensable. In the present work, the conservatively estimated tritium inventory in KSTAR is used with one of the most severe accident in ITER; Fusion Power Termination System(FPTS) failure with

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

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

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

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

    2015-04-15

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

  10. Multiple application coded switch development report

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

    1979-03-01

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

  11. Modeling of a confinement bypass accident with CONSEN, a fast-running code for safety analyses in fusion reactors

    Highlights: • The CONSEN code for thermal-hydraulic transients in fusion plants is introduced. • A magnet induced confinement bypass accident in ITER has been simulated. • A comparison with previous MELCOR results for the accident is presented. -- Abstract: The CONSEN (CONServation of ENergy) code is a fast running code to simulate thermal-hydraulic transients, specifically developed for fusion reactors. In order to demonstrate CONSEN capabilities, the paper deals with the accident analysis of the magnet induced confinement bypass for ITER design 1996. During a plasma pulse, a poloidal field magnet experiences an over-voltage condition or an electrical insulation fault that results in two intense electrical arcs. It is assumed that this event produces two one square meters ruptures, resulting in a pathway that connects the interior of the vacuum vessel to the cryostat air space room. The rupture results also in a break of a single cooling channel within the wall of the vacuum vessel and a breach of the magnet cooling line, causing the blow down of a steam/water mixture in the vacuum vessel and in the cryostat and the release of 4 K helium into the cryostat. In the meantime, all the magnet coils are discharged through the magnet protection system actuation. This postulated event creates the simultaneous failure of two radioactive confinement barrier and it envelopes all type of smaller LOCAs into the cryostat. Ice formation on the cryogenic walls is also involved. The accident has been simulated with the CONSEN code up to 32 h. The accident evolution and the phenomena involved are discussed in the paper and the results are compared with available results obtained using the MELCOR code

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

    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.

  13. Spectroscopic calculation code ASPECT and its application

    The Code ASPECT is available for calculations of electronic levels of atoms and ions by the intermediate coupling scheme. This scheme is characterized by the simultaneous diagonalization of Hamiltonians for electronic repulsion, spin orbit interaction and crystal field effect. ASPECT performs the sorting of microstates involved in the electronic configuration in problem, calculation of matrix elements of these Hamiltonians, and diagonalization of the summed matrix. As input data, the calculation needs only parameter values of Slater integrals. ASPECT is also applied to calculate transition probabilities between the electronic levels obtained by this code. ASPECT is particularly focused on complex configurations containing f-electrons as met in Lanthanides and Actinides, which are not easily treated by an algebraic method. For convenience of users, Slater integral values for configurations fn of Lanthanides and Actinides are installed in the code so that users may select merely the atomic number. This document is composed of three parts. The first part (Chapter 1-3) describes quantum mechanical principles to calculate matrix elements of each unperturbed Hamiltonian and transition probabilities. The second part (Chapter 4) explains the structure of the code, and the last part (Chapter 5) serves as the manual for applications of this code, in which some samples are included. The third part (Chapter 6) is added as supplement for users who will improve this code. (author)

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

    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

  15. A computer code (WETBERAN) for wet sequence behavior of radioactive nuclides in LWR plant at accident conditions

    The WETBERAN code has been developed to simulate the isotopic- and time-dependent behavior fission products (FP) which leak through the multiple paths of liquid and gas flow within an LWR plant under accident conditions. In this code, emphasis is put on the phenomena pertinent to the presence of water. The TMI, SL-1, and Ginna accidents are analyzed to show the code capability. The TMI 40 day analysis gives detailed informations of FP behavior, both leaking from and remaining in the plant, and proves the effectiveness of the network model for describing the multiple leakage paths. The SL-1 analysis is made to study halogen reduction by water, which cannot be taken into account by CORRAL. The Ginna analysis has been made to check iodine transport by droplets usually generated by primary water flashing at SG tube rupture

  16. Overview of the IMPACT severe accident analysis code SAMPSON - On core degradation and lower plenum debris behavior in the main

    The first version of the IMPACT-SAMPSON code was completed. SAMPSON is the best estimate integral code for severe accident analysis with modular structure. Each module can run independently and communication with multiple analysis modules supervised by the analysis control module makes an integral analysis possible while appearing to users to be a single code. Multi-dimensional mechanistic models and theoretical-base equations were applied. An execution of enormous amount of calculation steps becomes possible with the use of a parallel processing computer. Models in each module were verified by test analyses. Final integral verification by PHEBUS test analyses is in progress and integral analyses of light water nuclear power plants will be performed to demonstrate quantitatively that adequate safety margin exists to cope with severe accidents. (authors)

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

    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. Network Coding Protocols for Data Gathering Applications

    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...

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

    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)

  20. Accident analysis of Fukushima Daiichi NPP Unit-1 with SAMPSON code

    The progress of the core disruption of the Fukushima Daiichi NPP Unit-1 was analyzed by the severe accident analysis code SAMPSON. The code includes new modellings of the phenomena that occurred which have been deemed specific to the Fukushima Daiichi NPP: (1) steam leakage from the gasket of the safety relief valve (SRV) and from the buckling portion of the guide tubes (GTs) of some in-core monitors (source range monitors (SRMs) and intermediate range monitors (IRMs)): (2) melting of SRM/IRM GTs at the bottom of the reactor pressure vessel (RPV); and (3) incorporation of continuous drainage pathways for debris relocation. During the early phase of the accident after the reactor scram, the isolation condensers (ICs) had intermittently worked until the loss of AC and DC power supplies by the tsunami. The analysis reproduced well the RPV pressure transient during the IC operation period. After the loss of AC and DC power supplies, the SRV had repeated its opening and closing to keep the RPV pressure constant at about 7.5 MPa for about 4.5 hours, resulting in a gradual decrease of water level in the core. Then the SRV stopped working due to depressurization by the direct steam release from the buckling portions of the SRM/IRM GTs and from the SRV gasket. The eutectic B4C (control rod material) and steel reacted, resulting in the initiation of melting at about 4.5 h after the scram when the collapsed water level was getting closer to the bottom of active fuel, followed by melting of steel, zircalloy, and eutectics of UO2+Zr. All 12 SRM/IRM GTs had sequentially melted at about 6.5 h after the scram, resulting in fall down of melts onto the pedestal floor. Since there was no intentional core cooling for about 14 hours after the termination of the ICs until the alternative water injection by a fire engine, the core disruption continued. When the alternative water injection was started at 05:46, March 12 (15 h after the scram), 85% of the core materials had already become

  1. Severe accident research at the Transuranium Institute Karlsruhe: A review of past experience and its application to future challenges

    Highlights: • Severe accident research at the Transuranium Institute, Karlsruhe has been reviewed. • Large (Phébus, TMI-2) and smaller tests have improved understanding of core degradation. • Cladding/structural materials interaction and attack of fuel are important in degradation. • Formation and composition of molten fuel pool in the lower bundle was reproducible. • This mechanistic knowledge has greatly assisted severe accident modelling. - Abstract: With the current situation in Japan one should examine previous research into severe accidents and the current state of European severe accident research to assess what are the priorities for research for existing and future nuclear reactors. The European Commission’s SARNET 2 (Severe Accident NETwork of Excellence) programme and its SARP (Severe Accident Research Priorities) assessments have been made and have outlined the future needs as seen from the EU point of view. There is already considerable research that will be very valuable in analysing and guiding the investigation and remediation activities at Fukushima Dai-ichi. This includes investigations into previous major accidents and international severe fuel damage projects. Facilities using analogue materials are able to analyse large-scale behaviour of materials, while smaller-scale testing of irradiated fuel for detailed property measurements are important for mechanistic studies. The final (and very important) aspect is application of this information to formulate codes to model the identified mechanisms and also to have their predictions validated by the data. This paper will take examples from the Transuranium Institute’s (ITU Karlsruhe’s) contribution to projects such as the TMI-2 accident investigation and the Phébus PF bundle and fission product deposit investigations as well as some of the smaller scale testing and modelling support that ITU has performed over the last 20 years. This will show what has been learnt about fuel and

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

    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

  3. Investigation of NPP behavior in case of loss of coolant accident based on comparison of different ASTEC computer code versions

    The paper presents the work performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) and Bhabha Atomic Research Centre (BARC), India in the frame of SARNET2 project. The performed work continues the effort in the field of nuclear safety and cooperation between INRNE-BAS and BARC. The main target is development and validation of ASTEC (Accident Source Term Evaluation Code) at the further, a tool for level-2 PSA analysis for better understanding of accident progression during in-vessel phase until reactor vessel failure. (authors)

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

    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

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

    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.

  6. Writing robust C++ code for critical applications

    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...

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

    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

  8. Rod Ejection Accident by the Coupled System Code ATHLET-QUABOX/CUBBOX

    Perin, Yann; Velkov, Kiril; Pasichnyk, Igor; Langenbuch, Siegfried

    The paper considers a Rod Ejection Accident (REA) which has been calculated by the coupled-code system ATHLET-QUABOX/CUBBOX. For the present study, a MOX/UOX mixed core loading was developed on the basis of a generic PWR. The results are particularly focused on the fuel enthalpy rise which is the main safety criterion for such transient. A parametric REA study has been performed, showing the influence of some important thermal-hydraulic and neutron-physical parameters. Simulations have been performed using realistic or artificially decreased delayed neutron fractions for two different core states (HZP and 30% of the nominal power). Effective fuel rod temperature influence (i.e. Doppler coefficient) has been studied by using different correlations (0.5/0.5 weighting factors or the typical TDoppler = 0.7 TSurface + 0.3 TCenter) or by changing the fuel gap conductance. It is shown that the maximum enthalpy (and enthalpy increase) does not always appear in the affected fuel assembly but can also appear in the neighboring ones. This result is a direct consequence of the burn up dependence of the enthalpy. The paper also considers the case of local delayed neutron parameters and briefly describes the future REA studies foreseen at GRS such as an investigation of quantitative uncertainty propagation from the nuclear data to the transient behavior.

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

    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

  10. Application of the MELCOR code to design basis PWR large dry containment analysis.

    Phillips, Jesse; Notafrancesco, Allen (USNRC, Office of Nuclear Regulatory Research, Rockville, MD); Tills, Jack Lee (Jack Tills & Associates, Inc., Sandia Park, NM)

    2009-05-01

    The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

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

    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

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

    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)

  13. Simulation of experiment on aerosol behaviour at severe accident conditions in the LACE experimental facility with the ASTEC CPA code

    The experiment LACE LA4 on thermal-hydraulics and aerosol behavior in a nuclear power plant containment, which was performed in the LACE experimental facility, was simulated with the ASTEC CPA module of the severe accident computer code ASTEC V1.2. The specific purpose of the work was to assess the capability of the module (code) to simulate thermal-hydraulic conditions and aerosol behavior in the containment of a light-water-reactor nuclear power plant at severe accident conditions. The test was simulated with boundary conditions, described in the experiment report. Results of thermal-hydraulic conditions in the test vessel, as well as dry aerosol concentrations in the test vessel atmosphere, are compared to experimental results and analyzed. (author)

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

    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)

  15. Applicability research of RELAP5/MOD3.3 for small break loss of coolant accident of NPP with passive safety system

    The passive core cooling system is used in AP1000 to mitigate the small break loss of coolant accident (SBLOCA). The RELAP5/MOD3.3 code is generally applicable to the traditional NPP SBLOCA research, but for the passive NPP SBLOCA, its applicability will need further study and evaluation. Based on the analysis of the important phenomenon of the SBLOCA of the passive NPP, the RELAP5/MOD3.3 code was assessed and modified. In order to verify the applicability of the modified RELAP5/MOD3.3 code, the DBA-02 and NRC-05 cases of APEX-1000 which was the test facility for verifying AP1000 small break loss of coolant accident, were simulated. It shows good agreement between the results of the modified RELAP5/MOD3.3 code and experiment data. (authors)

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

    SAMPSON was designed as a large scale simulation system of inter-connected hierarchical modules covering a wide spectrum of scenarios ranging from normal operation to severe accidents. The code was validated by a wide range of analyses for separate-effect tests, and integral tests mainly through participation in the Organisation for Economic Co-operation and Development projects. In the previous analysis of TEPCO’s Fukushima Daiichi Nuclear Power Plant (1F) with the SAMPSON code, melt retention at a core plate was assumed based on observations after the Three Mile Island Unit 2 accident. The melt relocation to the core plate occurred when the water level was below the core plate in the SAMPSON analysis of the 1F accident. Therefore debris relocation phenomena were investigated using the Molten Core Relocation Analysis (MCRA) module of SAMPSON. The detailed model of the MCRA module was applied to the XR2-1 BWR metallic relocation experiment first. Molten material in the control rod area accumulated on the velocity limiter in the XR2-1 experiment and this phenomenon was reproduced by the SAMPSON analysis. A part of the molten metal fell directly through the inlet orifice in both the XR2-1 experiment and the SAMPSON analysis. Then the detailed model of the MCRA module was applied to the relocation phenomena of actual fuel bundles. The molten material accumulation on the velocity limiter and direct falling of the molten material through the inlet orifice were also observed in the analysis of actual fuel bundles. Based on the observations described above, MCRA noding for the system calculation was modified as follows. (1) The velocity limiters and control guide tubes were newly taken into account. (2) The flow path of debris was modified so that the molten materials could go to the lower plenum after passing through the inlet orifice without forced accumulation at the core plate. (author)

  17. Application of Electronic Business in Safe Accident Prevention and Control on Coalface

    Lu, Guozhi; Tang, Jianquan; Yao, Chunhui; Yang, Lei

    In this paper, by analyzing the coal mine safety accident of present stage, the author has come to a conclusion that the safe accidents on coalface accounting for a lot of coal mine safety accident, and has brought forward the cause leading to this phenomenon. Then, through the discussion about "Overlying Strata Movement Law", this author has suggested that Electronic Business can be used for the coal mine to prevent and control safe accident on coalface, and has given out the operating pattern of Electronic Business innovatively. This conclusions are most instructive to Chinese coal mine in managing safe accident on coalface and innovative for application of Electronic Business in coal mine safety.

  18. Simulation Code Development and Its Applications

    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.

  19. Application of coupled code technique to a safety analysis of a standard MTR research reactor

    Accident analyses in nuclear research reactors have been performed, up to now, using simple computational tools based on conservative physical models. These codes, developed to focus on specific phenomena in the reactor, were widely used for licensing purposes. Nowadays, the advances in computer technology make it possible to switch to a new generation of computational tools that provides more realistic description of the phenomena occurring in a nuclear research reactor. Recent International Atomic Energy Agency (IAEA) activities have emphasized the maturity in using Best Estimate (BE) Codes in the analysis of accidents in research reactors. Indeed, some assessments have already been performed using BE thermal-hydraulic system codes such as RELAP5/Mod3. The challenge today is oriented to the application of coupled code techniques for research reactors safety analyses. Within the framework of the current study, a Three-Dimensional Neutron Kinetics Thermal-Hydraulic Model (3D-NKTH) based on coupled PARCS and RELAP5/Mod3.3 codes has been developed for the IAEA High Enriched Uranium (HEU) benchmark core. The results of the steady state calculations are sketched by comparison to tabulated results issued from the IAEA TECDOC 643. These data were obtained using conventional diffusion codes as well as Monte Carlo codes. On the other hand, the transient analysis was assessed with conventional coupled point kinetics-thermal-hydraulic channel codes such as RELAP5 stand alone, RETRAC-PC, and PARET codes. Through this study, the applicability of the coupled code technique is emphasized with an outline of some remaining challenges.

  20. Application of coupled code technique to a safety analysis of a standard MTR research reactor

    Hamidouche, Tewfik [Division de l' Environnement, de la Surete et des Dechets Radioactifs, Centre de Recherche Nucleaire d' Alger (CRNA), Alger (Algeria); Laboratoire de Mecanique des Fluides Theorique et Appliquee, Faculte de Physique, Universite Des Sciences et de la Technologie Houari Boumediene, (USTHB), Bab-Ezzouar, Alger (Algeria)], E-mail: t.hamidouche@crna.dz; Bousbia-Salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione-Facolta di Ingegneria, Universita di Pisa, Pisa (Italy)], E-mail: b.salah@ing.unipi.it; Si-Ahmed, El Khider [Laboratoire de Mecanique des Fluides Theorique et Appliquee, Faculte de Physique, Universite Des Sciences et de la Technologie Houari Boumediene, (USTHB), Bab-Ezzouar, Alger (Algeria)], E-mail: esi-ahmed@usthb.dz; Mokeddem, Mohamed Yazid [Division de la Physique et des Applications Nucleaires, Centre de Recherche Nucleaire de Draria (CRND) (Algeria); D' Auria, Franscesco [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione-Facolta di Ingegneria, Universita di Pisa, Pisa (Italy)

    2009-10-15

    Accident analyses in nuclear research reactors have been performed, up to now, using simple computational tools based on conservative physical models. These codes, developed to focus on specific phenomena in the reactor, were widely used for licensing purposes. Nowadays, the advances in computer technology make it possible to switch to a new generation of computational tools that provides more realistic description of the phenomena occurring in a nuclear research reactor. Recent International Atomic Energy Agency (IAEA) activities have emphasized the maturity in using Best Estimate (BE) Codes in the analysis of accidents in research reactors. Indeed, some assessments have already been performed using BE thermal-hydraulic system codes such as RELAP5/Mod3. The challenge today is oriented to the application of coupled code techniques for research reactors safety analyses. Within the framework of the current study, a Three-Dimensional Neutron Kinetics Thermal-Hydraulic Model (3D-NKTH) based on coupled PARCS and RELAP5/Mod3.3 codes has been developed for the IAEA High Enriched Uranium (HEU) benchmark core. The results of the steady state calculations are sketched by comparison to tabulated results issued from the IAEA TECDOC 643. These data were obtained using conventional diffusion codes as well as Monte Carlo codes. On the other hand, the transient analysis was assessed with conventional coupled point kinetics-thermal-hydraulic channel codes such as RELAP5 stand alone, RETRAC-PC, and PARET codes. Through this study, the applicability of the coupled code technique is emphasized with an outline of some remaining challenges.

