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Sample records for accident analysis code

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

    Directory of Open Access Journals (Sweden)

    Muhammad Hashim

    2013-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-03-15

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

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

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    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-15

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

  6. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-15

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and err000.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  14. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-02-01

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

  15. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

    Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Jow, H.N. (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 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.

  16. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-02-01

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

  17. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  20. FGR Evaluation Code for CANDU Fuel under an Accidents: REDOU

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-12-15

    When an end fitting failure is occurred, the sheath is broken and the fission gases are released promptly from the gap between the pellet and sheath and also released continuously from the inside of the pellet due to the oxidation of the pellet. Thus, the accurate calculation of the fission gas release from the gap and from the inside of the pellet is an essential to the accident analysis of the fuel behavior and to prepare the safety strategy. In this report, among the performance analysis or the transient behavior prediction computer codes, the REDOU code which is the fission gas calculating code for the postulated accident scenario such as an end fitting failure is introduced. Then, the user manual for REDOU code is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and error.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  2. Severe accident analysis using dynamic accident progression event trees

    Science.gov (United States)

    Hakobyan, Aram P.

    In present, the development and analysis of Accident Progression Event Trees (APETs) are performed in a manner that is computationally time consuming, difficult to reproduce and also can be phenomenologically inconsistent. One of the principal deficiencies lies in the static nature of conventional APETs. In the conventional event tree techniques, the sequence of events is pre-determined in a fixed order based on the expert judgments. The main objective of this PhD dissertation was to develop a software tool (ADAPT) for automated APET generation using the concept of dynamic event trees. As implied by the name, in dynamic event trees the order and timing of events are determined by the progression of the accident. The tool determines the branching times from a severe accident analysis code based on user specified criteria for branching. It assigns user specified probabilities to every branch, tracks the total branch probability, and truncates branches based on the given pruning/truncation rules to avoid an unmanageable number of scenarios. The function of a dynamic APET developed includes prediction of the conditions, timing, and location of containment failure or bypass leading to the release of radioactive material, and calculation of probabilities of those failures. Thus, scenarios that can potentially lead to early containment failure or bypass, such as through accident induced failure of steam generator tubes, are of particular interest. Also, the work is focused on treatment of uncertainties in severe accident phenomena such as creep rupture of major RCS components, hydrogen burn, containment failure, timing of power recovery, etc. Although the ADAPT methodology (Analysis of Dynamic Accident Progression Trees) could be applied to any severe accident analysis code, in this dissertation the approach is demonstrated by applying it to the MELCOR code [1]. A case study is presented involving station blackout with the loss of auxiliary feedwater system for a

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

    Directory of Open Access Journals (Sweden)

    Itamar Iliuk

    2016-01-01

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

  4. Test Data for USEPR Severe Accident Code Validation

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe

    2007-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  6. MELCOR analysis of the TMI-2 accident

    Energy Technology Data Exchange (ETDEWEB)

    Boucheron, E.A.

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01

    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. Accident investigation and analysis

    NARCIS (Netherlands)

    Kampen, J. van; Drupsteen, L.

    2013-01-01

    Many organisations and companies take extensive proactive measures to identify, evaluate and reduce occupational risks. However, despite these efforts things still go wrong and unintended events occur. After a major incident or accident, conducting an accident investigation is generally the next ste

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-06-01

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

  12. Accident tolerant fuel analysis

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis [Idaho National Laboratory; Chichester, Heather [Idaho National Laboratory; Johns, Jesse [Texas A& M University; Teague, Melissa [Idaho National Laboratory; Tonks, Michael Idaho National Laboratory; Youngblood, Robert [Idaho National Laboratory

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced ''RISMC toolkit'' that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional ''accident-tolerant'' (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant

  13. Accident Tolerant Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; Heather Chichester; Jesse Johns; Melissa Teague; Michael Tonks; Robert Youngblood

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and

  14. Study on the code system for the off-site consequences assessment of severe nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

  15. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Ronaldo Celem

    2001-10-15

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

  18. Radioactive materials transport accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    McSweeney, T.I.; Maheras, S.J.; Ross, S.B. [Battelle Memorial Inst. (United States)

    2004-07-01

    Over the last 25 years, one of the major issues raised regarding radioactive material transportation has been the risk of severe accidents. While numerous studies have shown that traffic fatalities dominate the risk, modeling the risk of severe accidents has remained one of the most difficult analysis problems. This paper will show how models that were developed for nuclear spent fuel transport accident analysis can be adopted to obtain estimates of release fractions for other types of radioactive material such as vitrified highlevel radioactive waste. The paper will also show how some experimental results from fire experiments involving low level waste packaging can be used in modeling transport accident analysis with this waste form. The results of the analysis enable an analyst to clearly show the differences in the release fractions as a function of accident severity. The paper will also show that by placing the data in a database such as ACCESS trademark, it is possible to obtain risk measures for transporting the waste forms along proposed routes from the generator site to potential final disposal sites.

  19. OPR1000 Control Rod Drop Accident Simulation using the SPACE Code

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chang Keun; Ha, Sang Jun; Moon, Chan Kook [Korea Hydro and Nuclear Power, Daejeon (Korea, Republic of)

    2012-05-15

    The Korea nuclear industry has developed a best estimated two-phase three-filed thermal-hydraulic analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plants), for safety analysis and design of a PWR (Pressurized Water Reactor). As the first phase, the demo version of the SPACE code was released in March 2010. The code has been verified and improved according to the Validation and Verification (V and V) matrix prepared for the SPACE code as the second phase of the development. In this study, a Control Rod Drop accident has been simulated using the SPACE code as one aspect of the V and V work. The results from this test were compared with tests of the RETRAN and CESEC codes

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-15

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

  1. Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code; Verbundprojekt WASA-BOSS: Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt B: Druckwasserreaktor-Stoerfallanalysen unter Verwendung des Severe-Accident-Codes ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

    2017-02-15

    Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

  2. MELCOR DB Construction for the Severe Accident Analysis DB

    Energy Technology Data Exchange (ETDEWEB)

    Song, Y. M.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    The Korea Atomic Energy Research Institute (KAERI) has been constructing a severe accident analysis database (DB) under a National Nuclear R and D Program. In particular, an MAAP (commercial code being widely used in industries for integrated severe accident analysis) DB for many scenarios including a station blackout (SBO) has been completed. This paper shows the MELCOR DB construction process with examples of SBO scenarios, and the results will be used for a comparison with the MAAP DB

  3. Practical approaches in accident analysis

    Science.gov (United States)

    Stock, M.