  1. Applications of Derandomization Theory in Coding

    Cheraghchi, Mahdi

    2011-07-01

    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 whether a defective item is present within a specified group of items. We use randomness condensers to explicitly construct optimal, or nearly optimal, group testing schemes for a setting where the query outcomes can be highly unreliable, as well as the threshold model where a query returns positive if the number of defectives pass a certain threshold. Finally, we design ensembles of error-correcting codes that achieve the information-theoretic capacity of a large class of communication channels, and then use the obtained ensembles for construction of explicit capacity achieving codes. [This is a shortened version of the actual abstract in the thesis.

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

    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)

  3. Strict optical orthogonal codes for purely asynchronous code-division multiple-access applications.

    Zhang, J G

    1996-12-10

    Strict optical orthogonal codes are presented for purely asynchronous optical code-division multiple-access (CDMA) applications. The proposed code can strictly guarantee the peaks of its cross-correlation functions and the sidelobes of any of its autocorrelation functions to have a value of 1 in purely asynchronous data communications. The basic theory of the proposed codes is given. An experiment on optical CDMA systems is also demonstrated to verify the characteristics of the proposed code. PMID:21151299

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

    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

  5. Two Applications of the Hamming-Golay Code

    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.

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

    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)

  7. Applicability of coupled code RELAP5/GOTHIC to NPP Krsko MSLB calculation

    Usual way to analyze Main Steam Line Break (MSLB) accident in PWR plants is to calculate core and containment responses in two separate calculations. In first calculation system code is used to address behaviour of nuclear steam supply system and containment is modelled mainly as a boundary condition. In second calculation mass and energy release data are used to perform containment analysis. Coupled code R5G realized by direct explicit coupling of system code RELAP5/MOD3.3 and containment code GOTHIC is able to perform both calculations simultaneously. In this paper R5G is applied to calculation of MSLB accident in large dry containment of NPP Krsko. Standard separate calculation is performed first and then both core and containment responses are compared against corresponding coupled code results. Two versions of GOTHIC code are used, one old ver 3.4e and the last one ver 7.2. As expected, differences between standard procedure and coupled calculations are small. The performed analyses showed that classical uncoupled approach is applicable in case of large dry containment calculation, but that new approach can bring some additional insight in understanding of the transient and that can be used as simple and reliable procedure in performing MSLB calculation without any significant calculation overhead. (author)

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

    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)

  9. Performance and scenario evaluation of PAFS through the LOFW accident in APR1400 by using MARS code

    In order to enhance the safety feature of the APR1400 through the passive ways, the passive auxiliary feedwater system(PAFS) is under preliminary consideration by KAERI. For the successful adaptation of PAFS, accident scenario evaluation of PWR plant that is assumed to have the PAFS system should be performed. Condensing heat exchanger assemblies are installed at the exterior boundary of the containment building per one steam generator. The performance of the heat exchanger is designed to remove the decay heat of the fuel completely. In normal operation condition, PAFS system is not connected with the steam and feed lines. A Total Loss of Feed Water(TLOFW) accident is selected for the performance and scenario evaluation after the severity check. The PAFS connection valves are open at the signal of 25% level trip of steam generator. With the single failure assumption of PAFS open valve, the scenario propagations are calculated by using MARS code

  10. Performance and scenario evaluation of PAFS through the LOFW accident in APR1400 by using MARS code

    Bae, Sung Won; Bae, Byoung Uhn; Yun, Byong Jo [Korea Atomic Energy Institute, Daejeon (Korea, Republic of)

    2009-07-01

    In order to enhance the safety feature of the APR1400 through the passive ways, the passive auxiliary feedwater system(PAFS) is under preliminary consideration by KAERI. For the successful adaptation of PAFS, accident scenario evaluation of PWR plant that is assumed to have the PAFS system should be performed. Condensing heat exchanger assemblies are installed at the exterior boundary of the containment building per one steam generator. The performance of the heat exchanger is designed to remove the decay heat of the fuel completely. In normal operation condition, PAFS system is not connected with the steam and feed lines. A Total Loss of Feed Water(TLOFW) accident is selected for the performance and scenario evaluation after the severity check. The PAFS connection valves are open at the signal of 25% level trip of steam generator. With the single failure assumption of PAFS open valve, the scenario propagations are calculated by using MARS code.

  11. Application of Westinghouse NEXUS/ANC9 cross-section model for PWR accident analyses

    NEXUS/ANC9 is the latest licensed PWR core design code system developed by Westinghouse. This system has demonstrated capabilities of modeling advanced core designs with improved accuracy in core reactivity and power distribution predictions. NEXUS/ANC9 system is being rolled out to replace the current APA system (ALPHA/PHOENIX-P/ANC) for routine core calculations. In addition to the standard core design calculations, investigations are underway to explore the possibility to expand the NEXUS/ANC9 application for safety analysis, especially at accident conditions. The main focus of the investigation is the evaluation of the NEXUS/ANC9 cross-section representation model conditions like high void and significant change of core pressure. Comparisons of the predicted parameters among ANC9, PARAGON lattice code and MCNP calculations are presented. The results show that NEXUS/ANC9 is able to model the cross-section behavior and accurately reproduce lattice code results at all simulated conditions. (author)

  12. Applications of bar code technology at nuclear power plants

    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

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

    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

  14. Codes for 3-dimensional thermohydraulic calculation of fast reactor core in steady state, transient and accident conditions

    For the analysis of transient and emergency processes during reactor operation it is necessary to have a set of codes, which calculate physical processes with a various degree of accuracy. Codes CORT and BUMT for three-dimensional thermohydraulic calculation of fast reactor core in steady state, transient and accident conditions are described in this paper. The code CORT calculates thermohydraulics of the whole fast reactor core or group of subassemblies in simplified approximation. The core is described as a set of coupled one-dimensional channels or is divided into a set of ring zones, each of those is also represented by one subassembly (S/A). The detailed three-dimensional calculation of particular S/A is carried out by code BUMT. For description of S/A thermohydraulics the authors have chosen so called 'subchannel model. In this model the S/A is split into number of channels exchanging one by one with mass, momentum and energy. The coefficients of inter channel exchange are calculated on the basis of empirical correlations. The subchannel model is supplemented by detailed (two-dimensional in each axial cross-section) calculation of fuel pin and S/A wrapper temperatures. For solution of hydrodynamic equations the full-implicit scheme is used. Code BUMT was verified using experimental data for S/A-simulators and results of calculations obtained by other codes. These codes when used in complex with neutronic code and first circuit thermohydraulic code could describe in detail the thermal state of coolant and performance of fuel pins and construction elements of reactor during steady and transient states of its operation. (author)

  15. Applications of Derandomization Theory in Coding

    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...

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

    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)

  17. 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

    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)

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

    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

  19. Development and application of the waste code

    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

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

    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.

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

    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

  2. Lessons learnt from the EC/USNRC expert judgement study on probabilistic accident consequence codes applied in the COSYMA uncertainty analyses

    Two probabilistic accident consequence codes, COSYMA and MACCS respectively, estimate the risks and other endpoints associated with hypothetical accidents from nuclear installations. A joint EC/USNRC project for an uncertainty analysis of these two codes was initiated to systematically derive credible and traceable probability distributions for the respective code input variables. A formal expert judgement elicitation and evaluation process was used as the best available technique to accomplish that objective. These input distributions were used in an uncertainty analysis of the COSYMA package. This paper will show the overall process and highlights the lessons learnt from the projects. (author)

  3. Application of RUNTA code in flood analyses

    Flood probability analyses carried out to date indicate the need to evaluate a large number of flood scenarios. This necessity is due to a variety of reasons, the most important of which include: - Large number of potential flood sources - Wide variety of characteristics of flood sources - Large possibility of flood-affected areas becoming inter linked, depending on the location of the potential flood sources - Diversity of flood flows from one flood source, depending on the size of the rupture and mode of operation - Isolation times applicable - Uncertainties in respect of the structural resistance of doors, penetration seals and floors - Applicable degrees of obstruction of floor drainage system Consequently, a tool which carries out the large number of calculations usually required in flood analyses, with speed and flexibility, is considered necessary. The RUNTA Code enables the range of possible scenarios to be calculated numerically, in accordance with all those parameters which, as a result of previous flood analyses, it is necessary to take into account in order to cover all the possible floods associated with each flood area

  4. Accident analysis of flow blockage to coolant channels of upgraded JRR-3, using EUREKA-2 code, (1)

    This report describes the results about thermo-hydraulic behavior in the accident of flow blockage to coolant channels of upgraded JRR-3. Analysis was carried out using EUREKA-2 code. Flow blockage to coolant channels accident occur by some extraneous things which come from outside of the reactor pool, may block the coolant flow channels of the core. If flow blockage to coolant channels would occur, fuel temperature will increase due to flow rate decrease of coolant channels. And at last, fission products will be released from inside of fuel plates to the primary cooling system due to failure of fuel plates. In the analysis, one standard type fuel element was supposed as flow blockage channels, in the same way sa one of credible accidents, which postulated in the JRR-3 safety assessment. From the results, it was shown that about 16.7 % of the fuel element which was supposed as flow blockage channels, would fail, assuming that fuel plates might fail when the fuel meat temperatures riseover 400 deg C. (author)

  5. Recent revisions to MAAP4 for U.S. EPR severe accident applications

    A revision of the MAAP4 code (i.e., version 4.0.7) has been developed to address the severe accident evaluation needs of the U.S. EPR. The U.S. EPR design employs an ex-vessel severe accident strategy involving specific containment regions devoted to debris stabilization and long term cooling. The modifications performed to the MAAP4 code address both the phenomenological aspects and the spatial modeling flexibility consistent with MAAP4's Generalized Containment Model framework. In addition, enhancements have been included in MAAP4 to improve the modeling of other severe accident mitigation features and to improve code usage and level of detail available to the user. The implementation of several new models and enhancements into the MAAP4 code provides the necessary integral analysis capability of U.S. EPR severe accidents from the initiating event, through reactor vessel failure, to long-term containment and ex-vessel melt stabilization. This paper presents a detailed description of the code enhancements supporting the U.S. EPR design certification. (authors)

  6. Application of Coating Technology for Accident Tolerant Fuel Cladding

    To commercialize the ATF cladding concepts, various factors are considered, such as safety under normal and accident conditions, economy for the fuel cycle, and developing development challenges, and schedule. From the proposed concepts, it is known that the cladding coating, FeCrAl alloy, and Zr-Mo claddings are considered as a near/mid-term application, whereas the SiC material is considered as a long-term application. Among them, the benefit of cladding coating on Zr-based alloys is the fuel cycle economy regarding the manufacturing, neutron cross section, and high tritium permeation characteristics. However, the challenge of cladding coating on Zr-based alloys is the lower oxidation resistance and mechanical strength at high-temperature than other concepts. Another important point is the adhesion property between the Zr-based alloy and coating materials. As an improved coating technology compared to a previous study, a 3D laser coating technology supplied with Cr powders is considered to make a coated cladding because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. We are systematically studying the laser beam power, inert gas flow, cooling of the cladding tube, and powder control as key points to develop 3D laser coating technology. After Cr-coating on the Zr-based cladding, ring compression and ring tensile tests were performed to evaluate the adhesion property between a coated layer and Zr-based alloy tube at room temperature (RT), and a high-temperature oxidation test was conducted to evaluate the oxidation behavior at 1200 .deg. C of the coated tube samples. A 3D laser coating method supplied with Cr powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a Cr-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured

  7. Post-test analysis of two accident management experiments performed at the BETHSY test facility using the code ATHLET

    In the framework of the external validation of the thermal-hydraulic code ATHLET, which has been developed by the GRS, post test analyses of two experiments were done, which were performed at the french integral test facility BETHSY. During the experiment 5.2 C the complete loss of steam generator feedwater was simulated. The de-pressurization of the primary circuit and high pressure injection is assumed as an emergency measure. During the experiment 9.3 the break of a steam generator U-tube is simulated. The failure of the high pressure injection is assumed. As accident management measures, the depressurization of the steam generator secondary sides and finally of the primary circuit by opening of the pressurizer valve were investigated. The results show, that the code ATHLET is able to describe the complex scenario in good accordance with the experiment. For both tests the safety related statement could be reproduced. (author)

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

    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)

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

    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.

  10. Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

    Pavlovitchev, A.M.

    2000-03-08

    The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes.

  11. Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

    The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes

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

    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)

  13. Evaluation on operation of liquid relief valves for steam line break accidents by RELAP5/CANDU+ code

    A development of RELAP5/CANDU+ code for regulatory audits of accident analysis of CANDU nuclear power plants is on progress. This paper is undertaken in a procedure of a verification and validation for RELAP5/CANDU+ code by analyzing main steam line break accidents of WS 2/3/4. Following the ECC injection in sequence of the steam line breaks, the mismatch in heat transfer between the primary and the secondary systems makes pressure of the primary system instantly peaked to the open setpoint of liquid relief valves. The event sequence follows the result of WS 2/3/4 FSAR, but there is a difference in pressure transient after ECC injection. Sensitivity analysis for main factors dependent on the peak pressure such as control logics of liquid relief valves. ECC flow path and feedwater flow is performed. Because the pressure increase is continued for a long time and its peaking is high, open and close of the liquid relief valves are repeated several times, which is obviously different from those of WS 2/3/4 FSAR. As a result, it is evaluated that conservative modeling for the above variables is required in the analysis

  14. Quality assurance procedures for the CONTAIN severe reactor accident computer code

    The CONTAIN quality assurance program follows a strict set of procedures designed to ensure the integrity of the code, to avoid errors in the code, and to prolong the life of the code. The code itself is maintained under a code-configuration control system that provides a historical record of changes. All changes are incorporated using an update processor that allows separate identification of improvements made to each successive code version. Code modifications and improvements are formally reviewed and checked. An exhaustive, multilevel test program validates the theory and implementation of all codes changes through assessment calculations that compare the code-predicted results to standard handbooks of idealized test cases. A document trail and archive establish the problems solved by the software, the verification and validation of the software, software changes and subsequent reverification and revalidation, and the tracking of software problems and actions taken to resolve those problems. This document describes in detail the CONTAIN quality assurance procedures. 4 refs., 21 figs., 4 tabs

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

    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.

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

    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

  17. Learning Vector Coding Methods of ART1 and Their Applications

    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.

  18. Uncertainty analysis for control rod ejection accidents simulated by KIKO3D/TRABCO code system

    Recently, considerable conservatism must be applied in the traditional safety analyses for taking into account the uncertainties originating from the input parameters, approximations in the models, due to the safety reserves, etc. The extreme values for all of the input parameters are supposed in the traditional safety analysis at the same time. Additionally it must be mentioned that the selection of the input parameter values leading to conservative results often is not easy. The main goal of this paper is to present a more realistic methodology for the case of control rod ejection accidents. The applied consistent statistical approach leads to conservative results also, but avoids the unnecessary cumulative conservatism. A method based on a mathematical model ('Two-Sided Statistical Tolerance Intervals', [1-2]) was chosen for the realization of uncertainty analyses of Reactivity Initiated Accidents (RIA). (author)

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

    MABEL-2 has been developed to predict the extent of cladding deformation in PWR fuel rods during a loss of coolant accident. The user notes describe how to run MABEL. They include case preparation and input data, the job control language, a description of the output tables available, and aids to debugging. The input data and results for two sample cases are given. (U.K.)

  20. Development of a computer code concerning the diffusion of radioactive effluents and radiological exposure following an accident

    Cheap and clean energy can be supplied with nuclear power plant whose accident probability is very low comparing with that of the other industrial facilities. However, the consequences of severe accident of nuclear power plant may result in a critical impact on population and ecosystem over a wide area due to the radioactive effluents released into the atmosphere as was in the case of the Chernobyl accident. Therefore, it is necessary to prepare an effective computer analysis system for real-time dose assessment against a potential severe accident of nuclear facility. The objective of this study is a development of the real-time assessment system of environmental exposure dose in an emergency and the installation of its prototype system. Wind field model satisfying the mass conservation and Monte Carlo diffusion model to save computing time were developed in order to assess real-time diffusion of radioactive effluents in an accident. Comparing studies were done in 3 kinds of typical topographic situation dependent on every atmospheric stability. Data library of exposure doses was documented by several computer calculation so that computation time could be reduced considerably. These results of the computation with data library showed a good agreement with the real calculation by the existing mode in which numerical integration model was used. And a graphic package was established that enabled to display the wind vector, concentration field of radioactives, and distribution of exposure dose every time step on the color graphic terminal. Work station was composed of host computer CDC-CYBER, graphic application processor - TEK 4301, graphic CRT - TEK 4125 and hard copier - TEK 4693 D. (Author)

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

    MABEL-2D is a code to predict the extent of cladding deformation in PWR fuel rods during a LOCA. An earlier version of the code, MABEL-2C, was issued in 1982, but this has been extensively modified. The modifications to convert MABEL-2C to MABEL-2D are described. (author)

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

    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)

  3. Preliminary assessment of the accident probabilities of a NERVA derivative reactor for space power applications

    Gas cooled, solid core graphite reactors based on the Nuclear Engine Rocket Vehicle Application (NERVA) technology are attractive space nuclear power sources for SDI multimegawatt applications. A reactor of this type is called a NERVA Derivative Reactor (NDR). A simplified accident probability analysis of a NDR reactor for space power mission was carried out. The objective of the analysis was to identify the worst case accident and critical reactor design areas where the incorporation of appropriate design features may reduce the risks involved

  4. Applications of Coding in Network Communications

    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…

  5. 'Turbo' coding for deep space applications

    Andersen, Jakob Dahl

    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...