    An accident analysis technique based on successive application of structural response, explosion dynamics, gas cloud formation, and plant operation failure mode models is proposed. The method takes into account the nonideal explosion characteristic of a deflagration in the unconfined cloud. The resulting pressure wave differs significantly from a shock wave and the response of structures like lamp posts and walls can differ correspondingly. This gives a more realistic insight into explosion courses than a simple TNT-equivalent approach.

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

    Institute of Scientific and Technical Information of China (English)

    李世锐; 任丽霞; 胡文军; 乔鹏瑞

    2016-01-01

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

  5. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  6. Validation and application of the system code ATHLET-CD for BWR severe accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

    2016-10-15

    Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

  7. Development of Database for Accident Analysis in Indian Mines

    Science.gov (United States)

    Tripathy, Debi Prasad; Guru Raghavendra Reddy, K.

    2016-10-01

    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.

  8. Development of Database for Accident Analysis in Indian Mines

    Science.gov (United States)

    Tripathy, Debi Prasad; Guru Raghavendra Reddy, K.

    2015-08-01

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

  9. Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

    Energy Technology Data Exchange (ETDEWEB)

    Lázaro, A., E-mail: aurelio.lazaro-chueca@ec.europa.eu [JRC-IET European Commission—Westerduinweg 3, PO Box-2, 1755 ZG Petten (Netherlands); UPV—Universidad Politecnica de Valencia, Cami de vera s/n-46002, Valencia (Spain); 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.; Massara, S. [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); Mikityuk, K. [PSI—Paul Scherrer Institut, 5232 Villigen Switzerland (Switzerland); Schikorr, M.; Bubelis, E.; Ponomarev, A.; Kruessmann, R. [KIT—Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen Germany (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, PO Box 9034 6800 ES, Arnhem (Netherlands)

    2014-01-15

    Highlights: • Ten system-code models of the ESFR were developed in the frame of the CP-ESFR project. • Eight different thermohydraulic system codes adapted to sodium fast reactor's technology. • Benchmarking exercise settled to check the consistency of the calculations. • Upgraded system codes able to simulate the reactivity feedback and key safety parameters. -- Abstract: The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.

  10. Analysis of local subassembly accident in KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min; Jeong, Kwan Seong; Hahn, Do Hee

    2000-10-01

    Subassembly Accidents (S-A) in the Liquid Metal Reactor (LMR) may cause extensive clad and fuel melting and are thus regarded as a potential whole core accident initiator. The possibility of S-A occurrence must be very low frequency by the design features, and reactor must have specific instrumentation to interrupt the S-A sequences by causing a reactor shutdown. The evaluation of the relevant initiators, the event sequences which follow them, and their detection are the essence of the safety issue. Particularly, the phenomena of flow blockage caused by foreign materials and/or the debris from the failed fuel pin have been researched world-widely. The foreign strategies for dealing with the S-A and the associated safety issues with experimental and theoretical R and D results are reviewed. This report aims at obtaining information to reasonably evaluate the thermal-hydraulic effect of S-A for a wire-wrapped LMR fuel pin bundle. The mechanism of blockage formation and growth within a pin bundle and at the subassembly entrance is reviewed in the phenomenological aspect. Knowledge about the recent LMR subassembly design and operation procedure to prevent flow blockage will be reflected for KALIMER design later. The blockage analysis method including computer codes and related analytical models are reviewed. Especially SABRE4 code is discussed in detail. Preliminary analyses of flow blockage within a 271-pin driver subassembly have been performed using the SABRE4 computer code. As a result no sodium boiling occurred for the central 24-subchannel blockage as well as 6-subchannel blockage.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-12-01

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

  12. The Most Effective Safety Codes on Occupational Accidents Prevention in Construction Sites

    Directory of Open Access Journals (Sweden)

    Shirazeh Arghami

    2017-03-01

    Full Text Available Background: Rapid pace of the construction projects have led to safety negligence on the part of the contractors causing higher numbers of accidents in construction industry. Thus, over 2200 legal provisions have been established for the prevention of accidents on construction sites. Not only contractors are unable to implement all provisions, but also it is hard for regulatory bodies to implement all rules, but they are also hard to remember and attend to by regulatory forces. Therefore, identifying the set of codes with the highest effect on reducing accidents would be most beneficial. The aim of this study was to determine the most effective legal provisions in reducing accidents on construction sites in Zanjan city. Methods: In this study, 378 work-related accidents in construction sites in Zanjan city, already registered in the Labor Inspection Authority, were investigated .Then the codes that had not been respected in the construction industry accidents were identified. Effective codes were determined based on their frequency on Pareto chart. Results: Results indicated that out of 2200 legal matters, only 164 were broken in the accidents. Conclusion: Application of Pareto chart could raise safety professionals' awareness of the specific broken codes in accidents, thus draw higher attention to the codes.

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

    Science.gov (United States)

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

    2014-12-01

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

  14. Analysis of Fukushima Daiichi Accident Using HFACS

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Saeed Almheiri [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    The shadow of Fukushima Daiichi nuclear power plant (NPP) accident is still too big and will last long. On the other hand, it could still teach us lots of lessons to better design and operate nuclear power plants. In this paper, we will be focusing on the Fukushima Daiichi accident, especially on human organizational factors. We will analyze the accident using Human Factors Analysis and Classification System (HFACS) in order to better understand the organizational climate of TEPCO{sup 1} and NISA{sup 2} that led to Fukushima Daiichi Accident. HFACS was developed for the U. S. aviation industry and has been used at many industries like the rail and mining industries. We found that the HFACS to be greatly beneficial in investigating the latent and organizational causes for the accident. The application results show that the causes of Fukushima Daiichi accident were spread out from sharp end (i.e. Unsafe Act) to blunt end (i. e. Organizational Influences). This means that the corresponding countermeasures should cover from front line staff to management. Thus, we managed to develop a better understanding on how to prevent similar errors or violations. The incident and near-miss have a lot of helpful information because it may show the actual and latent deficiencies of complex systems. We applied the HFACS into Fukushima Daiichi accident to better locate the causes related to both sharp and blunt ends of operation of NPP. In order to derive useful lessons from the accident analysis, the analyst should try to find the similarities not differences from the incident. It is imperative that whatever accident/incident analysis systems we use, we should fully utilize the disastrous Fukushima accident.

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

    Energy Technology Data Exchange (ETDEWEB)

    Trambauer, K. [GRS, Garching (Germany)

    1997-07-01

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

  16. Criticality accident detector coverage analysis using the Monte Carlo Method

    Energy Technology Data Exchange (ETDEWEB)

    Zino, J.F.; Okafor, K.C.