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

    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)

  7. Physical sputtering code for fusion applications

    A computer code, DSPUT, has been developed to compute the physical sputtering yields for various plasma particles incident on candidate fusion-reactor first-wall materials. The code, which incorporates the energy and angular-dependence of the sputtering yield, treats both high- and low-Z incident particles bombarding high- and low-Z wall materials. The physical sputtering yield is expressed in terms of the atomic and mass numbers of the incident and target atoms, the surface binding energy of the wall materials, and the incident angle and energy of the particle. An auxiliary code has been written to provide sputtering yields for a Maxwellian-averaged incident particle flux. The code DSPUT has been used as part of a Monte Carlo code for analyzing plasma-wall interactions

  8. Bases of general calculation thermohydrodynamic means (codes) verification and validation methodology for accident analysis at nuclear power plants

    On the basis of previous known approaches' analysis the generalised calculation thermohydrodynamic means verification/validation (V/V) methodology for accident/transition processes' analysis at NPPs is offered in this article. Taking into account formulated requirements and principles the basic V/V procedures, their correlation and order are grounded and considered. The realisation order includes forming calculation means applicability assessment criteria system, analysing mathematical models adequacy to real processes, developing test data bases including a stands adequacy analysis to full-scale conditions, results generalisation methods for final calculation means applicability assessments for specific tasks at specific equipment

  9. Consistent Comparison of Full Core PWR Reactivity Initiated Accident with the Method Of Characteristic Code DeCART and the Coarse Mesh Nodal Code PARCS - 180

    The current state of the art in analysis of a control rod ejection event in a Pressurized Water Reactor (PWR) relies upon the assembly averaged power from a whole core nodal neutronics simulator and some type of pin power reconstruction within the fuel assembly. Recently, there has been interest in taking advantage of the DeCART code to perform a higher fidelity solution which could lead to more accurate pin-power results as well as provide intra-pin power information during the transient. The work described in this paper is the comparison of PARCS and DeCART analysis of two Reactivity Initiated Accidents. The methods used in PARCS and DeCART are briefly described as well as the approach to generate the needed temperature feedbacks. The generation of the macroscopic cross sections and kinetic parameters for PARCS is detailed. The results of both scenarios are shown and the main differences of both approaches are discussed. (authors)

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

    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

  11. Generalized polyphase representation and application to coding gain enhancement

    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.

  12. The coupled code TRAB-3D-SMABRE for 3D transient and accident analyses

    The three-dimensional TRAB-3D core dynamics code is being internally coupled to the thermal hydraulics system code SMABRE. The codes have previously been coupled with a parallel coupling scheme. VTT's reactor dynamics codes have performed well in all the situations that they have originally been designed for. The most important limitation of the present code models is their inability to handle coolant flow reversal in the core channel, a phenomenon that can be encountered in e.g. BWR ATWS cases or VVER power excursions. The new coupling of the two codes is realized on the level of each node of each channel in the core, with each fuel bundle described with its own channel. Necessary interfaces have been created, an improved version of SMABRE's thermal hydraulics solution method developed, and a steady state procedure developed. A satisfactorily working steady state solution has been achieved. The next step in the development will be testing of the transient calculation. Besides solving the flow reversal limitation of the present dynamics models, a successful coupling will allow expanding into more realistic modelling of an open core. (orig.)

  13. SLAC Parallel Tracking Code Development and Applications

    The increase in single processor speed based on Moore's law alone will not be able to deliver the dramatic speedup needed in many beam tracking simulations to uncover very slowly evolving effects in a reasonable time. SLAC has embarked on an effort to bring the power of parallel computing to bear on such computations with the goal to reduce the turnaround time by orders of magnitude so that the results may impact present facilities and future machine designs. This poster will describe the approaches adopted for parallelizing the LIAR code and the IONMAD code. The scalability of these tracking codes and their further improvement will be discussed

  14. From Barcode to QR Code Applications

    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.

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

    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)

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

    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.

  17. Development of debris coolability analysis module in severe accident analysis code SAMPSON for IMPACT project

    Debris coolability in the lower plenum of the reactor pressure vessel is an important factor for the evaluation of in-vessel debris retention. The debris coolability analysis module has been developed to predict more mechanistically the safety margin of the present reactor vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris spreading is solved by the explicit method on a quasi-three-dimensional scheme and debris coolability is solved on the basis of natural convection analysis with melting and solidification. The calculated results for spreading were compared with the results from a water spreading experiment on the floor and the results for coolability were compared with those from an n-octadecane melting experiment in the rectangular vessel. The comparisons showed the capability for predictions of the spearhead transportation in the debris spreading process and of the melting front transportation and time evolution of the fluid temperature in the melting process. The module provides a good tool for the prediction of the reactor pressure vessel safety margin in a severe accident through the analysis of debris spreading and coolability. (author)

  18. Development of debris coolability analysis module in severe accident analysis code SAMPSON for IMPACT project

    Ujita, Hiroshi [Advanced Simulation Systems Department, Nuclear Power Engineering Cooperation, Tokyo (Japan); Hidaka, Masataka; Susuki, Akira; Ishida, Naoyuki

    1999-10-01

    Debris coolability in the lower plenum of the reactor pressure vessel is an important factor for the evaluation of in-vessel debris retention. The debris coolability analysis module has been developed to predict more mechanistically the safety margin of the present reactor vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris spreading is solved by the explicit method on a quasi-three-dimensional scheme and debris coolability is solved on the basis of natural convection analysis with melting and solidification. The calculated results for spreading were compared with the results from a water spreading experiment on the floor and the results for coolability were compared with those from an n-octadecane melting experiment in the rectangular vessel. The comparisons showed the capability for predictions of the spearhead transportation in the debris spreading process and of the melting front transportation and time evolution of the fluid temperature in the melting process. The module provides a good tool for the prediction of the reactor pressure vessel safety margin in a severe accident through the analysis of debris spreading and coolability. (author)

  19. Validation of RALOC4 code for Ignalina NPP Accident Localisation System employing parameters measured during MSV opening

    Accident Localisation System (ALS) of Ignalina NPP is a pressure suppression type confinement. It consists of a number of interconnected compartments with 10 condensing pools to condense the accident-generated steam and to reduce the peak pressures that can be reached during any LOCA. The condensing pools are located at five elevations in two ALS towers. In the case of main safety valve (MSV) opening the released steam is directed to the top (5th) condensing pool of ALS. The ALS thermal hydraulic parameters measured during unintended opening of single MSV which appeared on November 8, 1998 at Ignalina NPP Unit 2 were used for validation of RALOC4 code (Germany). Post-event calculations performed and the calculated water temperatures and water levels in condensing pools as well as condenser tray cooling system (CTCS) parameters were compared with corresponding measured data. The results of the performed sensitivity analysis showed that in the best-estimate analysis the heat transfer coefficient in CTCS heat exchangers could be increased to 2500 W/(m2.K) compared to conservative value of 1000 W/(m2.K) applied in former calculations.(author)

  20. Severe accident analyses of a BWR with MAAP5 code. Station blackout and large-break LOCA

    Calculations were performed for a station blackout (TBU) sequence and a large-break loss-of-coolant accident (AE) sequence of a typical BWR-5 plant with modified Mark-II type containment by the MAAP5 code. The core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment including hydrogen production were investigated. Sensitivity analyses focusing on direct contact heating (DCH) and zirconium oxidation, which affect on the consequences of severe accidents, were also performed. If extensive DCH does not occur in the TBU sequence, failure of the containment vessel can be postponed. On the other hand, concrete ablation at a floor and a side wall in the pedestal due to molten core - concrete interaction (MCCI) significantly increases, because a large amount of debris with high temperature stays inside the pedestal. Although the hydrogen production is affected by the zirconium oxidation model, the differences of hydrogen production are within ± 10% in the case of TBU sequence. (author)

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

    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. Modelling of fission product release from TRISO fuel during accident conditions : benchmark code comparison / Ramlakan A.

    Ramlakan, Alastair Justin

    2011-01-01

    This document gives an overview of the proposed MSc study. The main goal of the study is to model the cases listed in the code benchmark study of the International Atomic Energy Agency CRP–6 fuel performance study (Verfondern & Lee, 2005). The platform that will be employed is the GETTER code (Keshaw & van der Merwe, 2006). GETTER was used at PBMR for the release calculations of metallic and some non–metallic long–lived fission products. GETTER calculates the transport of fission products ...

  3. Development and applicability analysis of ATHLET-SC code

    Research activities of supercritical water reactor (SCWR) have been carried out worldwide,aiming at cost reduction by system simplification and higher thermal efficiency. One of the most important tasks for the design and assessment of SCWR performance is to develop system analysis codes which are applicable under supercritical conditions. The paper presents the development of new system analysis code ATHLET-SC based on ATHLET 2.1A. Thermo-physical properties package valid for supercritical conditions are implemented into the existing ATHLET code to extend its application to safety analysis of SCWR. In to evaluate the applicability of the modified code, a core calculation model of mixed SCWR(SCWR-M) was proposed and analyzed, and transients of core power were simulated. Moreover, a simplified supercritical water cooled loop was proposed and its stability behaviors were analyzed. The results achieved up to now indicate a good applicability of the modified ATHLET code (ATHLET-SC) in simulation of SCWR. (authors)

  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

    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. Bar code application to nuclear material accountancy

    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

  6. Automatic Code Generation for Recurring Code Patterns in Web Based Applications and Increasing Efficiency of Data Access Code

    Senthil, J; Arumugam, S.; S Margret Anouncia; Abhinav Kapoor

    2012-01-01

    Today, a lot of web applications and web sites are data driven. These web applications have all the static and dynamic data stored in relational databases. The aim of this thesis is to generate automatic code for data access located in relational databases in minimum time.

  7. Accident probabilities of a NERVA derivative reactor for space power applications

    This paper presents a preliminary assessment of the accident probabilities of a NERVA Derivative Reactor used for multimegawatt space power applications. The analysis was based on a reactor design capable of passive reentry breakup and fuel dispersal in the upper atmosphere. The worst-case accident inferred was intact reentry and intact impact, followed by ejection of the control rods from the core and either water immersion or core compaction

  8. The 3D core thermohydraulics and neutronics solution in the TRAB-SMABRE accident and transient code

    The TRAB-SMABRE code is a result of code development efforts carried out at VTT Processes in Finland for calculating transient and accident behaviour in Finnish LWR plants. The operating plants are two 770 MWe BWR units in Olkiluoto and two 500 MWe PWR units of VVER-440 type in Loviisa. In addition a new PWR plant of the 1600 MWe EPR type will be built into Olkiluoto. The TRAB core model, two group neutronics solution with nodal expansion method, has been initially developed as a transient 3D transient code for the BWR plant transients, and HEXTRAN code for the 3D VVER transients having the hexagonal fuel geometry. The SMABRE model thermohydraulic model is a drift flux based LOCA model. HEXTRAN and SMABRE were coupled in parallel for making the ATWS analyses in VVER plants possible and TRAB and SMABRE were coupled in parallel for calculating the transients in the PWR plants with the squared core array. As the new step the TRAB core model was coupled internally with the SMABRE for making possible the BWR analyses with the flow reversal possible and as an optional tool for the PWR plant analyses with the squared core array. The model predicts well the 3D core thermohydraulics with the encapsulated fuel, but not for the open PWR core. To overcome this deficiency a thermohydraulics simulation model for the core was introduced based on the 3D porous media thermohydraulics solution PORFLO. In the paper the basic equations of the 3D neutronics, SMABRE thermohydraulics and PORFLO thermohydraulics are described. For the BWR plant the calculation results using the parallel coupling of neutronics and thermohydraulics and internal coupling will be compared for the two transients, MSIV closure in the steam line and partial load reduction, both compared against the real plant data. The calculation result proves that the internal coupling gives the most extensive possibilities for the core simulation and is recommended for the further BWR analyses for the PWR plant the control

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

    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.)

  10. BAR-MOM Code and Its Application

    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

  11. Description of THALES-CV1: a computer code for evaluating containment temperature and pressure during core meltdown accident

    THALES-CV1, a computer code for evaluating reactor containment pressure and temperature during core meltdown accident, was developed. In the code a whole free volume inside the containment is divided into several number of compartments which are connected to each other by junctions. Each compartment is further divided into two regions - gas region and liquid region including water droplets and steam bubbles respectively - by a movable mixture surface. THALES-CV1 evaluates mass and heat flows to/from the regions and calculates containment pressure and mass and temperature in each region, assuming uniform pressure in the containment and thermal equilibrium in each region of a compartment. A junction flow is calculated so that the volume of every compartment is kept constant and the phase of the flow is determined from the relationship between the mixture level and the junction elevation in the upstream compartment. In this code the following mass and heat transfers are considered: Blowdown flow from reactor cooling system, water boiling in reactor cavity, gas generation by melt-concrete interaction, heat transfer and mass addition by containment spray, ECCS flow addition to containment after reactor pressure vessel melt-through, cooling by containment air cooler, ventilation from one compartment to another, heat transfer to containment wall and internal structures, composition change and heat addition due to hydrogen burning, cooling by ice condenser, and leakage from containment to outside. All the models for these mass and heat transfers are very simple. If the sensitivity study for the code which is going to be performed hereafter finds some models are sensitive for containment pressure and temperature transient, such models will be sophisticated. (author)

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

    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

  13. Linking of FRAP-T, FRAPCON and RELAP-4 codes for transient analysis and accidents of light water reactors fuel rods

    The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)

  14. 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

    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.)

  15. Validation Calculations for the Application of MARS Code to the Safety Analysis of Research Reactors

    In order to investigate the applicability of MARS code to the accident analysis of the HANARO and other RRs, the following test data were simulated. Test data of the HANARO design and operation, Test data of flow instability and void fraction from published documents, IAEA RR transient data in TECDOC-643, Brazilian IEA-R1 experimental data. For the simulation of the HANARO data with finned rod type fuels at low pressure and low temperature conditions, MARS code, developed for the transient analysis of power reactors, was modified. Its prediction capability was assessed against the experimental data for the HANARO. From the assessment results, it can be said that the modified MARS code could be used for analyzing the thermal hydraulic transient of the HANARO. Some other simulations such as flow instability test and reactor transients were also done for the application of MARS code to RRs with plate type fuels. In the simulation for these cases, no modification was made. The results of simulated cases show that the MARS code can be used to the transient analysis of RRs with careful considerations. In particular, it seems that an improvement on a void model may be necessary for dealing with the phenomena in high void conditions

  16. Validation Calculations for the Application of MARS Code to the Safety Analysis of Research Reactors

    Park, Cheol; Kim, H.; Chae, H. T.; Lim, I. C

    2006-10-15

    In order to investigate the applicability of MARS code to the accident analysis of the HANARO and other RRs, the following test data were simulated. Test data of the HANARO design and operation, Test data of flow instability and void fraction from published documents, IAEA RR transient data in TECDOC-643, Brazilian IEA-R1 experimental data. For the simulation of the HANARO data with finned rod type fuels at low pressure and low temperature conditions, MARS code, developed for the transient analysis of power reactors, was modified. Its prediction capability was assessed against the experimental data for the HANARO. From the assessment results, it can be said that the modified MARS code could be used for analyzing the thermal hydraulic transient of the HANARO. Some other simulations such as flow instability test and reactor transients were also done for the application of MARS code to RRs with plate type fuels. In the simulation for these cases, no modification was made. The results of simulated cases show that the MARS code can be used to the transient analysis of RRs with careful considerations. In particular, it seems that an improvement on a void model may be necessary for dealing with the phenomena in high void conditions.

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

    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.