    1993-12-31

    As a result of the need for a more accurate computational methodology, the Los Alamos developed Monte Carlo code MCNP is used to show the implementation of a more advanced and accurate methodology in criticality accident detector analysis. This paper will detail the application of MCNP for the analysis of the areas of coverage of a criticality accident alarm detector located inside a concrete storage vault at the Savannah River Site. The paper will discuss; (1) the generation of fixed-source representations of various criticality fission sources (for spherical geometries); (2) the normalization of these sources to the ``minimum criticality of concern`` as defined by ANS 8.3; (3) the optimization process used to determine which source produces the lowest total detector response for a given set of conditions; and (4) the use of this minimum source for the analysis of the areas of coverage of the criticality accident alarm detector.

  17. An analysis of aircraft accidents involving fires

    Science.gov (United States)

    Lucha, G. V.; Robertson, M. A.; Schooley, F. A.

    1975-01-01

    All U. S. Air Carrier accidents between 1963 and 1974 were studied to assess the extent of total personnel and aircraft damage which occurred in accidents and in accidents involving fire. Published accident reports and NTSB investigators' factual backup files were the primary sources of data. Although it was frequently not possible to assess the relative extent of fire-caused damage versus impact damage using the available data, the study established upper and lower bounds for deaths and damage due specifically to fire. In 12 years there were 122 accidents which involved airframe fires. Eighty-seven percent of the fires occurred after impact, and fuel leakage from ruptured tanks or severed lines was the most frequently cited cause. A cost analysis was performed for 300 serious accidents, including 92 serious accidents which involved fire. Personal injury costs were outside the scope of the cost analysis, but data on personnel injury judgements as well as settlements received from the CAB are included for reference.

  18. Severe accident analysis of a small LOCA accident using MAAP-CANDU support level 2 PSA for the Point Lepreau station refurbishment project

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Brown, M.J.; Mathew, P.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station. The MAAP4-CANDU code was used to calculate the progression of postulated severe core damage accidents and fission product releases. Five representative severe core damage accidents were selected: Station Blackout, Small Loss-of-Coolant Accident, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State Accident. Analysis results for only the reference Small LOCA Accident scenario (which is a very low probability event) are discussed in this paper. (author)

  19. Reactivity Insertion Accident (RIA) Capability Status in the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Folsom, Charles Pearson [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States); Veeraraghavan, Swetha [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-05-01

    One of the Challenge Problems being considered within CASL relates to modelling and simulation of Light Water Reactor LWR) fuel under Reactivity Insertion Accident (RIA) conditions. BISON is the fuel performance code used within CASL for LWR fuel under both normal operating and accident conditions, and thus must be capable of addressing the RIA challenge problem. This report outlines required BISON capabilities for RIAs and describes the current status of the code. Information on recent accident capability enhancements, application of BISON to a RIA benchmark exercise, and plans for validation to RIA behavior are included.

  20. Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J., E-mail: brad.merrill@inl.gov; Bragg-Sitton, Shannon M., E-mail: shannon.bragg-sitton@inl.gov; Humrickhouse, Paul W., E-mail: paul.humrickhouse@inl.gov

    2017-04-15

    Highlights: • Accident tolerant fuels (ATF) systems are currently under development for LWRs. • Many performance analysis tools are specifically developed for UO{sub 2}–Zr alloy fuel. • Modifications were made to the MELCOR code for candidate ATF cladding. • Preliminary analysis results for SiC and FeCrAl cladding concepts are presented. - Abstract: A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO{sub 2}–Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This paper describes modifications

  1. Analysis on the severe accidents in KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myoung Jae; Cheong, Y. H.; Choi, Y. S.; Cheon, E. J. [PlaGen, Seoul (Korea, Republic of)

    2003-11-15

    The establishment of regulatory and approval systems for KSTAR (Korea Superconducting Tokamak Advanced Research) has been demanded as the facility is targeted to be completed in the year of 2005. Such establishment can be achieved by performing adequate and in-depth analyses on safety issues covering radiological and chemical hazard materials, radiation protection, high vacuum, very low temperature, etc. The loss of coolant accidents and the loss of vacuum accident in fusion facilities have been introduced with summary of simulation results that were previously reported for ITER and JET. Computer codes that are actively used for accident simulation research are examined and their main features are briefly described. It can be stated that the safety analysis is indispensable to secure the safety of workers and individual members of the public as well as to establish the regulatory and approval systems for KSTAR tokamak.

  2. Safety analysis of surface haulage accidents

    Energy Technology Data Exchange (ETDEWEB)

    Randolph, R.F.; Boldt, C.M.K.

    1996-12-31

    Research on improving haulage truck safety, started by the U.S. Bureau of Mines, is being continued by its successors. This paper reports the orientation of the renewed research efforts, beginning with an update on accident data analysis, the role of multiple causes in these accidents, and the search for practical methods for addressing the most important causes. Fatal haulage accidents most often involve loss of control or collisions caused by a variety of factors. Lost-time injuries most often involve sprains or strains to the back or multiple body areas, which can often be attributed to rough roads and the shocks of loading and unloading. Research to reduce these accidents includes improved warning systems, shock isolation for drivers, encouraging seatbelt usage, and general improvements to system and task design.

  3. A study on the core analysis methodology for SMART CEA ejection accident-I

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Kyun; Lee, Chung Chan; Kim, Kyo Yoon; Cho, Byung Oh

    1999-04-01

    A methodology to analyze the fuel enthalpy is developed based on MASTER that is a time dependent 3 dimensional core analysis code. Using the proposed methodology, SMART CEA ejection accident is analyzed. Moreover, radiation doses are estimated at the exclusion area boundary and low population zone to confirm the criteria for the accident. (Author). 31 refs., 13 tabs., 18 figs.