  18. Calculation of accidents on the +14.7 m level at the Dukovany WWER-440 NPP, using code MELCOR 1.8.2

    Computer calculations are reported of tube rupture accidents in the steam pipeline room at the 14.7 m level in the Dukovany NPP (reactor WWER-440/213). The MELCOR 1.8.2 code and a Hewlett-Packard 9000/720 computer were used. As input to MELCOR, the input data used in the Severe Accident Simulator for the Jaslovske Bohunice NPP were used with minor modifications. The calculations were aimed at determining the pressure, temperature, humidity and condensed water level in the room after the following accidents: main steam header rupture; steam line rupture before fast acting valve; main feedwater header rupture. The results are presented in numerous graphs in 3 appendices. The calculations showed the temperature and humidity in the steam pipeline room to be the limiting factors for the equipment important to reactor safety. Pressure will increase only a little after the accident and water on the floor will not reach a critical level. 15 refs

  19. Qualification of the ARROTTA code for light water reactor accident analysis

    Qualification efforts have been performed by the Taiwan Power Company (TPC) and the Institute of Nuclear Energy Research (INER) for the three-dimensional spatial kinetics code ARROTTA for light water reactor (LWR) core transient analysis. Together TPC and INER started a 5-yr project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multidimensional kinetics calculations such as rod ejection for pressurized water reactors and rod drop for boiling water reactors (BWR). To qualify ARROTTA for evaluation of the Final Safety Analysis Report licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicate that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include the Nuclear Energy Agency Committee on Reactor Physics rod ejection problem, the three-dimensional Langenbuch-Maurer-Werner LWR rod withdrawal/insertion problem, and the three-dimensional linear regression analysis BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared with other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multidimensional core transients for LWRs

  20. The transplantation, augmentation and application of ORIGEN-2 code

    ORIGEN-2 is a versatile point depletion and decay computer code. Its functions, major characteristics, and theoretical models are briefly introduced. The transplantation of the code on IBM 4300 series computers and its augmentation of output photon data and application in calculating gamma-ray spectra of fission products are described

  1. Fuel thermal/mechanical behaviour under loss of coolant accident conditions as predicted by the FACTAR code

    FACTAR (Fuel And Channel Temperature And Response) is a computer code developed to simulate the transient thennal and mechanical behaviour of 37-element or 28-element fuel bundles within a single CANDU fuel channel for moderate (ie., sheath temperatures less than the melting point of Zircaloy) loss-of-coolant accident (LOCA) conditions including transition and large break LOCAs with emergency coolant injection assumed available. FACTAR's predictions of fuel temperature and sheath failure times are used for subsequent assessment of fission product releases and fuel string expansion. In this paper, model capabilities and calculated quantities of the code are summarised. The results from overly severe test cases are presented in order to clearly demonstrate the effect on calculated fuel channel behaviour of a mechanistic assessment of fuel-to-sheath heat transfer, and the impact of using a diffusion-limited model for Zircaloy/steam reaction (i.e., FROM) as opposed to a reaction rate correlation, coupled with the assumption of unlimited steam supply. (author)

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

    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)

  3. Applications of autoassociative neural networks for signal validation in accident management

    The OECD Halden Reactor Project has been working for several years with computer based systems for determination on plant status including early fault detection and signal validation. The method here presented explores the possibility to use a neural network approach to validate important process signals during normal and abnormal plant conditions. In BWR plants, signal validation has two important applications: reliable thermal limits calculation and reliable inputs to other computerized systems that support the operator during accident scenarious. This work shows how a properly trained autoassociative neural network can promptly detect faulty process signal measurements and produce a best estimate of the actual process value. Noise has been artificially added to the input to evaluate the network ability to respond in a very low signal to noise ratio environment. Training and test datasets have been simulated by the real time transient simulator code APROS. Future development addresses the validation of the model through the use of real data from the plant. (author). 5 refs, 17 figs

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

    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

  5. BAR-MOM code and its application

    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

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

    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

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

    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...

  8. Application of thermal-hydraulic codes in the nuclear sector

    Use of thermal-hydraulic codes is extended all over many different aspects of nuclear engineering. This article groups and briefly describes the main features of some of the well known codes as an introduction to their recent applications in the Spain nuclear sector. the broad range and quality of applications highlight the maturity achieved both in industry and research organizations and universities within the Spanish nuclear sector. (Author)

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

    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.

  10. Qualification of the CFD code TRIO-U for full scale nuclear reactor applications

    Numerical and experimental research on nuclear safety is in the end dedicated to understand, on a plant scale, the fundamental physical phenomena which are associated to specific accident scenarios. Hence, the results derived from single effect experiments or reduced scale analysis have to be extrapolated to plant scale whereas plant scale experiments should be evaluated with respect to their applicability to the physics of the specific scenario. For several years, IRSN and CEA have used Computational Fluid Dynamics (CFD) codes for detailed nuclear safety analyses on plant scale. The paper presents a procedure which has been used to qualify the Trio-U code for the prediction of the boron concentration at the core inlet of a French Pressurized Water Reactor (PWR) in accidental conditions (inherent dilution problem) 1. A ROCOM experiment as well as an UPTF Tram-C3 experiment has been used for this purpose. (authors)

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

    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

  12. PERSPECTIVES ON A DOE CONSEQUENCE INPUTS FOR ACCIDENT ANALYSIS APPLICATIONS

    (NOEMAIL), K; Jonathan Lowrie, J; David Thoman (NOEMAIL), D; Austin Keller (NOEMAIL), A

    2008-07-30

    Department of Energy (DOE) accident analysis for establishing the required control sets for nuclear facility safety applies a series of simplifying, reasonably conservative assumptions regarding inputs and methodologies for quantifying dose consequences. Most of the analytical practices are conservative, have a technical basis, and are based on regulatory precedent. However, others are judgmental and based on older understanding of phenomenology. The latter type of practices can be found in modeling hypothetical releases into the atmosphere and the subsequent exposure. Often the judgments applied are not based on current technical understanding but on work that has been superseded. The objective of this paper is to review the technical basis for the major inputs and assumptions in the quantification of consequence estimates supporting DOE accident analysis, and to identify those that could be reassessed in light of current understanding of atmospheric dispersion and radiological exposure. Inputs and assumptions of interest include: Meteorological data basis; Breathing rate; and Inhalation dose conversion factor. A simple dose calculation is provided to show the relative difference achieved by improving the technical bases.

  13. Application of Inertia Ellipse in Code Marker Matching

    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.

  14. From Barcode to QR Code Applications

    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 ...

  15. Rod ejection accident 3D-dynamic analysis in Trillo NPP with RELAP5/PARCS V2.7 coupled codes

    The Rod Ejection Accident (REA) belongs to the Reactivity Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA consist of a rod ejection due to the failure of its driving mechanism. The evolution is driven by a continuous reactivity insertion. In previous works, we have analyzed this transient in Trillo NPP at different power levels at the beginning of cycle (BOC) and at the end of cycle (EOC) using the coupled code RELAP5-MOD3.3/PARCSv2.7. In this work, we present the results of the REA analysis at 30% of the rated power at BOC. In the thermalhydraulic model used, each fuel assemblies has been modelled as an independent channel, for that, the RELAP5 source code has been modified and recompiled to accept this large number of channels. The neutronic nodal discretization consists of 177 x 32 active nodes, considering 28 different fuel elements with 867 neutronic compositions. The cross-sections sets are obtained from CASMO4-SIMULATE3 using the SIMTAB methodology developed in UPV. The transient departs from an initially critical core, being the withdrawal speed of the control rod a typical bounding value. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions, in this way the conclusions will be realistic. The aim is to improve the understanding of these accidents using advanced methods. (authors)

  16. Rod ejection accident 3D-dynamic analysis in Almaraz NPP with RELAP5/PARCS V2.7 coupled codes

    The Rod Ejection Accident (REA) belongs to the Reactivity Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA consist of a rod ejection due to the failure of its driving mechanism. The evolution is driven by a continuous reactivity insertion. In previous works, we have analyzed this transient in Almaraz NPP at different power levels at the beginning of cycle (BOC) and at the end of cycle (EOC) using the coupled code RELAP5-MOD3.3/PARCS v2.7. In this work, we present the results of the REA analysis at hot zero power at BOC with all control rods inserted. In the thermal-hydraulic model used, each fuel assemblies has been modelled as an independent channel, for that, the RELAP5 source code has been modified and recompiled to accept this large number of channels. The neutronic nodal discretization consists of 157 x 24 active nodes, considering 13 different fuel elements with 291 neutronic compositions. The cross-sections sets are obtained from CASMO4-SIMULATE3 using the SIMTAB methodology developed in UPV. The transient departs from an initially critical core, being the withdrawal speed of the control rod a typical bounding value. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions, in this way the conclusions will be realistic. The aim is to improve the understanding of these accidents using advanced methods. (authors)

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

    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). [PWR; BWR

    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.

  19. PASLOCA: two-dimensional code for loss of primary coolant accident analysis in pool type reactors for use in micro computers

    In order to improve the better performance of the MILOCA code, a 2-dimensional code for Loss of Coolant Accident Analysis in pool type research reactor for use in IBM-PC, an adaptation of the code was made from FORTRAN to PASCAL. This paper presents also the heat transfer model calculation from the fuel elements to the air after the draining of the pool water. As an example, the analysis made for the IAEA 2 MW reference reactor is presented. Differences in computing time between the two versions are also shown. (author)

  20. Analysis of transients and accidents with the system code ATHLET for the Krsko Nuclear Power Plant

    Main aspects of the cooperation between the republic of Croatia and the F.R. of Germany in the field of NPP safety research are overviewed in the paper. The GRS system code ATHLET developed for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors is now being available at the University of Zagreb. A very comprehensive ATHLET standard input data set for the NPP Krsko has been established. This data set was validated by calculation of the event 'Main Steam Isolation Valve Closure' that occurred at the PWR NPP Krsko in 1995 and comparing the resulting characteristic parameters with the corresponding measured data. (author)

  1. Aerosol transport analysis of LWR high-consequence accidents using the HAA-4A code

    Use of the HAA-4A code to calculate removal of aerosol in containment due to inherent behavior mechanisms is described. Results for a PWR TMLB' scenario showed a source reduction of about a factor of 50 in CsI available for release to the environment through a catastrophic containment failure. Respirable CsI entering containment from the primary coolant system and melt-through blowdown was a factor of 25 less than the source. The principal removal mechanisms were particle growth due to Brownian and differential settling agglomeration and subsequent fallout. Sensitivities to important and uncertain parameters are discussed. Increased removal due to turbulent agglomeration and a larger expected source particle size are indicated. A seven control volume analysis took less than 1 minute of CPU time on an IBM 3033

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

    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

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

    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

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

    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

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

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

  6. Quantifying reactor safety margins. Pt. 1; An overview of the code scaling, applicability, and uncertainty evaluation methodology

    Boyack, B.E. (Los Alamos National Lab. (LANL), NM (USA). Nuclear Technology and Engineering Div.); Catton, I. (California Univ., Los Angeles (USA)); Duffey, R.B.; Katsma, K.R.; Wilson, G.E. (Idaho National Engineering Lab., Idaho Falls (USA)); Griffith, P. (Massachusetts Inst. of Tech., Cambridge (USA)); Lellouche, G.S.; Levy, S. (Levy (S.), Inc., Campbell, CA (USA)); Rohatgi, U.S.; Wulff, W. (Brookhaven National Lab., Upton, NY (USA)); Zuber, N. (Nuclear Regulatory Commission, Washington, DC (USA))

    1990-05-01

    The CSAU methodology and an example application, described in this set of six papers, demonstrate that uncertainties in complex phenomena can be quantified. The methodology is structured, traceable, and practical, as is needed in the regulatory arena. It addresses in a comprehensive and systematic manner questions concerned with: (1) code capability to scale-up processes from test facility to full-scale nuclear power plant (NPP), (2) code applicability to safety studies of a postulated accident scenario in a specified NPP, and (3) quantifying uncertainties of calculated results. The methodology combines a 'top-down' approach to define the dominant phenomena with a 'bottom-up' approach to quantify uncertainty. The methodology is able to address both: (1) uncertainties for which bias and distribution are quantifiable, and (2) uncertainties for which only a bounding value is quantifiable. The methodology is general, and therefore applicable to a variety of scenarios, plants, and codes. This paper provides an overview of the CSAU evaluation methodology and its application to a postulated cold-leg, large-break loss-of-coolant accident in a Westinghouse four-loop pressurized water reactor with 17x17 fuel. The code selected for this demonstration of the CSAU methodology was TRAC-PF1/MOD1, Version 14.3. (orig./GL).

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

    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

  8. Accident consequence calculations and risk assessments for pressurized light water reactors with the computer code UFOMOD/B3

    With respect to the application of the accident consequence model of the German Risk Study (GRS) for light water reactors to risk assessments of other reactor types (high temperature reactor HTR-1160, fast breeder reactor SNR-300), the improved version UFOMOD/B3 was developed. The modifications mainly concern the deposition parameters, the resuspension process, the ingestion model and the dose factors. To make results comparable, recalculations for pressurized light water reactors were performed with the release categories of the GRS. The results show in contrast to the findings of the GRS a significant reduction of the acute fatality risk by a factor of 3.6. This essentially results from the smaller deposition parameters. The latent fatality risk was calculated nearly unchanged. (orig.)

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

    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.

  10. Application of the CATHARE advanced code to numerical benchmark exercises

    In this work the CATHARE V1.3 code was applied to two Numerical Benchmark Exercises proposed in the aim to test the behaviour of thermalhydraulic codes against some spurious numerical effects. In the analysis of the first exercise, which is related to a two-phase flow along a convergent-divergent nozzle, it was seen that the CATHARE code well predicts the expected physical trends in good agreement with the majority of other codes. The application of the code to the second exercise, concerning the water packing phenomenon, showed the presence of this numerical effect. From a sensitivity analysis on time step and mesh size, it was seen that the amplitude of the pressure spikes can be reduced increasing time step and decreasing the space increments

  11. Application of numerical models and codes

    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.

  12. Modeling of continuous withdrawal and falling out of CPS control rods accident, using QUABOX/CUBBOX-HYCA code

    At present, at the Ignalina NPP the process of a wider use of the new uranium-erbium fuel of higher saturation and the manual control rods of new design is going on. These actions are directed to reducing the reactor control and protection system (CPS) cooling circuit voiding effect and to improving the technical and economical reactor operation parameters. Continuous withdrawal and falling out of CPS control rods lead to the reactivity and power changes in the reactor core. Therefore, important for safety is the evaluation of the CPS ability to compensate for the resulting excess reactivity in the reactor core, having the changed core loading conditions during such accidents. This article presents the calculation results of the continuous withdrawal and falling out of CPS control rods for the specific reactor core conditions of the Ignalina NPP Unit 2, i.e. during its operation on the maximum allowed power level of 4200 MW. The German code QUABOX/CUBBOX-HYCA with the improved CPS logic was used for the simulation of the above-mentioned transients. (author)

  13. Chemical speciation code CHEMSPEC and its applications

    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.

  14. Verification of fuel performance simulation codes: application of distributed computing

    The Canadian Nuclear Industry as a whole is attempting to find accurate, efficient and cost-effective methods to either test or demonstrate that complex simulation codes are behaving as they were intended to. One method that has been used at Ontario Power Generation is 'stress testing': running the code or code suite a sufficient number of times with small variations in input data, and tracking the resulting change in calculated results. This technique has only recently become practical for large simulation codes with the increasing power of desktop computers and the ability to dispatch batch jobs across a network; the method used for the work performed to date is described in detail in. The stress testing methodology is a balance between safety analysis needs and software engineering principles. It is not in general possible to completely cover the range of all input parameters and their interactions in a reasonable amount of time. Hence the number of input parameters to test was restricted to those that have the largest impact on the key output results (those outputs that are actually used directly or indirectly as safety criteria). Of particular value in identifying code deficiencies was 'parameter scanning', where one input parameter was varied using a fine increment over a range encompassing the normal variation. The key output results were then plotted against the varying input parameter, and the results studied to ensure that the predicted trends were correct and no non-physical discontinuities were present. This paper reports the results from two such stress testing exercises as examples of applications of the stress testing method. One application is on the ELESIM/ELOCA code suite to show how stress testing can identify problems in code testing and verification. The other application is from the FACTAR code suite focusing on how the stress testing method can be applied to a large code suite. (author)

  15. Parallelizing the MARS15 Code with MPI for shielding applications

    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

  16. Modification and application of the system analysis code ATHLET to trans-critical simulations

    Highlights: ► The pseudo two-phase method is proposed and utilized to modify ATHLET code. ► A smooth transition of void fraction under trans-critical transient can be realized by this method. ► The newly developed ATHLET-SC code can be adopted to simulate the blowdown process of a simplified model. - Abstract: During the loss of coolant accident (LOCA) of supercritical water cooled reactor (SCWR), the pressure in the reactor system will undergo a rapid decrease from supercritical to subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. Using the current version of system code (e.g. ATHLET, REALP), calculation will be terminated due to the abrupt change of void fraction across the critical point (22.064 MPa). To solve this problem, a pseudo two-phase method is proposed by introducing a fictitious region of latent heat (enthalpy of vaporization hfg∗) at pseudo-critical temperatures. A smooth transition of void fraction can be realized by using liquid-field conservation equations at temperatures lower than the pseudo-critical temperature, and vapor-field conservation equations at temperatures higher than the pseudo-critical temperature. Adopting this method, the system code ATHLET is modified to ATHLET-SC mod 2 on the basic of the previous version ATHLET-SC mod 1 modified by Shanghai Jiao Tong University. When the fictitious region of latent heat is kept as a small region, the code can achieve an acceptable accuracy. Moreover, the ATHLET-SC mod 2 code is applied to simulate the blowdown process of a simplified model. The results achieved so far indicate a good applicability of the new modified code for the trans-critical transient.

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

    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)

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

    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)

  19. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    Lazaro, A., E-mail: aulach@iqn.upv.es [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); Ammirabile, L. [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Schmitt, D. [EDF, 1 Avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St. Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)

    2014-10-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

  20. Smart phone Application using Morse Code and Inaudible Frequency

    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...

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

    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

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

    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

  3. Video over DSL with LDGM Codes for Interactive Applications

    Laith Al-Jobouri

    2016-05-01

    Full Text Available Digital Subscriber Line (DSL network access is subject to error bursts, which, for interactive video, can introduce unacceptable latencies if video packets need to be re-sent. If the video packets are protected against errors with Forward Error Correction (FEC, calculation of the application-layer channel codes themselves may also introduce additional latency. This paper proposes Low-Density Generator Matrix (LDGM codes rather than other popular codes because they are more suitable for interactive video streaming, not only for their computational simplicity but also for their licensing advantage. The paper demonstrates that a reduction of up to 4 dB in video distortion is achievable with LDGM Application Layer (AL FEC. In addition, an extension to the LDGM scheme is demonstrated, which works by rearranging the columns of the parity check matrix so as to make it even more resilient to burst errors. Telemedicine and video conferencing are typical target applications.

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

    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

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

    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.