  4. Frozen Accident Pushing 50: Stereochemistry, Expansion, and Chance in the Evolution of the Genetic Code.

    Science.gov (United States)

    Koonin, Eugene V

    2017-05-23

    Nearly 50 years ago, Francis Crick propounded the frozen accident scenario for the evolution of the genetic code along with the hypothesis that the early translation system consisted primarily of RNA. Under the frozen accident perspective, the code is universal among modern life forms because any change in codon assignment would be highly deleterious. The frozen accident can be considered the default theory of code evolution because it does not imply any specific interactions between amino acids and the cognate codons or anticodons, or any particular properties of the code. The subsequent 49 years of code studies have elucidated notable features of the standard code, such as high robustness to errors, but failed to develop a compelling explanation for codon assignments. In particular, stereochemical affinity between amino acids and the cognate codons or anticodons does not seem to account for the origin and evolution of the code. Here, I expand Crick's hypothesis on RNA-only translation system by presenting evidence that this early translation already attained high fidelity that allowed protein evolution. I outline an experimentally testable scenario for the evolution of the code that combines a distinct version of the stereochemical hypothesis, in which amino acids are recognized via unique sites in the tertiary structure of proto-tRNAs, rather than by anticodons, expansion of the code via proto-tRNA duplication, and the frozen accident.

  5. Severe accident analysis of a station blackout accident using MAAP-CANDU for the Point Lepreau station refurbishment project level 2 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Petoukhov, S.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station, using the MAAP-CANDU code to simulate the progression of severe core damage accidents and fission product releases. Five representative severe accidents were selected: Station Blackout, Small Loss-of-Coolant, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State. Analysis results for the reference station blackout accident are discussed in this paper. (author)

  6. Inclusion of models to describe severe accident conditions in the fuel simulation code DIONISIO

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Daverio, Hernando [Gerencia Reactores y Centrales Nucleares, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Denis, Alicia [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina)

    2017-04-15

    The simulation of fuel rod behavior is a complex task that demands not only accurate models to describe the numerous phenomena occurring in the pellet, cladding and internal rod atmosphere but also an adequate interconnection between them. In the last years several models have been incorporated to the DIONISIO code with the purpose of increasing its precision and reliability. After the regrettable events at Fukushima, the need for codes capable of simulating nuclear fuels under accident conditions has come forth. Heat removal occurs in a quite different way than during normal operation and this fact determines a completely new set of conditions for the fuel materials. A detailed description of the different regimes the coolant may exhibit in such a wide variety of scenarios requires a thermal-hydraulic formulation not suitable to be included in a fuel performance code. Moreover, there exist a number of reliable and famous codes that perform this task. Nevertheless, and keeping in mind the purpose of building a code focused on the fuel behavior, a subroutine was developed for the DIONISIO code that performs a simplified analysis of the coolant in a PWR, restricted to the more representative situations and provides to the fuel simulation the boundary conditions necessary to reproduce accidental situations. In the present work this subroutine is described and the results of different comparisons with experimental data and with thermal-hydraulic codes are offered. It is verified that, in spite of its comparative simplicity, the predictions of this module of DIONISIO do not differ significantly from those of the specific, complex codes.

  7. HTGR severe accident sequence analysis

    Energy Technology Data Exchange (ETDEWEB)

    Harrington, R.M.; Ball, S.J.; Kornegay, F.C.

    1982-01-01

    Thermal-hydraulic, fission product transport, and atmospheric dispersion calculations are presented for hypothetical severe accident release paths at the Fort St. Vrain (FSV) high temperature gas cooled reactor (HTGR). Off-site radiation exposures are calculated for assumed release of 100% of the 24 hour post-shutdown core xenon and krypton inventory and 5.5% of the iodine inventory. The results show conditions under which dose avoidance measures would be desirable and demonstrate the importance of specific release characteristics such as effective release height. 7 tables.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Vaurio, J. K.; Mueller, C.

    1978-06-01

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

  10. The analysis of thermal-hydraulic models in MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. H.; Hur, C.; Kim, D. K.; Cho, H. J. [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)

    1996-07-15

    The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed.

  11. Thermal Hydraulic Analysis of K-DEMO Single Blanket Module for Preliminary Accident Analysis using MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

  12. [Hanggliding accidents. Distribution of injuries and accident analysis].

    Science.gov (United States)

    Ballmer, F T; Jakob, R P

    1989-12-01

    Paragliding--a relatively new sport to Switzerland--brought 23 patients with 48 injuries (38% lower limb and 29% spinal) within a period of 8 months to the Inselspital University hospital in Berne. The aim of the study in characterizing these injuries is to formulate some guidelines towards prevention. With over 90% of accidents occurring at either take off or landing, emphasis on better training for the beginner is proposed with strict guidelines for the more experienced pilot flying in unfavourable conditions.

  13. Users' guide to CACECO containment analysis code. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Peak, R.D.

    1979-06-01

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. The code is included in the National Energy Software Center Library at Argonne National Laboratory as Program No. 762. This users' guide describes the CACECO code and its data input requirements. The code description covers the many mathematical models used and the approximations used in their solution. The descriptions are detailed to the extent that the user can modify the code to suit his unique needs, and, indeed, the reader is urged to consider code modification acceptable.

  14. Code Coupling for Multi-Dimensional Core Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-05-15

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  15. Developing techniques for cause-responsibility analysis of occupational accidents.

    Science.gov (United States)

    Jabbari, Mousa; Ghorbani, Roghayeh

    2016-11-01

    The aim of this study was to specify the causes of occupational accidents, determine social responsibility and the role of groups involved in work-related accidents. This study develops occupational accidents causes tree, occupational accidents responsibility tree, and occupational accidents component-responsibility analysis worksheet; based on these methods, it develops cause-responsibility analysis (CRA) techniques, and for testing them, analyzes 100 fatal/disabling occupational accidents in the construction setting that were randomly selected from all the work-related accidents in Tehran, Iran, over a 5-year period (2010-2014). The main result of this study involves two techniques for CRA: occupational accidents tree analysis (OATA) and occupational accidents components analysis (OACA), used in parallel for determination of responsible groups and responsibilities rate. From the results, we find that the management group of construction projects has 74.65% responsibility of work-related accidents. The developed techniques are purposeful for occupational accidents investigation/analysis, especially for the determination of detailed list of tasks, responsibilities, and their rates. Therefore, it is useful for preventing work-related accidents by focusing on the responsible group's duties.

  16. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  17. Development of accident management technology and computer codes -A study for nuclear safety improvement-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Kyu; Jae, Moo Sung; Jo, Young Gyun; Park, Rae Jun; Kim, Jae Hwan; Ha, Jae Ju; Kang, Dae Il; Choi, Sun Young; Kim, Si Hwan [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-01

    We have surveyed new technologies and research results for the accident management of nuclear power plants. And, based on the concept of using the existing plant capabilities for accident management, both in-vessel and ex-vessel strategies were identified and analyzed. When assessing accident management strategies, their effectiveness, adverse effects, and their feasibility must be considered. We have developed a framework for assessing the strategies with these factors in mind. We have applied the developed framework to assessing the strategies, including the likelihood that the operator correctly diagnoses the situation and successfully implements the strategies. Finally, the cavity flooding strategy was assessed by applying it to the station blackout sequence, which have been identified as one of the major contributors to risk at the reference plant. The thermohydraulic analyses with sensitivity calculations have been performed using MAAP 4 computer code. (Author).