  6. Establishment of Technical Collaboration basis between Korea and France for the development of severe accident assessment computer code under high burnup condition

    This project was performed by KAERI in the frame of construction of the international cooperative basis on the nuclear energy. This was supported from MOST under the title of 'Establishment of Technical Collaboration basis between Korea and France for the development of severe accident assessment computer code under high burn up condition'. The current operating NPP are converting the burned fuel to the wasted fuel after burn up of 40 GWD/MTU. But in Korea, burn up of more than 60 GWD/MTU will be expected because of the high fuel efficiency but also cost saving for storing the wasted fuel safely. The domestic research for the purpose of developing the fuel and the cladding that can be used under the high burn up condition up to 100 GWD/MTU is in progress now. But the current computer code adopts the model and the data that are valid only up to the 40 GWD/MTU at most. Therefore the current model could not take into account the phenomena that may cause differences in the fission product release behavior or in the core damage process due to the high burn up operation (more than 40 GWD/MTU). To evaluate the safety of the NPP with the high burn up fuel, the improvement of current severe accident code against the high burn up condition is an important research item. Also it should start without any delay. Therefore, in this study, an expert group was constructed to establish the research basis for the severe accident under high burn up conditions. From this expert group, the research items regarding the high burn up condition were selected and identified through discussion and technical seminars. Based on these selected items, the meeting between IRSN and KAERI to find out the cooperative research items on the severe accident under the high burn up condition was held in the IRSN headquater in Paris. After the meeting, KAERI and IRSN agreed to cooperate with each other on the selected items, and to co-host the international seminar, and to develop the model and to

  7. Use of MAAP code for identification of key plant vulnerabilities for the beyond design accidents and their mitigation at NPP Krsko

    NPP Krsko performed according to GL 88-20, Supplement 1-4 and RUJV requirement the Individual Plant Examination analyses. For the required deterministic analyses the MAAP 3.0B code was used. It was proven that such severe accident analysis can be used for evaluation of the overall level of safety improvement that can be gained with the different modifications and alternate design. In this paper one such important outcomes from these analyses will be presented. (author)

  8. Utilization of the RELAP4/MOD5/SAS code version in loss of coolant accident in the Angra 1 nuclear power station

    A new version of computer code RELAP4/MOD5 was developed to improve the output. The new version, called RELAP4/MOD5/SAS, prints the main variables in graphical form. In order to check the program, a 36 - volume simulation of the Loss-of-Coolant Accident for Angra - I was performed and the results compared to those of a existing 44 - volume simulation showed a satisfactory agreement with a substantial reduction in computing time. (author)

  9. Metal oxide aerosol dry deposition in laminar pipe flow at high thermal gradients and comparison with SOPHAEROS module of ASTEC reactor accident analysis code

    Highlights: • Experiments to simulate aerosol deposition in pipes in severe accident condition. • Laminar flow conditions under high thermal gradient have been used. • Results of the experiments have been compared with SOPHAEROS module of the ASTEC code. - Abstract: During a severe nuclear reactor accident involving core melt down conditions, the deposition of fission product aerosols inside the reactor coolant system affects the final source term available to the containment and subsequently to the environment. Towards quantifying the aerosol deposition under varying flow conditions and thermal gradients, as may be encountered in the heat transport systems, experiments were performed to investigate the dry deposition behavior of metal oxide aerosols in a 3.6 m long stainless steel piping test assembly. This assembly consisted of divergent and convergent sections, horizontal and vertical sections and right angle bends. Tin oxide aerosols, generated by a plasma torch aerosol generator, were transported into the test assembly using argon carrier gas. Temperature sensors coupled to data loggers were used to record the pipe inner wall and carrier gas temperatures. The experimental deposition results were found to be within 8% of those estimated by the SOPHAEROS module of the accident analysis code ASTEC (Accident Source Term Evaluation Code). Code results for experimental input parameters showed that for sections at higher temperature gradients the dominant deposition mechanism was thermophoresis, while in sections for low thermal gradients, gravitational settling dominated. The micrographs obtained using Transmission Electron Microscopy (TEM) showed that the deposited Tin oxide particles were mostly spherical and bimodal in nature. The X-ray diffraction (XRD) analysis showed that plasma torch generated aerosols exhibit tetragonal SnO and SnO2 phases

  10. The Monte Carlo code TRAMO - Capabilities and instructions for application

    The report is intended for readers familiar with the fundamentals of the Monte Carlo method. Those readers might be interested in learning about successful generalisations as well as new ideas for curbing the statistical errors involved. Another intention however is to explain the significant basic features of the multigroup Monte Carlo code TRAMO, including the required input, so that readers will be able to performing the required adjustments to the specific calculation technique and develop their own tools for performing their specific calculations. An indispensable code needed for such TRAMO applications is the TRAWEI Monte Carlo code which calculates he required weightings for applications of the variance reducing Weight Window Method; other codes required are those for generating the neutron cross-section data and the group data. The TRAMO code calculates, with given source distribution of neutrons in multigroup approximation, multigroup flux data, integrated group flux data, and dose values for given partial volumes and surfaces. There are further code versions for calculation of neutron and gamma fluxes, or criticality data, but these are not considered in the report. (orig./CB)

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

    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)

  12. CONSUL code package application for LMFR core calculations

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

    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)

  14. 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

    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.)

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

    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

  16. Adaptive coded aperture imaging: progress and potential future applications

    Gottesman, Stephen R.; Isser, Abraham; Gigioli, George W., Jr.

    2011-09-01

    Interest in Adaptive Coded Aperture Imaging (ACAI) continues to grow as the optical and systems engineering community becomes increasingly aware of ACAI's potential benefits in the design and performance of both imaging and non-imaging systems , such as good angular resolution (IFOV), wide distortion-free field of view (FOV), excellent image quality, and light weight construct. In this presentation we first review the accomplishments made over the past five years, then expand on previously published work to show how replacement of conventional imaging optics with coded apertures can lead to a reduction in system size and weight. We also present a trade space analysis of key design parameters of coded apertures and review potential applications as replacement for traditional imaging optics. Results will be presented, based on last year's work of our investigation into the trade space of IFOV, resolution, effective focal length, and wavelength of incident radiation for coded aperture architectures. Finally we discuss the potential application of coded apertures for replacing objective lenses of night vision goggles (NVGs).

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

    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)

  18. Assessment of Radiological and Economic Consequences of a Hypothetical Accident for ETRR-2, Egypt Utilizing COSYMA Code

    A comprehensive probabilistic study of an accident consequence assessment (ACA) for loss of coolant accident (LOCA) has accomplished to the second research reactor ETRR-2, located at Inshas Nuclear Research Center, Cairo, Egypt. PC-COSYMA, developed with the support of European Commission, has adopted to assess the radiological and economic consequences of a proposed accident. The consequences of the accident evaluated in case of early and late effects. The effective doses and doses in different organs carried out with and without countermeasures. The force mentioned calculations were required the following studies: the core inventory due to the hypothetical accident, the physical parameters of the source term, the hourly basis meteorological parameters for one complete year, and the population distribution around the plant. The hourly stability conditions and height of atmospheric boundary layers (ABL) of the concerned site were calculated. The results showed that, the nuclides that have short half-lives (few days) give the highest air and ground concentrations after the accident than the others. The area around the reactor requires the early and late countermeasures action after the accident especially in the downwind sectors. Economically, the costs of emergency plan are effectively high in case of applying countermeasures but countermeasures reduce the risk effects

  19. Consistent Code Qualification Process and Application to WWER-1000 NPP

    Calculation analysis by application of the system codes are performed to evaluate the NPP or the facility behavior during a postulated transient or to evaluate the code capability. The calculation analysis constitutes a process that involves the code itself, the data of the reference plant, the data about the transient, the nodalization, and the user. All these elements affect one each other and affect the results. A major issue in the use of mathematical model is constituted by the model capability to reproduce the plant or facility behavior under steady state and transient conditions. These aspects constitute two main checks that must be satisfied during the qualification process. The first of them is related to the realization of a scheme of the reference plant; the second one is related to the capability to reproduce the transient behavior. The aim of this paper is to describe the UMAE (Uncertainty Method based on Accuracy Extrapolation) methodology developed at University of Pisa for qualifying a nodalization and analysing the calculated results and to perform the uncertainty evaluation of the system code by the CIAU code (Code with the capability of Internal Assessment of Uncertainty). The activity consists with the re-analysis of the Experiment BL-44 (SBLOCA) performed in the LOBI facility and the analysis of a Kv-scaling calculation of the WWER-1000 NPP nodalization taking as reference the test BL-44. Relap5/Mod3.3 has been used as thermal-hydraulic system code and the standard procedure adopted at University of Pisa has been applied to show the capability of the code to predict the significant aspects of the transient and to obtain a qualified nodalization of the WWER-1000 through a systematic qualitative and quantitative accuracy evaluation. The qualitative accuracy evaluation is based on the selection of Relevant Thermal-hydraulic Aspects (RTAs) and is a prerequisite to the application of the Fast Fourier Transform Based Method (FFTBM) which quantifies

  20. The combined thermohydraulics-neutronics code TRAB-SMABRE for 3D plant transient and accident analyses

    of the flow reversal, which was not possible with the parallel coupling. This is not a large advancement, but with the new work the code package becomes more transportable than earlier. The advance using the old reactor physical basis is that all experiences tailored into the code since 1970's are included at the same. Choosing BWR as the first reference case was natural, because the 3-D core simulation for BWR plants is more challenging than that for PWR plants. After BWR application and code application range will be expanded for PWR reactors as well. The first validation cases have been related to the BWR start-up tests and real plant transients. For the Olkiluoto plant the load rejection experiment with a partial shutdown, a start-up test for the Olkiluoto plant, resulting with an asymmetric neutron flux profile has been calculated. The comparison results are displayed. An over-pressurization transient occurred in the Olkiluoto plant due to an erroneous function of the pressure controller, and the transient has been calculated against the available plant data. The comparison results are presented. (authors)

  1. Optical code division multiple access fundamentals and applications

    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

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

    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

  3. Application experiences of ASME nuclear codes in Korea

    ASME Section III has been the most important Code for the design and construction of nuclear components for over 30 years in Korea. During that time, some difficulties have been recognized in compliance with the Code due to the differences in industrial practices and regulatory system between Korea and the U.S. In case of NCA(General Requirements for Division 1 and 2), the administrative and procedural requirements have been applied as modified to suit domestic industry practices. For technical aspects, supplemental requirements have been added to ASME Code in order to satisfy regulatory guides. Preheating and PWHT requirements of ASME Section III are slightly different from those of ASME B 31.1. The differences are discussed in this presentation. With the issuance of MOST(Ministry of Science and Technology) notice 96-32 in 1996 regarding KEPIC(Korea Electric Power Industry Code) application, KEPIC MN has been used in selected documents instead of ASME Sec. III. Ulchin 5 and 6 is the first project that applied to KEPIC. The application of KEPIC - MN has been gradually expanding to the subsequent nuclear projects. For concrete settlement of KEPIC, some measures for foreign vendor's application of KEPIC should be considered

  4. Application of advanced statistical methods in assessment of the late phase of a nuclear accident

    Hofman, Radek

    Praha: ČVUT, 2008, s. 1-4. [Dny radiacni ochrany /30/.. Liptovsky Jan (SK), 10.11.2008-14.11.2008] R&D Projects: GA ČR(CZ) GA102/07/1596 Institutional research plan: CEZ:AV0Z10750506 Keywords : radiation protection Subject RIV: DI - Air Pollution ; Quality http://library.utia.cas.cz/separaty/2008/AS/hofman-application of advanced statistical methods in assessment of the late phase of a nuclear accident.pdf

  5. Automatic differentiation of codes in nuclear engineering applications

    We discuss our experience in applying automatic differentiation (AD) to calculations in nuclear reactor applications. The document is intended as a guideline on how to apply AD to Fortran codes with significant legacy components; it is also a part of a larger research effort in uncertainty quantification using sampling methods augmented with derivative information. We provide a brief theoretical description of the concept of AD, explain the necessary changes in the code structure, and remark on possible ways to deal with non-differentiability. Numerical experiments were carried out where the derivative of a functional subset of the SAS4A/SASSYS code was computed in forward mode with several AD tools. The results are in good agreement with both the real and complex finite-difference approximations of the derivative.

  6. Automatic differentiation of codes in nuclear engineering applications.

    Alexe, M.; Roderick, O.; Utke, J.; Anitescu, M.; Hovland, P.; Fanning, T.; Virginia Polytechnic Inst. and State Univ.; Unv. of Chicago

    2009-12-01

    We discuss our experience in applying automatic differentiation (AD) to calculations in nuclear reactor applications. The document is intended as a guideline on how to apply AD to Fortran codes with significant legacy components; it is also a part of a larger research effort in uncertainty quantification using sampling methods augmented with derivative information. We provide a brief theoretical description of the concept of AD, explain the necessary changes in the code structure, and remark on possible ways to deal with non-differentiability. Numerical experiments were carried out where the derivative of a functional subset of the SAS4A/SASSYS code was computed in forward mode with several AD tools. The results are in good agreement with both the real and complex finite-difference approximations of the derivative.

  7. The FLUKA code for space applications: recent developments

    Andersen, V.; Ballarini, F.; Battistoni, G.; Campanella, M.; Carboni, M.; Cerutti, F.; Empl, A.; Fasso, A.; Ferrari, A.; Gadioli, E.; Garzelli, M. V.; Lee, K.; Ottolenghi, A.; Pelliccioni, M.; Pinsky, L. S.; Ranft, J.; Roesler, S.; Sala, P. R.; Wilson, T. L.; Townsend, L. W. (Principal Investigator)

    2004-01-01

    The FLUKA Monte Carlo transport code is widely used for fundamental research, radioprotection and dosimetry, hybrid nuclear energy system and cosmic ray calculations. The validity of its physical models has been benchmarked against a variety of experimental data over a wide range of energies, ranging from accelerator data to cosmic ray showers in the earth atmosphere. The code is presently undergoing several developments in order to better fit the needs of space applications. The generation of particle spectra according to up-to-date cosmic ray data as well as the effect of the solar and geomagnetic modulation have been implemented and already successfully applied to a variety of problems. The implementation of suitable models for heavy ion nuclear interactions has reached an operational stage. At medium/high energy FLUKA is using the DPMJET model. The major task of incorporating heavy ion interactions from a few GeV/n down to the threshold for inelastic collisions is also progressing and promising results have been obtained using a modified version of the RQMD-2.4 code. This interim solution is now fully operational, while waiting for the development of new models based on the FLUKA hadron-nucleus interaction code, a newly developed QMD code, and the implementation of the Boltzmann master equation theory for low energy ion interactions. c2004 COSPAR. Published by Elsevier Ltd. All rights reserved.

  8. Evaluation of confinement capability of radioactive materials under the fire accident in nuclear fuel facility with CELVA-1D (Cell Ventillation Analysis Code-1D)

    To reduce the clogging of smoke on the HEPA filters under the fire accident, some of ventilation systems in the plant are equipped with the pre-filters in front of the HEPA filters for collecting the relatively large smoke particles. Appropriate correspondence such as the exchange of the pre-filter is important for confinement of radioactive materials in the ventilation system under the fire accident. To study smoke generation behavior due to the burnable wastes and clogging properties of the ventilation filters by smoke loading, the verification test has been performed. The cell ventilation system analysis code, CELVA-1D was used for analysis of smoke generation and the rising of pressure drop at both the pre-filter and the HEPA filter. With the change of source term, the breakage of time of the pre-filter was also estimated. (author)

  9. Application of Best Estimate Thermalhydraulic Codes for the Safety Analysis of Research Reactors

    With the progress in computer technology and numerical methods, the capabilities of computer codes have been substantially enlarged. Consequently, advanced safety analysis and design optimisations are increasingly performed. The application of Best Estimate (BE) method and tools allows getting more realistic simulation of the complex processes taking place during the steady state operation and transients in Research Reactors. This paper presents the application of two international codes RELAP5 and CATHENA to the German FRJ-2 Research Reactor (RR) for the analysis of a hypothetical accident. The work mainly aims at the comparison of the results of the two sophisticated codes under the consideration of the modelling differences and the numerical approaches respectively. The two codes have been developed for the analysis of the steady state and transient behaviour of reactor core and piping components on the basis of complex heat transfer and hydraulic models. The work consists of the simulation of a hypothetical fast reactivity transient, which is assumed to be initiated by the failure of one shutdown arm causing a reactivity insertion by an amount of 2.7 percent dk/k during the reactor operation. The case has been chosen due to the importance of the models for the precise description of complex phenomenon of subcooled boiling and two phase flow taking place during the transient. For the calculation, the fuel element assembly including the individual cooling channels were modelled in detail. The primary circuit was included in the whole model in order to consider the interaction with the core and the individual parallel channels of the fuel elements. In general the results of the two codes are in agreement and comparable with respect to the thermalhydraulic processes taking place in the course of the transient. The comparison reveals that models, constitutive relations and correlations employed in both codes have considerable influence on the results of the

  10. Adaptation of the ATHLET code to ADS applications

    Full text of publication follows: In Europe, heavy liquid metal (HLM), such as lead bismuth eutectic (LBE), is considered as the reference coolant for both the subcritical reactor core and the spallation target of an accelerator driven subcritical reactor system (ADS) due to its low melting point, efficient heat removal properties and high production rate of neutrons. System analysis is required for the thermal hydraulic design of both the reactor cooling system and the spallation target cooling system of ADS. Since many years, the Forschungszentrum Karlsruhe has been strongly involved in the design of different kinds of spallation targets. Two independent one-dimensional system codes, HETRAF and HERETA, have been developed and applied to the design analysis of various spallation targets of ADS. Most recently, the system analysis code ATHLET, which was developed at the Gesellschaft fuer Reaktorsicherheit (GRS) and widely applied to water cooled nuclear systems, has being modified for ADS applications. The numerical scheme was improved to cope with the application to multi-fluid systems. The code structure was modified in such a way that the user can easily adapt the code for various fluids. In the present version, the packages of heat transfer and physical properties suitable for various fluids, such as lead-bismuth eutectic and organic diathermic oil, were implemented into the code. The modified ATHLET code, ATHLET-ADS, is applied to the cooling systems of various spallation targets, e.g. the target of the European Experimental ADS (XADS). The LBE cooled XADS target is a pool type compact module. The LBE is circulated by natural circulation as it heated up by the proton beam in the spallation zone below the beam window, while it is cooled down in the upper part of the target unit, where the heat is removed through a heat exchanger by means of diathermic oil. Analyses are performed for steady state operation and various transient scenarios, e.g. beam power switch

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

    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

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

    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

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

    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....