  18. Hindsight Bias in Cause Analysis of Accident

    Institute of Scientific and Technical Information of China (English)

    Atsuo Murata; Yasunari Matsushita

    2014-01-01

    It is suggested that hindsight becomes an obstacle to the objective investigation of an accident, and that the proper countermeasures for the prevention of such an accident is impossible if we view the accident with hindsight. Therefore, it is important for organizational managers to prevent hindsight from occurring so that hindsight does not hinder objective and proper measures to be taken and this does not lead to a serious accident. In this study, a basic phenomenon potentially related to accidents, that is, hindsight was taken up, and an attempt was made to explore the phenomenon in order to get basically insights into the prevention of accidents caused by such a cognitive bias.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  20. Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yongjae; Seo, Seungwon; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of); Ha, Kwang Soon; Kim, Hwan-Yeol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Assessment of the safety features over the hypothesized severe accidents may be performed experimentally or numerically. Due to the considerable time and expenditures, experimental assessment is implemented only to the limited cases. Therefore numerical assessment has played a major role in revisiting severe accident analysis of the existing or newly designed power plants. Computer codes for the numerical analysis of severe accidents are categorized as the fast running integral code and detailed code. Fast running integral codes are characterized by a well-balanced combination of detailed and simplified models for the simulation of the relevant phenomena within an NPP in the case of a severe accident. MAAP, MELCOR and ASTEC belong to the examples of fast running integral codes. Detailed code is to model as far as possible all relevant phenomena in detail by mechanistic models. The examples of detailed code is SCDAP/RELAP5. Using the MELCOR, Carbajo. investigated sensitivity studies of Station Black Out (SBO) using the MELCOR for Peach Bottom BWR. Park et al. conduct regulatory research of the PWR severe accident. Ahn et al. research sensitivity analysis of the severe accident for APR1400 with MELCOR 1.8.4. Lee et al. investigated RCS depressurization strategy and developed a core coolability map for independent scenarios of Small Break Loss-of-Coolant Accident (SBLOCA), SBO, and Total Loss of Feed Water (TLOFW). In this study, three initiating cases were selected, which are SBLOCA without SI, SBO, and TLOFW. The initiating cases exhibit the highest probability of transitioning into core damage according to PSA 1 of OPR 1000. The objective of this study is to investigate the reactor core coolability during hypothesized severe accidents of OPR1000. As a representative indicator, we have employed Jakob number and developed JaCET and JaMCT using the MELCOR simulation. Although the RCS pressures for the respective accident scenarios were different, the JaMCT and Ja

  1. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.; PIEPHO, M.G.

    2000-03-23

    This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  2. Canister storage building design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    KOPELIC, S.D.

    1999-02-25

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

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

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2012-01-01

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

  4. Extending the application range of a fuel performance code from normal operating to design basis accident conditions

    Science.gov (United States)

    Van Uffelen, P.; Győri, C.; Schubert, A.; van de Laar, J.; Hózer, Z.; Spykman, G.

    2008-12-01

    Two types of fuel performance codes are generally being applied, corresponding to the normal operating conditions and the design basis accident conditions, respectively. In order to simplify the code management and the interface between the codes, and to take advantage of the hardware progress it is favourable to generate a code that can cope with both conditions. In the first part of the present paper, we discuss the needs for creating such a code. The second part of the paper describes an example of model developments carried out by various members of the TRANSURANUS user group for coping with a loss of coolant accident (LOCA). In the third part, the validation of the extended fuel performance code is presented for LOCA conditions, whereas the last section summarises the present status and indicates needs for further developments to enable the code to deal with reactivity initiated accident (RIA) events.

  5. Cold Vacuum Drying (CVD) Facility Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    PIEPHO, M.G.

    1999-10-20

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR.

  6. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Bunt, R. [Southern Nuclear, Atlanta, GA (United States); Corradini, M. [Univ. of Wisconsin, Madison, WI (United States); Ellison, Paul B. [GE Power and Water, Duluth, GA (United States); Francis, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Gabor, John D. [Erin Engineering, Walnut Creek, CA (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Henry, C. [Fauske and Associates, Burr Ridge, IL (United States); Linthicum, R. [Exelon Corp., Chicago, IL (United States); Luangdilok, W. [Fauske and Associates, Burr Ridge, IL (United States); Lutz, R. [PWR Owners Group (PWROG); Paik, C. [Fauske and Associates, Burr Ridge, IL (United States); Plys, M. [Fauske and Associates, Burr Ridge, IL (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Robb, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wachowiak, R. [Electric Power Research Inst. (EPRI), Knovville, TN (United States)

    2015-01-31

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  7. Analysis of traffic accidents in children

    Directory of Open Access Journals (Sweden)

    Pavlekić Snežana

    2006-01-01

    Full Text Available Introduction: Violent health damages of different origin (accidents, murders, suicides in children and youth are one of the main causes of death and disabilities in this group of population in most countries. Objective: Objective of our paper was to analyze all related factors of traffic accidents involving children and to propose adequate measures of their prevention. Method: The analysis of fatal traffic accidents of children and youth aged to 18 years on the territory of Belgrade, within the period from 1998 to 2002. Results: In relation to other forms of violent death, the traffic mortality rate in children and youth holds the leading position, accounting for 56.9% with pedestrians as the most frequent category (57.4%. The most frequent age was between 7 and 9 years (46.8% and the boys were more frequently injured than the girls. It was established that the majority of children (51.9% was either running across the street outside the pedestrian/ zebra crossings or they were carelessly running out in the street, especially in April, July, August and September. More than a half of them (55.5%, predominantly school children, were injured by the end of working week, on Thursday and Friday. Conclusion: Results of our research have shown that the traffic education of children in our region is inadequate. Due to the abovementioned, it is primarily necessary to establish long-term and permanent education of this category of population. In addition, some public investments in the City infrastructure will be required in order to reduce the risk of traffic injuries in children.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  9. Development of the Severe Accident Analysis DB for the Severe Accident Management Expert System (I)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Ahn, Kwang Il [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    This report contains analysis methodologies and calculation results of 5 initiating events of the severe accident analysis database system. The Ulchin 3,4 NPP has been selected as reference plants. Based on the probabilistic safety analysis of the corresponding plant, 54 accident scenarios, which was predicted to have more than 10-10 /ry occurrence frequency, have been analyzed as base cases for the Large loss of Coolant sequence database. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results in this report will be utilized as input data to develop the database system

  10. Accident progression event tree analysis for postulated severe accidents at N Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M. (Sandia National Labs., Albuquerque, NM (USA)); Medford, G.T. (Science Applications International Corp., Albuquerque, NM (USA))

    1990-06-01

    A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied.