  14. CONSUL code package application for LMFR core calculations

    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)

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

    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

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

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

  17. Review of aviation safety measures which have application to aviation accident prevention.

    Doughtery, J D

    1975-01-01

    Introduction of certain human-factors techniques has been followed by market reduction in military and airline accident rates. In this study, these safety measures are analyzed to determine the value of their application to general aviation activity. Some techniques are already in use. They are: 1. medical evaluation of iarcrews; 2. aeronautical innovations which tailor the machine to the man; 3. imporvement of precision navigational air traffic control and flight procedures; 4. standardization of flight training and flight procedures. A remaining field of interest, and one which appears to be underused, is that of supervision. After ending his association with the flight instructor, the general aviation pilot is essentially unsupervised. Accident data gathered over several years show that with increases in the proportion of pilots who have not maintained an association with a flight instructor, the general aviation fatal accident rate is increased. Current regulations, which require revalidation of airman's certificates, provide a method by which this association can be maintained. The flight instructor, or some similar aviation professional, can maintain an element of supervision with otherwise independent general aviation pilots. Data from previous years supports the hypothesis that such a program would make a substantial improvement in general aviation safety. PMID:1115703

  18. Radiation application technologies used for countermeasures against Fukushima Daiichi Nuclear Power Station accident

    Radiation application technologies had contributed to countermeasures against Fukushima Daiichi Nuclear Power Station Accident. After the accident, such activities had been continued to assure the safety of damaged cores towards the settlement of the accident and countermeasures against radioactive materials discharged and diffused into the environment. As for recovery and removal of radioactive materials in the environment, development of cesium adsorbent using graft polymerization and trial of environmental restoration of agricultural land using plant breeding suitable for cesium absorption had been started. Database was available for evaluation of radiation resistance of polymer materials and equipments. Related with general-purpose cranes and robots used under high radiation environment, guideline of radiation life assessment of universal electric parts was proposed to limit total radiation doses. Radiation resistance of semiconductor devices of electric parts used for Quince robots was tested at cobalt 60 gamma-ray irradiation facility. Radiolysis of boiling pure water was tested under gamma-ray irradiation to clarify hydrogen generation mechanism. (T. Tanaka)

  19. ATHLET code and its application in low temperature heating reactor

    The thermal-hydraulic computer code ATHLET (Analysis of Thermal-Hydraulics of Leaks and Transients) is developed by the GRS (Gesellschaft fuer Anlagen-und Reaktorsicherheit) for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors. Besides conventional and advanced PWR and BWR, the range of applicability is being extended to the Russian reactor types VVER and RBMK and Canadian CANDU. The PC-based version of ATHELT MOD 1.2 A code was introduced and the analysis results of steady-state natural circulation for 5 MW low temperature heating reactor of Tsinghua University was reported. The comparison has shown that the calculation results agree satisfactorily well with the operational test data of 5 MW heating reactor

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

    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.)

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

    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

  2. Progress in Addressing DNFSB Recommendation 2002-1 Issues: Improving Accident Analysis Software Applications

    VINCENT, ANDREW

    2005-04-25

    Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1 (''Quality Assurance for Safety-Related Software'') identified a number of quality assurance issues on the use of software in Department of Energy (DOE) facilities for analyzing hazards, and designing and operating controls to prevent or mitigate potential accidents. Over the last year, DOE has begun several processes and programs as part of the Implementation Plan commitments, and in particular, has made significant progress in addressing several sets of issues particularly important in the application of software for performing hazard and accident analysis. The work discussed here demonstrates that through these actions, Software Quality Assurance (SQA) guidance and software tools are available that can be used to improve resulting safety analysis. Specifically, five of the primary actions corresponding to the commitments made in the Implementation Plan to Recommendation 2002-1 are identified and discussed in this paper. Included are the web-based DOE SQA Knowledge Portal and the Central Registry, guidance and gap analysis reports, electronic bulletin board and discussion forum, and a DOE safety software guide. These SQA products can benefit DOE safety contractors in the development of hazard and accident analysis by precluding inappropriate software applications and utilizing best practices when incorporating software results to safety basis documentation. The improvement actions discussed here mark a beginning to establishing stronger, standard-compliant programs, practices, and processes in SQA among safety software users, managers, and reviewers throughout the DOE Complex. Additional effort is needed, however, particularly in: (1) processes to add new software applications to the DOE Safety Software Toolbox; (2) improving the effectiveness of software issue communication; and (3) promoting a safety software quality assurance culture.

  3. Progress in Addressing DNFSB Recommendation 2002-1 Issues: Improving Accident Analysis Software Applications

    Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1 (''Quality Assurance for Safety-Related Software'') identified a number of quality assurance issues on the use of software in Department of Energy (DOE) facilities for analyzing hazards, and designing and operating controls to prevent or mitigate potential accidents. Over the last year, DOE has begun several processes and programs as part of the Implementation Plan commitments, and in particular, has made significant progress in addressing several sets of issues particularly important in the application of software for performing hazard and accident analysis. The work discussed here demonstrates that through these actions, Software Quality Assurance (SQA) guidance and software tools are available that can be used to improve resulting safety analysis. Specifically, five of the primary actions corresponding to the commitments made in the Implementation Plan to Recommendation 2002-1 are identified and discussed in this paper. Included are the web-based DOE SQA Knowledge Portal and the Central Registry, guidance and gap analysis reports, electronic bulletin board and discussion forum, and a DOE safety software guide. These SQA products can benefit DOE safety contractors in the development of hazard and accident analysis by precluding inappropriate software applications and utilizing best practices when incorporating software results to safety basis documentation. The improvement actions discussed here mark a beginning to establishing stronger, standard-compliant programs, practices, and processes in SQA among safety software users, managers, and reviewers throughout the DOE Complex. Additional effort is needed, however, particularly in: (1) processes to add new software applications to the DOE Safety Software Toolbox; (2) improving the effectiveness of software issue communication; and (3) promoting a safety software quality assurance culture

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

    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

  5. Best Estimate Thermal-Hydraulic System Analysis using the MARS Code for the Steam Generator Tube Rupture Accident in the APR1400

    A postulated SGTR (Steam Generator Tube Rupture) accident of the APR1400 was analysed using the best estimate safety analysis code, MARS (Multidimensional Analysis of Reactor Safety). The SGTR accident is one of the design basis accidents, which has a unique feature of the penetration of the barrier between the reactor coolant system (RCS) and the secondary system resulting from the failure of a steam generator U-tube. The SGTR has an importance in safety due to a concern of a containment bypass of radioactive inventory. In the course of the SGTR, the radioactivity leaking from a broken steam generator Utube mixes with the shell-side water in an affected steam generator. Leak flow from ruptured U-tubes can increase a water level and a pressure of the affected steam generator. Following a reactor trip and a turbine trip, the main steam safety valves (MSSVs) can be open to mitigate an increase in the secondary system pressure. Meanwhile, the SGTR can provide a direct flow path from the primary to the secondary system resulting in the release of fission products into the atmosphere. As one of the most limiting SGTR accidents, a leak flow equivalent to a double-ended rupture of five Utubes was analysed in this study. 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 the HSGL (High Steam Generator Level) trip and the LPP (Low Pressurizer Pressure) trip on the thermal-hydraulic system response

  6. Investigation of Local Effects Influence on Results of Design Basis Accident Analysis of WWER-440 Reactor Using RELAP5-3D Code

    One of the most important tasks in today's nuclear power plant safety analysis is a simulation of physical processes at nuclear facilities which accounts for 3-dimensional effects in the core and downcomer of reactor. System coupled thermo-hydraulic/neutron-kinetic code RELAP5-3D, which is a modeling tool provided to University of Kyiv by US DOE in a frame of International Nuclear Safety Program, allows simulation of variable in time spatial distribution of neutron flux in a core and also includes special components for 3D modeling of thermo-hydraulics. A model of Rivne NPP Unit 1 with WWER-440/V-213 type reactor has been developed for RELAP5-3D code. A scenario of 'Main steam line break' design basis accident has been calculated using this model. Such a problem can be characterized by intensive overcooling of a primary coolant in affected loop and, taking into account partial mixing of coolant from different primary loops, a non-uniform cooling of reactor core. Obtained results have been compared with the results obtained by model, which has been used at Design Based Accidents analysis, performed at specified unit.(author)

  7. Assessment of induction elbows for an ASME Code application

    The ASME Nuclear Codes impose some specific requirements on the wall thickness uniformity and the out-of-roundness of cross sections of the elbows for Nuclear Power Plant applications. Due to some of these requirements, manufacturing and installation of these elbows can be time consuming and quite expensive. This paper explores the feasibility of using induction elbows for nuclear application from the stress analysis point of view. To this end, three different sizes of 90deg elbows have been analyzed based on the geometry of an 'ASME Code' elbow and an elbow formed by induction bending. The analysis is carried out for internal pressure, in-plane and out-of-plane loads. Based on the results of these three carbon steel elbows, the use of induction elbows in some of the CANDU-PHW (CANadian Deuterium Uranium-Pressurized Heavy Water) power plant applications seems encouraging. However, before the feasibility can be fully confirmed analysis and induction bending tests over a wider range of geometries, loading conditions, and materials are required. (author)

  8. 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

    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)

  9. All-optical code-division multiple-access applications: 2(n) extended-prime codes.

    Zhang, J G; Kwong, W C; Mann, S

    1997-09-10

    A new family of 2(n) codes, called 2(n) extended-prime codes, is proposed for all-optical code-division multiple-access networks. Such 2(n) codes are derived from so-called extended-prime codes so that their cross-correlation functions are not greater than 1, as opposed to 2 for recently proposed 2(n) prime codes. As a result, a larger number of active users can now be supported by the new codes for a given bit-error rate than can be by 2(n) prime codes, while power-efficient, waveguide-integrable all-serial coding and correlating configurations proposed for the 2(n) prime codes can still be employed. PMID:18259529

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

    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)

  11. FISH: A 3D parallel MHD code for astrophysical applications

    Kaeppeli, R; Scheidegger, S; Pen, U -L; Liebendörfer, M

    2009-01-01

    FISH is a fast and simple ideal magneto-hydrodynamics code that scales to ~10 000 processes for a Cartesian computational domain of ~1000^3 cells. The simplicity of FISH has been achieved by the rigorous application of the operator splitting technique, while second order accuracy is maintained by the symmetric ordering of the operators. Between directional sweeps, the three-dimensional data is rotated in memory so that the sweep is always performed in a cache-efficient way along the direction of contiguous memory. Hence, the code only requires a one-dimensional description of the conservation equations to be solved. This approach also enable an elegant novel parallelisation of the code that is based on persistent communications with MPI for cubic domain decomposition on machines with distributed memory. This scheme is then combined with an additional OpenMP parallelisation of different sweeps that can take advantage of clusters of shared memory. We document the detailed implementation of a second order TVD ad...

  12. Investigation of the applicability of MCNP code to complicated geometries

    Applicability of MCNP code, which is a general purpose Monte Carlo code for particle transport problems, to complicated geometries, has been investigated as a study in Human Acts Simulation Program (HASP), in which basic studies for intelligent robot for patrol and inspection of nuclear facilities are being performed. In HASP, basic software systems simulating the behavior of intelligent robot of human shape working in Japan Research Reactor No.3 are being developed. The aim of Dose Evaluation system in HASP is to establish the methodology to evaluate irradiation damage of the LSI/VLSI circuits embedded within a robot body and to give design criteria of intelligent robot. Monte Carlo method is used to solve particle transport problem in a complicated geometry such as robot body. Preliminary evaluation to establish the methodology has been conducted using continuous energy Monte Carlo code, MCNP with the anthropomorphic phantom. The phantom has the same degree of geometric complexity as robot body and is widely used for the calculation of the effective dose equivalent for radiological protection. It allowed us to verify the validity of the methodology by comparison of calculation results with the data in ICRP Pub. 51. In this report, the method used in the calculation of effective dose equivalent, visualization system supporting visualization of input data for complicated geometry and the results in the evaluation of validity of the method by the comparison of the calculated results with the data in the ICRP publication are described. (author)

  13. The Application of the PEBBED Code Suite to the PBMR-400 Coupled Code Benchmark - FY 2006 Annual Report

    2006-09-01

    This document describes the recent developments of the PEBBED code suite and its application to the PBMR-400 Coupled Code Benchmark. This report addresses an FY2006 Level 2 milestone under the NGNP Design and Evaluation Methods Work Package. The milestone states "Complete a report describing the results of the application of the integrated PEBBED code package to the PBMR-400 coupled code benchmark". The report describes the current state of the PEBBED code suite, provides an overview of the Benchmark problems to which it was applied, discusses the code developments achieved in the past year, and states some of the results attained. Results of the steady state problems generated by the PEBBED fuel management code compare favorably to the preliminary results generated by codes from other participating institutions and to similar non-Benchmark analyses. Partial transient analysis capability has been achieved through the acquisition of the NEM-THERMIX code from Penn State University. Phase I of the task has been achieved through the development of a self-consistent set of tools for generating cross sections for design and transient analysis and in the successful execution of the steady state benchmark exercises.

  14. Analysis of severe accidents in pressurized heavy water reactors

    Certain very low probability plant states that are beyond design basis accident conditions and which may arise owing to multiple failures of safety systems leading to significant core degradation may jeopardize the integrity of many or all the barriers to the release of radioactive material. Such event sequences are called severe accidents. It is required in the IAEA Safety Requirements publication on Safety of the Nuclear Power Plants: Design, that consideration be given to severe accident sequences, using a combination of engineering judgement and probabilistic methods, to determine those sequences for which reasonably practicable preventive or mitigatory measures can be identified. Acceptable measures need not involve the application of conservative engineering practices used in setting and evaluating design basis accidents, but rather should be based on realistic or best estimate assumptions, methods and analytical criteria. Recently, the IAEA developed a Safety Report on Approaches and Tools for Severe Accident Analysis. This publication provides a description of factors important to severe accident analysis, an overview of severe accident phenomena and the current status in their modelling, categorization of available computer codes, and differences in approaches for various applications of severe accident analysis. The report covers both the in- and ex-vessel phases of severe accidents. The publication is consistent with the IAEA Safety Report on Accident Analysis for Nuclear Power Plants and can be considered as a complementary report specifically devoted to the analysis of severe accidents. Although the report does not explicitly differentiate among various reactor types, it has been written essentially on the basis of available knowledge and databases developed for light water reactors. Therefore its application is mostly oriented towards PWRs and BWRs and, to a more limited extent, they can be only used as preliminary guidance for other types of reactors

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

    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.

  16. Comparison of published geochemical codes and application of the PHREEQE code to selected cases

    In future safety analyses of nuclear repositories the question about the implementation of geochemical codes will be raised. The use of geochemical codes in international excercises, particularly the CHEMVAL project, was discussed and general requirements to be met by such codes in safety analyses of final repositories were formulated. The performance characteristics of three geochemical codes - MINEQL, PHREEQE and EQ3/6 - were listed and general limitations were discussed when applying geochemical codes to natural systems. Additionally, the PHREEQE code was applied to two simple natural systems. (orig.)

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

    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

  18. Application of Bayesian nonparametric models to the uncertainty and sensitivity analysis of source term in a BWR severe accident

    A full-scope method is constructed to reveal source term uncertainties and to identify influential inputs during a severe accident at a nuclear power plant (NPP). An integrated severe accident code, MELCOR Ver. 1.8.5, is used as a tool to simulate the accident similar to that occurred at Unit 2 of the Fukushima Daiichi NPP. In order to figure out how much radioactive materials are released from the containment to the environment during the accident, Monte Carlo based uncertainty analysis is performed. Generally, in order to evaluate the influence of uncertain inputs on the output, a large number of code runs are required in the global sensitivity analysis. To avoid the laborious computational cost for the global sensitivity analysis via MELCOR, a surrogate stochastic model is built using a Bayesian nonparametric approach, Dirichlet process. Probability distributions derived from uncertainty analysis using MELCOR and the stochastic model show good agreement. The appropriateness of the stochastic model is cross-validated through the comparison with MELCOR results. The importance measure of uncertain input variables are calculated according to their influences on the uncertainty distribution as first-order effect and total effect. The validity of the present methodology is demonstrated through an example with three uncertain input variables. - Highlights: • A method of source term uncertainty and sensitivity analysis is proposed. • Source term in Fukushima Daiichi NPP severe accident is demonstrated. • Uncertainty distributions of source terms show non-standard shapes. • A surrogate model for integrated code is constructed by using Dirichlet process. • Importance ranking of influential input variables is obtained

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

    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)

  20. Coupling of Time-Dependent Neutron Transport Theory with the Thermal Hydraulics Code ATHLET and Application to the Research Reactor FRM-II

    We introduce a new coupled neutronics/thermal hydraulics code system for analyzing transients of nuclear power plants and research reactors, based on a neutron transport theory approach. For the neutron kinetics, we have developed the code DORT-TD, a time-dependent extension of the well-known discrete ordinates code DORT. DORT-TD uses a fully implicit time integration scheme and is coupled via a general interface to the thermal hydraulics system code ATHLET, a generally applicable code for the analyses of LWR accident scenarios. Feedback is accounted for by interpolating multigroup cross sections from precalculated libraries, which are generated in advance for user-specified, discrete sets of thermal hydraulic parameters, e.g., fuel and coolant temperature. The coupled code system is applied to the high-flux research reactor FRM-II (Germany). Several design basis accidents are considered, namely the unintended control rod withdrawal, the loss of offsite power, and the loss of the secondary heat sink as well as a hypothetical transient with large reactivity insertion

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

    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

  2. Network Coding Applications and Implementations on Mobile Devices

    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....