  11. Exploratory analysis of Spanish energetic mining accidents.

    Science.gov (United States)

    Sanmiquel, Lluís; Freijo, Modesto; Rossell, Josep M

    2012-01-01

    Using data on work accidents and annual mining statistics, the paper studies work-related accidents in the Spanish energetic mining sector in 1999-2008. The following 3 parameters are considered: age, experience and size of the mine (in number of workers) where the accident took place. The main objective of this paper is to show the relationship between different accident indicators: risk index (as an expression of the incidence), average duration index for the age and size of the mine variables (as a measure of the seriousness of an accident), and the gravity index for the various sizes of mines (which measures the seriousness of an accident, too). The conclusions of this study could be useful to develop suitable prevention policies that would contribute towards a decrease in work-related accidents in the Spanish energetic mining industry.

  12. Use of detailed thermochemical databases to model chemical interactions in the Severe Accident codes

    Energy Technology Data Exchange (ETDEWEB)

    Barrachin, M. [IPSN/DRS, CEA Cadarache (France)

    2001-07-01

    For the prevention, mitigation and management of severe accidents, many problems related to core melt have to be solved: fuel degradation, melting and relocation, convection in the core melt(s), coolability of the core melt(s), fission product release, hydrogen production, behavior of the materials of the protective layers, ex-vessel spreading of the core melt(s).. To solve these problems such properties like thermal conductivity, heat capacity, density, viscosity, evaporation or sublimation of melts, the solidification behavior (solid/liquid fraction), the tendency to trap or to release the fission products, the stratification of melts notably metallic and oxide, must be known. However most of these properties are delicate to measure directly at high temperature and/or in the radio-active environment produced by the fission products. Therefore some of them must be derived by calculations from the physical-chemical description of the melt: number of phases, phase compositions, proportions of solids and liquids and their respective oxidation state, miscibility of the liquids, solubility of one phase in another, etc. This information is given by the phase diagrams of the materials in presence. Since more than ten years, IPSN has developed in collaboration with THERMODATA (Grenoble, France) a very detailed thermochemical database for the complex system U-O-Zr-Fe-Ni-La-Ba-Ru-Sr-Si-Mg-Ca-Al-(H-Ar). The direct coupling between the severe accident (SA) Codes and a thermochemical code with its database is not actually possible because of the computer time consuming and the size of the database. For this reason, most of the Severe Accident codes usually have a very simplified description for the phase diagrams which are not in agreement with the status of the art. In this presentation, alternative methodologies are detailed with their respective difficulties, the goal being to build an interface between a thermochemical database and a SA Code and to get a fast, accurate and

  13. Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.

  14. Combustion chamber analysis code

    Science.gov (United States)

    Przekwas, A. J.; Lai, Y. G.; Krishnan, A.; Avva, R. K.; Giridharan, M. G.

    1993-05-01

    A three-dimensional, time dependent, Favre averaged, finite volume Navier-Stokes code has been developed to model compressible and incompressible flows (with and without chemical reactions) in liquid rocket engines. The code has a non-staggered formulation with generalized body-fitted-coordinates (BFC) capability. Higher order differencing methodologies such as MUSCL and Osher-Chakravarthy schemes are available. Turbulent flows can be modeled using any of the five turbulent models present in the code. A two-phase, two-liquid, Lagrangian spray model has been incorporated into the code. Chemical equilibrium and finite rate reaction models are available to model chemically reacting flows. The discrete ordinate method is used to model effects of thermal radiation. The code has been validated extensively against benchmark experimental data and has been applied to model flows in several propulsion system components of the SSME and the STME.

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

    CERN Multimedia

    HSE Unit

    2015-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Lindley Benjamin A.

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Palmrose, D E; Yang, J M

    2007-05-10

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

  18. A COMPARISON OF SOME STATISTICAL TECHNIQUES FOR ROAD ACCIDENT ANALYSIS

    NARCIS (Netherlands)

    OPPE, S INST ROAD SAFETY RES, SWOV

    1992-01-01

    At the TRRL/SWOV Workshop on Accident Analysis Methodology, heldin Amsterdam in 1988, the need to establish a methodology for the analysis of road accidents was firmly stated by all participants. Data from different countries cannot be compared because there is no agreement on research methodology,

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-02-01

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

  20. Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

    2012-09-30

    Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

  1. TMI-2 accident: core heat-up analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ardron, K.H.; Cain, D.G.

    1981-01-01

    This report summarizes NSAC study of reactor core thermal conditions during the accident at Three Mile Island, Unit 2. The study focuses primarily on the time period from core uncovery (approximately 113 minutes after turbine trip) through the initiation of sustained high pressure injection (after 202 minutes). The transient analysis is based upon established sequences of events; plant data; post-accident measurements; interpretation or indirect use of instrument responses to accident conditions.

  2. A Comparative analysis for control rod drop accident in RETRAN DNB and CETOP DNB Model

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chang Keun; Kim, Yo Han; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    In Korea, the nuclear industries such as fuel manufacturer, the architect engineer and the utility, have been using the methodologies and codes of vendors, such as Westinghouse(WH), Combustion Engineering, for the safety analyses of nuclear power plants. Consequently the industries have kept up the many organizations to operate the methodologies and to maintain the codes for each vendor. It may occur difficulty to improve the safety analyses efficiency and technology related. So, the necessity another of methodologies and code systems applicable to Non- LOCA, beyond design basis accident and performance analyses for all types of pressurized water reactor(PWR) has been raised. Due to the above reason, the Korea Electric Power Research Institute(KEPRI) had decided to develop the new safety analysis code system for Korea Standard Nuclear Power Plants in Korea. As the first requirement, the best-estimate codes were required for applicable wider application area and realistic behavior prediction of power plants with various and sophisticated functions. After the investigation for few candidates, RETRAN-3D has been chosen as a system analysis code. As a part of the feasibility estimation for the methodology and code system, CRD(Control Rod Drop) accident which an event of Non-LOCA accidents for Uljin units 3 and 4 and Yonggwang 1 and 2 was selected to verify the feasibility of the methodology using the RETRAN-3D. In this paper, RETRAN DNB Model and CETOP DNB Model were analyzed by using comparative method.

  3. A cluster analysis on road traffic accidents using genetic algorithms

    Science.gov (United States)

    Saharan, Sabariah; Baragona, Roberto

    2017-04-01

    The analysis of traffic road accidents is increasingly important because of the accidents cost and public road safety. The availability or large data sets makes the study of factors that affect the frequency and severity accidents are viable. However, the data are often highly unbalanced and overlapped. We deal with the data set of the road traffic accidents recorded in Christchurch, New Zealand, from 2000-2009 with a total of 26440 accidents. The data is in a binary set and there are 50 factors road traffic accidents with four level of severity. We used genetic algorithm for the analysis because we are in the presence of a large unbalanced data set and standard clustering like k-means algorithm may not be suitable for the task. The genetic algorithm based on clustering for unknown K, (GCUK) has been used to identify the factors associated with accidents of different levels of severity. The results provided us with an interesting insight into the relationship between factors and accidents severity level and suggest that the two main factors that contributes to fatal accidents are "Speed greater than 60 km h" and "Did not see other people until it was too late". A comparison with the k-means algorithm and the independent component analysis is performed to validate the results.

  4. Applicability of simplified human reliability analysis methods for severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Boring, R.; St Germain, S. [Idaho National Lab., Idaho Falls, Idaho (United States); Banaseanu, G.; Chatri, H.; Akl, Y. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2016-03-15

    Most contemporary human reliability analysis (HRA) methods were created to analyse design-basis accidents at nuclear power plants. As part of a comprehensive expansion of risk assessments at many plants internationally, HRAs will begin considering severe accident scenarios. Severe accidents, while extremely rare, constitute high consequence events that significantly challenge successful operations and recovery. Challenges during severe accidents include degraded and hazardous operating conditions at the plant, the shift in control from the main control room to the technical support center, the unavailability of plant instrumentation, and the need to use different types of operating procedures. Such shifts in operations may also test key assumptions in existing HRA methods. This paper discusses key differences between design basis and severe accidents, reviews efforts to date to create customized HRA methods suitable for severe accidents, and recommends practices for adapting existing HRA methods that are already being used for HRAs at the plants. (author)

  5. Loss of Coolant Accident Analysis Methodology for SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Bae, K. H.; Lee, G. H.; Yang, S. H.; Yoon, H. Y.; Kim, S. H.; Kim, H. C

    2006-02-15

    The analysis methodology on the Loss-of-coolant accidents (LOCA's) for SMART-P is described in this report. SMART-P is an advanced integral type PWR producing a maximum thermal power of 65.5 MW with metallic fuel. LOCA's are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. Since SMART-P contains the major primary circuit components in a single Reactor Pressure Vessel (RPV), the possibility of a large break LOCA (LBLOCA) is inherently eliminated and only the small break LOCA is postulated. This report describes the outline and acceptance criteria of small break LOCA (SBLOCA) for SMART-P and documents the conservative analytical model and method and the analysis results using the TASS/SMR code. This analysis method is applied in the SBLOCA analysis performed for the ECCS performance evaluation which is described in the section 6.3.3 of the safety analysis report. The prediction results of SBLOCA analysis model of SMART-P for the break flow, system's pressure and temperature distributions, reactor coolant distribution, single and two-phase natural circulation phenomena, and the time of major sequence of events, etc. should be compared and verified with the applicable separate and integral effects test results. Also, it is required to set-up the feasible acceptance criteria applicable to the metallic fueled integral reactor of SMART-P. The analysis methodology for the SBLOCA described in this report will be further developed and validated as the design and licensing status of SMART-P evolves.

  6. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.

    1999-09-09

    This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  7. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for internal dosimetry. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harrison, J.D. [National Radiological Protection Board (United Kingdom); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1998-04-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA internal dosimetry models.

  8. Probabilistic accident consequence uncertainty analysis -- Late health effects uncertainty assessment. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    Little, M.P.; Muirhead, C.R. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA late health effects models.

  9. NASA's Accident Precursor Analysis Process and the International Space Station

    Science.gov (United States)

    Groen, Frank; Lutomski, Michael

    2010-01-01

    This viewgraph presentation reviews the implementation of Accident Precursor Analysis (APA), as well as the evaluation of In-Flight Investigations (IFI) and Problem Reporting and Corrective Action (PRACA) data for the identification of unrecognized accident potentials on the International Space Station.

  10. SAMPSON Parallel Computation for Sensitivity Analysis of TEPCO's Fukushima Daiichi Nuclear Power Plant Accident

    Science.gov (United States)

    Pellegrini, M.; Bautista Gomez, L.; Maruyama, N.; Naitoh, M.; Matsuoka, S.; Cappello, F.

    2014-06-01

    On March 11th 2011 a high magnitude earthquake and consequent tsunami struck the east coast of Japan, resulting in a nuclear accident unprecedented in time and extents. After scram started at all power stations affected by the earthquake, diesel generators began operation as designed until tsunami waves reached the power plants located on the east coast. This had a catastrophic impact on the availability of plant safety systems at TEPCO's Fukushima Daiichi, leading to the condition of station black-out from unit 1 to 3. In this article the accident scenario is studied with the SAMPSON code. SAMPSON is a severe accident computer code composed of hierarchical modules to account for the diverse physics involved in the various phases of the accident evolution. A preliminary parallelization analysis of the code was performed using state-of-the-art tools and we demonstrate how this work can be beneficial to the nuclear safety analysis. This paper shows that inter-module parallelization can reduce the time to solution by more than 20%. Furthermore, the parallel code was applied to a sensitivity study for the alternative water injection into TEPCO's Fukushima Daiichi unit 3. Results show that the core melting progression is extremely sensitive to the amount and timing of water injection, resulting in a high probability of partial core melting for unit 3.