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

    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.

  4. Development of simplified evaluation models for the first power peak during a criticality accident and its verification by the TRACE code simulated results based on CRAC experimental data

    In a reprocessing facility or a part of uranium fuel manufacturing facility where nuclear fuel solution is processed, one could frequently observe a series of power peaks with the first highest right after a criticality accident. The criticality alarm system (CAS) is designed to detect the first power peak and immediately warn workers around the reacting material by any means such as sounding alarms. Consequently, exposure of the workers could be minimized by an immediate and effective evacuation. Therefore in the design and installation of CAS, it is necessary to estimate the magnitude of the first power peak and to set up the threshold point for CAS initiating alarm. Furthermore, it is necessary to estimate the potential level of accidental exposure of workers so as to decide whether or not it is appropriate to install CAS for any compartment. In this report, simplified evaluation models to estimate the minimum scale of the first power peak and the released energy during a criticality accident are derived only by theoretical consideration for use in the design of CAS to set up the threshold point triggering the alarm signal. Other simplified evaluation models are in the same way derived to estimate the maximum scale of the first power peak and the released energy and to predict possible exposure level of workers to be used to judge the appropriateness of CAS installation. These evaluation models are shown to have adequate margin in predicting the minimum and maximum scale of criticality accidents by comparing their results with French CRAC experiment data. Furthermore, comparison of the maximum scale of the first power peak simplified evaluation, has been made with simulated results by the TRACE code based on the extrapolated conditions predicted by the CRAC experiment data to verify the effectiveness of the derived evaluation models

  5. The FLUKA code for space applications Recent developments

    Andersen, V; Battistoni, G; Campanella, M; Carboni, M; Cerutti, F; Empl, A; Fassò, A; Ferrari, A; Gadioli, E; Garzelli, M V; Lee, K; Ottolenghi, A; Pelliccioni, M; Pinsky, L S; Ranft, J; Roesler, S; Sala, P R; Wilson, T L

    2004-01-01

    The FLUKA Monte Carlo transport code is widely used for fundamental research, radioprotection and dosimetry, hybrid nuclear energy system and cosmic ray calculations. The validity of its physical models has been benchmarked against a variety of experimental data over a wide range of energies, ranging from accelerator data to cosmic ray showers in the earth atmosphere. The code is presently undergoing several developments in order to better fit the needs of space applications. The generation of particle spectra according to up-to- date cosmic ray data as well as the effect of the solar and geomagnetic modulation have been implemented and already successfully applied to a variety of problems. The implementation of suitable models for heavy ion nuclear interactions has reached an operational stage. At medium/high energy FLUKA is using the DPMJET model. The major task of incorporating heavy ion interactions from a few GeV/n down to the threshold for inelastic collisions is also progressing and promising results h...

  6. UNICOS CPC6: automated code generation for process control applications

    The Continuous Process Control package (CPC) is one of the components of the CERN Unified Industrial Control System framework (UNICOS). As a part of this framework, UNICOS-CPC provides a well defined library of device types, a methodology and a set of tools to design and implement industrial control applications. The new CPC version uses the software factory UNICOS Application Builder (UAB) to develop CPC applications. The CPC component is composed of several platform oriented plug-ins (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 generator (based on PVSS) and the CPC wizard as a dedicated plug-in created to provide the user a friendly GUI (Graphical User Interface). A tool called UAB Bootstrap will manage the different UAB components, like CPC, and its dependencies with the resource packages. This tool guides the control system developer during the installation, update and execution of the UAB components. (authors)

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

    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...

  8. Comparison of different variants of ATWS type accident calculations by means of the ATHLET and RELAP codes

    The results of thermal hydraulic analyses of anticipated transients without scram (ATWS) served as the basis for the new Emergency Operating Procedures for WWER-440/V-213 reactors. Because of the differences in the behavior of parameters in the calculations by the ATHLET code (for the Dukovany NPP) and by the RELAP code (for the Bohunice V2 plant), the major parameters in selected calculations were compared and the differences were explained on graphs. The starting calculations, in which no operator intervention was taken into account, were used for the comparison. (P.A.)

  9. Study of application of protective measures for the public and remediation of contaminated areas in case of nuclear and / or radiological accidents in Brazil

    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 137Cs, 90Sr 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 real people

  10. 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

    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

  11. 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

    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.

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

    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)

  13. Utilisation of gas pipelines - Application of new codes

    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.

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

    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

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

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

    2015-01-15

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

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

    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

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

    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)

  18. Application of advanced statistical methods in assessment of the late phase of a nuclear accident

    The paper presents a new methodology for improving of estimates of radiological situation on terrain in the late phase of a nuclear accident. Methods of Bayesian filtering are applied to the problem. The estimates are based on combination of modeled and measured data provided by responsible authorities. Exploiting information on uncertainty of both the data sources, we are able to produce improved estimate of the true situation on terrain. We also attempt to account for model error, which is unknown and plays crucial role in accuracy of the estimates. The main contribution of this paper is application of an approach based on advanced statistical methods, which allows for estimating of model error covariance structure upon measurements. Model error is estimated on basis of measured-minus-observed residuals evaluated upon measured and modeled values. The methodology is demonstrated on a sample scenario with simulated measurements. (authors)

  19. Application of bulk material commodity code in nuclear engineering

    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)

  20. ICARE/CATHARE V1. Application to a PWR 900 MWe severe accident sequence

    Zabiego, Magali; Fichot, Florian; Guillard, Valia; Barrachin, Marc; Melis, Stephane; Chatelard, Patrick; Camous, Francine [Commissariat a l' Energie Atomique, Institut de Protection et de Surete Nucleaire, DRS/SEMAR/LECTA, Centre d' Etudes de Cadarache, St-Paul-Les-Durance Cedex (France); Lefevre, Bertrand [Communications et Systems, Centre d' Etudes de Cadarache, St-Paul-Lez-Durance (France)

    2000-11-01

    In the first part of this paper, a brief description of the V1 version of the ICARE/CATHARE software is presented. The new models developed by IPSN are described here. An application of this version to a French PWR 900 MWe is then shown. Although a description of the whole circuit is possible with ICARE/CATHARE V1, the present application is restricted to the reactor vessel. The results show the progression of the core degradation and the relocation of an important amount of corium in the reactor lower head. This calculation demonstrates the capacity of the code to provide a physically grounded simulation of the whole scenario. (author)

  1. BUSH: A computer code for calculating steady state heat transfer in LWR rod bundles under accident conditions

    The computer code BUSH has been developed for the calculation of steady state heat transfer in a rod bundle. For a given power, flow and geometry it can calculate the temperatures in the rods, coolant and shroud assuming that at any axial level each rod can be described by one temperature and the coolant fluid is also radially uniform at this level. Heat transfer by convection and radiation are handled and the geometry is flexible enough to model nearly all types of envisaged shroud design for the SUPERSARA test series. The modular way in which BUSH has been written makes it suitable for future development, either within the present BUSH framework or as part of a more advanced code

  2. Validation of computer code THYNAC for analysis of loss of coolant accident in pressurised heavy water reactor

    The computer code THYNAC has been validated against the available experimental results of blowdown from RD-4 loop, a Canadian test facility designed to simulate LOCA in PHWR. In this paper available experimental results are compared with predictions made with THYNAC. In general, predictions show consistent trends on pressure transient during LOCA and conservative trends with respect to fuel sheath peak temperatures. (author). 4 refs., 5 figs., 2 tab s

  3. German (GRS) approach to accident analysis (part III). Status of uncertainty evaluations of thermal-hydraulic code results in Germany

    There is an increasing interest in computational reactor safety analysis to replace the conservative evaluation model calculations by best estimate calculations supplemented by a quantitative uncertainty analysis. Sources of uncertainties - code models, initial and boundary conditions, plant state, fuel parameters, scaling, and numerical solution algorithm. Measurements, which are the basis of computer code model, show a scatter around a mean value. For example, data for two-phase pressure drop show a scatter range of about ± 20 - 30%. A range of values should be taken into account for the respective model parameter instead of one discrete value only. The state of knowledge about all uncertain parameters is described by ranges and subjective probability distributions. Stochastic variability due to possible component failures of the reactor plant is not considered in an uncertainty analysis. The single failure criterion is taken into account in a deterministic way. The aim of the uncertainty analysis is at first to identify and quantify all potentially important uncertain parameters. Their propagation through computer code calculations provides subjective probability distributions (and ranges) for the code results. The evaluation of the margin to acceptance criteria, (= technical limit value) e.g. the maximum fuel rod clad temperature, should be based on the upper limit of this distribution for the calculated temperatures. Investigations are underway to transform data measured in experiments and post-test calculations into thermal-hydraulic model parameters with uncertainties. It is effective to concentrate on those uncertainties showing the highest sensitivity measures. The state of knowledge about these uncertain input parameters has to be improved, and suitable experimental as well as analytical information has to be selected. This is a general experience applying different uncertainty methods

  4. Application of natural adsorbents as decontamination agents for the elimination of the consequences of the Chernobyl reactor accident

    The scientific foundations of using natural adsorbents as ion exchangers,filtering media and adagulants for water purification ase presented. The results showing the efficiency of practical application of natural adsorbents for the decontamination of water, clothes, machinery, construction materials, etc. during the elimination of the consequences of the Chernobyl reactor accident in 1986-1987 are presented

  5. Simulations of PWR spray systems by ASTEC computer code

    In this paper, sequence of loss of feedwater of steam generators on a PWR 900 MWe is performed by application of integral code ASTEC. The influence of the spray on evolution and source term of severe accident in the containment and in the environment is mainly studied. The results are helpful for the investigation of mitigative measures of severe accident. (authors)

  6. WSPEEDI (worldwide version of SPEEDI): A computer code system for the prediction of radiological impacts on Japanese due to a nuclear accident in foreign countries

    A computer code system has been developed for near real-time dose assessment during radiological emergencies. The system WSPEEDI, the worldwide version of SPEEDI (System for Prediction of Environmental Emergency Dose Information) aims at predicting the radiological impact on Japanese due to a nuclear accident in foreign countries. WSPEEDI consists of a mass-consistent wind model WSYNOP for large-scale wind fields and a particle random walk model GEARN for atmospheric dispersion and dry and wet deposition of radioactivity. The models are integrated into a computer code system together with a system control software, worldwide geographic database, meteorological data processor and graphic software. The performance of the models has been evaluated using the Chernobyl case with reliable source terms, well-established meteorological data and a comprehensive monitoring database. Furthermore, the response of the system has been examined by near real-time simulations of the European Tracer Experiment (ETEX), carried out over about 2,000 km area in Europe. (author)

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

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

  8. Thermal-hydraulic analysis best-estimate of an accident in the containment a PWR-W reactor with GOTHIC code using a 3D model detailed; Analisis termo-hidraulico best-estimate de un accidente en contencion de un reactor PWR-W con el codigo GOTHIC mediante un modelo 3D detallado

    Bocanegra, R.; Jimenez, G.

    2013-07-01

    The objective of this project will be a model of containment PWR-W with the GOTHIC code that allows analyzing the behavior detailed after a design basis accident or a severe accident. Unlike the models normally used in codes of this type, the analysis will take place using a three-dimensional model of the containment, being this much more accurate.

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

    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

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

    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

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

    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...

  12. The combined thermohydraulics-neutronics code TRAB-SMABRE for 3D plant transient and accident analyses

    TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenized fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensable part in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but the rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation. (authors)

  13. Accidents - Chernobyl accident

    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.)

  14. 26 CFR 521.102 - Applicable provisions of the Internal Revenue Code.

    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....

  15. Charged and neutral particle transport methods and applications: The CALOR code system

    Gabriel, T.A.; Charlton, L.A.

    1997-04-01

    The CALOR code system, which is a complete radiation transport code system, is described with emphasis on the high-energy (> 20 MeV) nuclear collision models. Codes similar to CALOR are also briefly discussed. A current application using CALOR which deals with the development of the National Spallation Neutron Source is also given.

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

    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)

  17. Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

    Bae, Byoung-Uhn, E-mail: bubae@kaeri.re.kr; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

    2014-08-15

    Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection.

  18. Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

    Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection

  19. The application of bar coding technology at WIPP

    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

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

    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.

  1. User Effect on Code Application and Qualification Needs

    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

  2. Arithmetic error codes - Cost and effectiveness studies for application in digital system design.

    Avizienis, A.

    1971-01-01

    The application of error-detecting or error-correcting codes in digital computer design requires studies of cost and effectiveness tradeoffs to supplement the knowledge of their theoretical properties. General criteria for cost and effectiveness studies of error codes are developed, and results are presented for arithmetic error codes with the low-cost check modulus 2 super a - 1. Both separate (residue) and nonseparate (AN) codes are considered. The class of multiple arithmetic error codes is developed as an extension of low-cost single codes.

  3. QR CODE GENERATION AND APPLICATION OF ANTI-FALSIFICATION

    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...

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

    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.

  5. Description and application of the AERIN Code at LLNL

    The AERIN code was written at the Lawrence Livermore National Laboratory in 1976 to compute the organ burdens and absorbed dose resulting from a chronic or acute inhalation of transuranic isotopes. The code was revised in 1982 to reflect the concepts of ICRP-30. This paper will describe the AERIN code and how it has been used at LLNL to study more than 80 cases of internal deposition and obtain estimates of internal dose. A comparison with the computed values of the committed organ dose is made with ICRP-30 values. The benefits of using the code are described. 3 refs., 3 figs., 6 tabs

  6. Severe accident analysis methodology in support of accident management

    The author addresses the implementation at BELGATOM of a generic severe accident analysis methodology, which is intended to support strategic decisions and to provide quantitative information in support of severe accident management. The analysis methodology is based on a combination of severe accident code calculations, generic phenomenological information (experimental evidence from various test facilities regarding issues beyond present code capabilities) and detailed plant-specific technical information

  7. [Application of the new coding standards for introducing diagnosis related groups (DRGs) in ophthalmology].

    Stiefelmeyer, Sandra; Neubauer, Aljoscha Steffen; Ehrt, Oliver; Kampik, Anselm

    2002-08-01

    For all in-patients a fundamental change of hospital financing based only on diagnosis related groups (DRGs) is planned in Germany for 2004. Patients are grouped into different DRGs according to their coded diagnoses, operations and procedures. Hence, the quality of coding is essential for adequate reimbursement of cases and thus for hospital financing. The legal framework of coding is supplied by the so-called "general and special coding guidelines for coding diseases and procedures", presented jointly by the associations representing hospitals and public and private health insurance companies in early 2002. These German coding standards regulate how coding must be performed in daily practice and therefore have a major influence on the grouping of patients into DRGs. Because of their importance this paper describes the practical application of current coding standards. Specifically in Ophthalmology the problem exists that for many procedures adequate coding numbers are still missing, thus impairing further development of the DRG system. PMID:12353175

  8. Two Application Examples of Concrete Containment Structures under Accident Load Conditions Using Finite Element Analysis

    Oskarshamn 1 is the oldest nuclear plant in operation in Sweden. It was designed and erected at the end of the 1960's. During the last five years an extensive upgrading process of the power plant has been carried out. Within the frame of one of these upgrading projects the outer and inner Main Steam Isolation Valves (MSN) have been replaced. As a consequence of these replacements it was necessary to make an overhaul investigation of the basic general concept regarding pipe rupture descriptions and rupture locations, in order to attain a design of the pipe whip restraints in accordance with requirements of modern standards. In this paper two application examples regarding finite element analysis of concrete containment structures under accident loading conditions are presented. Each example includes a brief introduction of the problem and the object of the commission. The finite element model and the structural response analysis are described and the results are discussed. The application examples are: 1. Non-linear structural analysis of a reinforced concrete culvert affected by internal over-pressurization and impulse load effects of pipe rupture reactions. 2. Non-linear thermal stress analysis around a steel penetration of a reactor containment

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

    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.

  10. A Non-MDS Erasure Code Scheme For Storage Applications

    Kiani, Abbas

    2011-01-01

    This paper investigates the use of redundancy and self repairing against node failures in distributed storage systems, using various strategies. In replication method, access to one replication node is sufficient to reconstruct a lost node, while in MDS erasure coded systems which are optimal in terms of redundancy-reliability tradeoff, a single node failure is repaired after recovering the entire stored data. Moreover, regenerating codes yield a tradeoff curve between storage capacity and repair bandwidth. The current paper aims at investigating a new storage code. Specifically, we propose a non-MDS (2k, k) code that tolerates any three node failures and more importantly, it is shown using our code a single node failure can be repaired through access to only three nodes.

  11. ULOF and unprotected blockage accident analyses of the 400MWth-class EFIT accelerator driven transmuter with the SIMMER-III code

    In the EUROTRANS Programme of the 6th FP of the EC the EFIT, the European Facility for Industrial Transmutation, a generic conceptual design of a full 400 MWth ADS transmuter loaded with a CERCER U-free fuel with an MgO matrix, is developed. The safety objectives for the EFIT are achieved on the basis of the defense-in-depth concept. For the safety assessment various protected and unprotected design basis conditions (DBC) and design extension conditions (DEC) transients have been analyzed as e.g. the protected and unprotected loss of flow (PLOF/ULOF), beam trip transients, over power transients and especially the unprotected blockage accidents (UBAs). Key safety analyses of the EFIT core have been performed with the SIMMER-III code, currently the only code which can handle core disruptive conditions for heavy liquid metal reactors. Unprotected transients set upper safety limits and play an important role in the overall safety assessment and transient behaviors of MgO based fuel and T91 cladding. Especially the high temperature range is still connected with significant uncertainties, therefore, unprotected accidents with a potential of fuel failure and gas release deserve special attention. In this paper, analyses of the ULOF and the UBA will be presented. Simulations of the core under the steady operation state have been carried out first, whose results show good agreement with the design data. In addition the safety parameters as e.g. the void worth of the core and the Doppler constant have been determined. In the ULOF analysis, the pump head is assumed to be completely lost within 1 or 5 seconds. Due to the loss of the pump head, coolant mass flow rate in the reactor will decrease and finally arrive at its new steady state because of the remained natural convection. Calculated results indicate that, under the current assumptions, the EFIT design can survive a ULOF without fuel pin failure whilst a larger safety margin exists if the pump halving time is longer

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

    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 ...

  13. Developing and validating severe accident management guidelines using SAMPSON-RELAP/SCDAPSIM.MOD3.4

    The development and validation of Severe Accident Management Guidelines (SAMGs) must consider complex thermal-hydraulic and severe accident phenomena. Yet, many of the simplified integral Severe Accident codes, that have been used widely to develop SAMGs in Europe, Asia, and the United States, cannot accurately predict many of these complex interactions. By contrast, detailed codes such as SAMPSON-RELAP/SCDAPSIM have shown, through comparison with the TMI-2 accident and experiments, that they can predict such complex behavior. This paper describes the merger of SAMPSON with RELAP/SCDAPSIM/MOD3.4, reviews the severe accident phenomena important for Severe Accident Management, and then describes the potential impact of using SAMPSON-RELAP/SCDAPSIM on the development and validation of SAMGs. A companion paper, being presented at this conference provides an example of the application of SAMPSON-RELAP/SCDAPSIM for the development and validation of a SAMG for a Nuclear Power Plant. (authors)

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

    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.

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

    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)

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

    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.

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

    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

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

    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)

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

    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

  20. Experimental Analysis with RANNS Code on Boiling Heat Transfer from Fuel Rod Surface to Coolant Water Under Reactivity-Initiated Accident Conditions

    In order to promote a better understanding of the temperature evolution of fuel rod under reactivity-initiated accident (RIA) conditions, we have investigated the effects of coolant subcooling, flow velocity, pressure, and cladding pre-irradiation on the heat transfer from fuel rod surface to coolant water during RIA boiling transient. The study was based on a computational analysis, with the RANNS code, on the transient data from RIA-simulating experiments in the nuclear safety research reactor (NSRR); boiling heat transfer coefficients were estimated by inverse-heat-conduction calculations using the histories of measured cladding temperature and estimated heat generation in pellets, and the effects of coolant condition were analyzed by a two-phase laminar boundary layer model for stable film boiling. The experimental data used in this study cover coolant conditions with subcoolings of ~10–80 K, flow velocities of 0 to ~3 m/s, pressures of 0.1 to ~16 MPa, and fuel burnups of 0–69 GWd/tU. The analysis showed that the film boiling heat transfer coefficients during RIA boiling transient increase with coolant subcooling, flow velocity, and pressure as predicted by the model for stable film boiling. The estimated boiling heat transfer coefficients were significantly larger than those predicted by semi-empirical correlations for stable film boiling: about 1.5 times larger for stagnant water condition and 2–8 times larger for forced flow condition, respectively. The analysis also suggested that the heat transfers during both transition and film boiling phases are strongly enhanced by pre-irradiation of the cladding. The irradiation effect was clearly seen at large subcooling of ~80 K and atmospheric coolant pressure, and was rather moderate at small subcooling of ~10 K and coolant pressure of ~7 MPa. These behaviors of boiling heat transfer are incorporated into the RANNS code mainly as modified empirical correlations for boiling heat transfer coefficient. (author)

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

    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

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

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

    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.

  4. Quantifying reactor safety margins: Part 1: An overview of the code scaling, applicability, and uncertainty evaluation methodology

    In August 1988, the Nuclear Regulatory Commission (NRC) approved the final version of a revised rule on the acceptance of emergency core cooling systems (ECCS) entitled ''Emergency Core Cooling System; Revisions to Acceptance Criteria.'' The revised rule states an alternate ECCS performance analysis, based on best-estimate methods, may be used to provide more realistic estimates of plant safety margins, provided the licensee quantifies the uncertainty of the estimates and included that uncertainty when comparing the calculated results with prescribed acceptance limits. To support the revised ECCS rule, the NRC and its contractors and consultants have developed and demonstrated a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology. It is an auditable, traceable, and practical method for combining quantitative analyses and expert opinions to arrive at computed values of uncertainty. This paper provides an overview of the CSAU evaluation methodology and its application to a postulated cold-leg, large-break loss-of-coolant accident in a Westinghouse four-loop pressurized water reactor with 17 /times/ 17 fuel. The code selected for this demonstration of the CSAU methodology was TRAC-PF1/MOD1, Version 14.3. 23 refs., 5 figs., 1 tab

  5. Coldleg Loss of Coolant Accident (LOCA) Analysis of the Modified Reactor Thermohydraulic Test Facility Using CATHENA Computer Code

    A LOCA analysis at coldleg of the reactor thermal hydraulic test facility using CATHENA computer code has been completely conducted. The analysis was performed by modeling the reactor thermal hydraulic test loop into generic models of the CATHENA such as PUMP, VALVE, VOLUME, ACCUMULATOR, TANK, RESERVOIR, DISCHARGE, GENERALIZED TANK, and GENHTP. The primary system was simulated at power of 1 MWatt with pressure and mass flow at 15.5 MPa and 9.4 kg/s respectively. At secondary side, feedwater flowed at 15.0 kg/s with temperature of 27 oC and pressure of 0.8 MPa. The calculation showed that during steady state the inlet and outlet temperature of test section were 121 oC and 146 oC. After calculating steady state condition, the calculation was followed by transient calculation. The transient was triggered by pipe break at coldleg with diameter of the break was 2 mm. Due to this break, the pressure decreased dramatically. When the pressure reached 4.2 MPa, the accumulator started supplying water into the system. A moment later, the pump was also tripped because of the continuing pressure drop that reached 4.0 MPa. As a consequence the coolant flow was also dropped At the coolant 40 % of normal flow, the power of heated rods then shut down. The result of calculation showed that during the transient, the maximum coolant temperature was 159 oC and the maximum temperature of heated rods was 223 oC. Based on these results, it can be concluded that during the transient, the heated rods were not in danger. (author)

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

    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. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was conservatively calculated

  8. Evolution of the ELESTRES code for application to extended burnups

    The computer code ELESTRES is frequently used at Atomic Energy of Canada Limited to assess the integrity of CANDU fuel under normal operating conditions. The code also provides initial conditions for evaluating fuel behaviour during high-temperature transients. This paper describes recent improvements in the code in the areas of pellet expansion and of fission gas release. Both of these are very important considerations in ensuring fuel integrity at extended burnups. Firstly, in calculations of pellet expansion, the code now accounts for the effect of thermal stresses on the volume of gas bubbles at the boundaries of UO2 grains. This has a major influence on the expansion of the pellet during power-ramps. Secondly, comparisons with data showed that the previous fission gas package significantly underpredicted the fission gas release at high burnups. This package has now been improved via modifications to the following modules: distance between neighbouring bubbles on grain boundaries; diffusivity; and thermal conductivity. The predictions of the revised version of the code show reasonable agreement with measurements of ridge strains and of fission gas release. An illustrative example demonstrates that the code can be used to identify a fuel design that would: reduce the sheath stresses at circumferential ridges by a factor of 2-10; and keep the gas pressure at very high burnups to below the coolant pressure

  9. The Genomic Code: Genome Evolution and Potential Applications

    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. Description of the COMRADEX code

    The COMRADEX Code is discussed briefly and instructions are provided for the use of the code. The subject code was developed for calculating doses from hypothetical power reactor accidents. It permits the user to analyze four successive levels of containment with time-varying leak rates. Filtration, cleanup, fallout and plateout in each containment shell can also be analyzed. The doses calculated include the direct gamma dose from the containment building, the internal doses to as many as 14 organs including the thyroid, bone, lung, etc. from inhaling the contaminated air, and the external gamma doses from the cloud. While further improvements are needed, such as a provision for calculating doses from fallout, rainout and washout, the present code capabilities have a wide range of applicability for reactor accident analysis

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

    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. Modeling of the thermal transfer inside a porous environment: application to nuclear reactors in accident situation

    The purpose of this report is to simulate heat exchanges occurring by conduction, by convection and by radiating in a porous medium made up of opaque particles in a semi-transparent fluid. Usually the determination of the macroscopic equations is based on homogenization techniques, but in the case of a major accident, the complexity of the problem is so overwhelming that semi-empirical methods are used to determine macroscopic coefficients. The author develops a new method to determine these coefficients, this method is based on the calculation of different tensors: the equivalent conductivity tensor, the radiative conductivity tensor, the thermal conductivity tensor and the heat exchange coefficient (hsf) between the solid phase and the fluid one. The first chapter briefly describes energy, impulse and mass balances. In the case of the energy balance the solid phase is not supposed to be in thermal equilibrium with the liquid phase. The second chapter presents an application of the porous media method to a one-dimensional and stationary problem, this application to a simple problem gives an idea of the performance of the method. The model allowing the calculation of hsf is developed, it is a wide range model. The second chapter ends with the presentation of the model allowing the computing of the effective conductivity of fuel rods. A comparison between results given by this new method and other numeric calculations or experimental data coming from benchmarks is presented in the third chapter. This chapter ends with the simulation of a reactor core in accidental situation, 2 cases are presented: with and without the presence of water steam. (A.C.)

  13. Calculation of accidents at the 14.7 m level in the Dukovany WWER-440 NPP by using the MELCOR 1.8.2 code. (Extended version of report)

    Calculations of pipe rupture accidents in the steam pipeline room at the +14.7 m level in the Dukovany nuclear power plant (WWER-440/V-213) were performed on a Hewlett-Packard 9000/720 computer by using the MELCOR 1.8.2 code. The calculations were aimed at determining the pressure, temperature, humidity and condensed water level for the following types of accident: (i) main steam header rupture; (ii) steam line rupture before the quick-acting valve; and (iii) main feedwater header rupture. The temperature and humidity in the steam pipeline room emerged as the limiting parameters in the qualification assessment of the safety-related equipment. Pressure increase following the accident will not be very marked and water spilled on the floor should not reach a hazardous level. 29 figs., 9 tabs., 18 refs

  14. 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

    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)

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

    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

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

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

  17. Researches and Applications of ESR Dosimetry for Radiation Accident Dose Assessment

    The aim of this work was to establish methods suitable for practical dose assessment of people involved in ionising radiation accidents. Some biological materials of the human body and materials possibly carried or worn by people were taken as detection samples. By using electron spin resonance (ESR) techniques, the basic dosimetric properties of selected materials were investigated in the range above the threshold dose of human acute haemopoietic radiation syndrome. The dosimetric properties involved included dose response properties of ESR signals, signal stabilities, distribution of background signals, the lowest detectable dose value, radiation conditions, environmental effects on the detecting process, etc. Several practical dose analytical indexes and detecting methods were set up. Some of them (bone, watch glass and tooth enamel) had also been successfully used in the dose assessment of people involved in three radiation accidents, including the Chernobyl reactor accident. This work further proves the important role of ESR techniques in radiation accident dose estimation. (author)

  18. Application of probabilistic methods to accident analysis at waste management facilities

    Probabilistic risk assessment is a technique used to systematically analyze complex technical systems, such as nuclear waste management facilities, in order to identify and measure their public health, environmental, and economic risks. Probabilistic techniques have been utilized at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, to evaluate the probability of a catastrophic waste hoist accident. A probability model was developed to represent the hoisting system, and fault trees were constructed to identify potential sequences of events that could result in a hoist accident. Quantification of the fault trees using statistics compiled by the Mine Safety and Health Administration (MSHA) indicated that the annual probability of a catastrophic hoist accident at WIPP is less than one in 60 million. This result allowed classification of a catastrophic hoist accident as ''not credible'' at WIPP per DOE definition. Potential uses of probabilistic techniques at other waste management facilities are discussed

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

    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.

  20. Traitement automatique de rapports d’incidents et accidents : application à la gestion du risque dans l’aviation civile

    Tulechki, Nikola

    2015-01-01

    This thesis describes the applications of natural language processing (NLP) to industrial risk management. We focus on the domain of civil aviation, where incident reporting and accident investigations produce vast amounts of information, mostly in the form of textual accounts of abnormal events, and where efficient access to the information contained in the reports is required. We start by drawing a panorama of the different types of data produced in this particular domain. We analyse the do...

  1. Development of Database for Accident Analysis in Indian Mines

    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.

  2. DWroidDump: Executable Code Extraction from Android Applications for Malware Analysis

    Dongwoo Kim; Jin Kwak; Jaecheol Ryou

    2015-01-01

    We suggest an idea to dump executable code from memory for malicious application analysis on Android platform. Malicious applications are getting enhanced in terms of antianalysis techniques. Recently, sophisticated malicious applications have been found, which are not decompiled and debugged by existing analysis tools. It becomes serious threat to services related to embedded devices based on Android. Thus, we have implemented the idea to obtain main code from the memory by modifying a part ...

  3. Review of the GOTHIC code and trial application

    A critical review of the performance of the generic computer code GOTHIC for the generation of thermalhydraulic information for containments was conducted. Several analyses were performed with GOTHIC to predict the flow behaviour and distribution of hydrogen concentration within containments whose geometrical complexity ranged from two simple interconnected rooms to a full scale reactor building. Sensitivity analysis studies were carried out to examine the effect of various modeling parameters. The implementation of physics by the code is reviewed and recommendations on its use for performing blowdown/hydrogen release analyses are made.(author) 5 refs., 9 tabs., 105 figs

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

    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

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

    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

  6. 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.

    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...

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

    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

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

    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...

  9. Application of multiple-error-correcting binary BCH codes to optical matrix-vector multipliers

    Caglar, A. Tankut; Krile, Thomas F.; Walkup, John F.

    1996-06-01

    The application of multiple error-correction codes to optical matrix-vector multipliers (OMVMs) can improve the computational accuracy level of these processors. A binary Bose Ray-Chaudhuri (BCH) code was applied to a simulated mod-2 OMVM. Based on the results obtained from the simulations, the conditions under which the use of such error-correction coding is feasible in OMVMs are discussed.

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

    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...

  11. Python-Assisted MODFLOW Application and Code Development

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

    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

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

    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)

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

    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)

  15. Application of software quality assurance to a specific scientific code development task

    This paper describes an application of software quality assurance to a specific scientific code development program. The software quality assurance program consists of three major components: administrative control, configuration management, and user documentation. The program attempts to be consistent with existing local traditions of scientific code development while at the same time providing a controlled process of development

  16. Bridging Inter-flow and Intra-flow Network Coding for Video Applications

    Hansen, Jonas; Krigslund, Jeppe; Roetter, Daniel Enrique Lucani;

    2013-01-01

    transmission approach to decide how much and when to send redundancy in the network, and a minimalistic feedback mechanism to guarantee delivery of generations of the different flows. Given the delay constraints of video applications, we proposed a simple yet effective coding mechanism, Block Coding On The Fly...

  17. Cesar: a simplified evolution code for reprocessing applications

    In the framework of a collaboration between CEA and COGEMA, CESAR has been developed to provide the required characterization data for burn up fuels from PWRs. BWRs and FRs. It can quickly calculate the evolution, in and out of pile, of material balances, and the activity, decay heat and neutron source emitted by the irradiated fuel, taking into account 104 actinides, 208 fission products and 125 activation products. CESAR can also make flux depletion calculations of radioactive sources, up to geological times (106 years). The neutronics data libraries (cross sections sets) are supplied by the CEA reference calculation codes for neutron physics: APOLLO for thermal spectrum systems and ERANOS for fast spectrum systems. In the near future, they will be supplied by the DARWIN package. For the principal nuclides, the code has been validated using two types of experimental results: isotopic analysis from burn up fuel rod samples and also from full assemblies dissolutions from the reprocessing plants. (author)

  18. Application of the FLIRA code to fit ERSEC experiments

    The FLIRA code was developped to calculate the maximum cladding temperature reached in the course of an accidental LOCA during reflooding of a water reactor. This code calculates the quench front progress. Its aim is to establish a model describing all the elementary mechanism involved in the reflooding phenomenon. This paper is devoted to the qualification of the FLIRA thermohydraulical model with help of ERSEC experiments performed with a tube (bottom reflooding). The model takes in account the seven following flow patterns. Four flow patterns occur when the wall is in contact with the liquid phase, upstream from the quenching front. And three flow patterns occur when the wall is in contact with the vapor phase, downstream from the quenching front

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

    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)

  20. Application of CFD Code PHOENICS for simulating CYCLONE SEPARATORS

    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