  11. National and regional analysis of road accidents in Spain.

    Science.gov (United States)

    Tolón-Becerra, A; Lastra-Bravo, X; Flores-Parra, I

    2013-01-01

    In Spain, the absolute fatality figures decreased almost 50 percent between 1998 and 2009. Despite this great effort, road mortality is still of great concern to political authorities. Further progress requires efficient road safety policy based on an optimal set of measures and targets that consider the initial conditions and characteristics in each region. This study attempts to analyze road accidents in Spain and its provinces in time and space during 1998-2009. First, we analyzed daily, monthly, and nationwide (NUTS 0) development of road accidents, the correlation between logarithmic transformations of road accidents and territorial and socioeconomic variables, the causality by simple linear regression of road accidents and territorial and socioeconomic variables, and preliminary frequency by fast Fourier transform. Then we analyzed the annual trend in accidents in the Spanish provinces (NUTS 3) and found a correlation between the logarithmic transformations of the mortality rate, fatalities per fatal accident, and accidents resulting in injuries per inhabitant variables and population, population density, gross domestic product (GDP), length of road network, and area. Finally, causality was analyzed by simple linear regression. The most outstanding results were the negative correlation between mortality rate and population density in Spanish provinces, which has increased over time, and that road accidents in Spain have an approximate periodicity of 57 days. The fast Fourier transform analysis of road accident frequency in Spain was useful in identifying the periodic, harmonic components of accidents and casualties. The periodicity observed both for the period 1998-2009 and by year showed that the highest intensity in road accidents was bimonthly, despite the lower number of accidents and casualties in the spectra of amplitude and power and efforts to reduce the intensity and concentration during off-season travel (summer and December).

  12. Pressure Load Analysis during Severe Accidents for the Evaluation of Late Containment Failure in OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The MAAP code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a level 2 probabilistic safety assessment or severe accident management strategy developments. The code employs lots of user-options for supporting a sensitivity and uncertainty analysis. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to in-vessel hydrogen generation, gas combustion in the containment, corium distribution in the containment after a reactor vessel failure, corium coolability in the reactor cavity, and molten-corium interaction with concrete. The phenomenology of severe accidents is extremely complex. In this paper, a sampling-based phenomenological uncertainty analysis was performed to statistically quantify uncertainties associated with the pressure load of a containment building for a late containment failure evaluation, based on the key modeling parameters employed in the MAAP code and random samples for those parameters. Phenomenological issues surrounding the late containment failure mode are highly complex. Included are the pressurization owing to steam generation in the cavity, molten corium-concrete interaction, late hydrogen burn in the containment, and the secondary heat removal availability. The methodology and calculation results can be applied for the optimum assessment of a late containment failure model. The accident sequences considered were a loss of coolant accidents and loss of offsite accidents expected in the OPR-1000 plant. As a result, uncertainties addressed in the pressure load of the containment building were quantified as a function of time. A realistic evaluation of the mean and variance estimates provides a more complete

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-09-01

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

  14. [Severe parachuting accident. Analysis of 122 cases].

    Science.gov (United States)

    Krauss, U; Mischkowsky, T

    1993-06-01

    Based on a population of 122 severely injured patients the causes of paragliding accidents and the patterns of injury are analyzed. A questionnaire is used to establish a sport-specific profile for the paragliding pilot. The lower limbs (55.7%) and the lower parts of the spine (45.9%) are the most frequently injured parts of the body. There is a high risk of multiple injuries after a single accident because of the tremendous axial power. The standard of equipment is good in over 90% of the cases. Insufficient training and failure to take account of geographical and meteorological conditions are the main determinants of accidents sustained by paragliders, most of whom are young. Nevertheless, 80% of our patients want to continue paragliding. Finally some advice is given on how to prevent paragliding accidents and injuries.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  18. Traffic Accident, System Model and Cluster Analysis in GIS

    Directory of Open Access Journals (Sweden)

    Veronika Vlčková

    2015-07-01

    Full Text Available One of the many often frequented topics as normal journalism, so the professional public, is the problem of traffic accidents. This article illustrates the orientation of considerations to a less known context of accidents, with the help of constructive systems theory and its methods, cluster analysis and geoinformation engineering. Traffic accident is reframing the space-time, and therefore it can be to study with tools of technology of geographic information systems. The application of system approach enabling the formulation of the system model, grabbed by tools of geoinformation engineering and multicriterial and cluster analysis.

  19. The relationship between alcohol and traffic accidents an actuarial analysis.

    NARCIS (Netherlands)

    Nelker, G.

    1970-01-01

    This analysis deals with the accident chances of the excessive drinkers, moderate drinkers and non-drinkers. The Swedish motor insurance results for total abstainers are compared to the average policyholders.

  20. Advanced accident sequence precursor analysis level 1 models

    Energy Technology Data Exchange (ETDEWEB)

    Sattison, M.B.; Thatcher, T.A.; Knudsen, J.K.; Schroeder, J.A.; Siu, N.O. [Idaho National Engineering Lab., Idaho National Lab., Idaho Falls, ID (United States)

    1996-03-01

    INEL has been involved in the development of plant-specific Accident Sequence Precursor (ASP) models for the past two years. These models were developed for use with the SAPHIRE suite of PRA computer codes. They contained event tree/linked fault tree Level 1 risk models for the following initiating events: general transient, loss-of-offsite-power, steam generator tube rupture, small loss-of-coolant-accident, and anticipated transient without scram. Early in 1995 the ASP models were revised based on review comments from the NRC and an independent peer review. These models were released as Revision 1. The Office of Nuclear Regulatory Research has sponsored several projects at the INEL this fiscal year to further enhance the capabilities of the ASP models. Revision 2 models incorporates more detailed plant information into the models concerning plant response to station blackout conditions, information on battery life, and other unique features gleaned from an Office of Nuclear Reactor Regulation quick review of the Individual Plant Examination submittals. These models are currently being delivered to the NRC as they are completed. A related project is a feasibility study and model development of low power/shutdown (LP/SD) and external event extensions to the ASP models. This project will establish criteria for selection of LP/SD and external initiator operational events for analysis within the ASP program. Prototype models for each pertinent initiating event (loss of shutdown cooling, loss of inventory control, fire, flood, seismic, etc.) will be developed. A third project concerns development of enhancements to SAPHIRE. In relation to the ASP program, a new SAPHIRE module, GEM, was developed as a specific user interface for performing ASP evaluations. This module greatly simplifies the analysis process for determining the conditional core damage probability for a given combination of initiating events and equipment failures or degradations.

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

    Directory of Open Access Journals (Sweden)

    Tate Warren P

    2008-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-15

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

  3. Analysis of Credible Accidents for Argonaut Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hawley, S. C.; Kathern, R. L.; Robkin, M. A.

    1981-04-01

    Five areas of potential accidents have been evaluated for the Argonaut-UTR reactors. They are: • insertion of excess reactivity • catastrophic rearrangement of the core • explosive chemical reaction • graphite fire • fuel-handling accident. A nuclear excursion resulting from the rapid insertion of the maximum available excess reactivity would produce only 12 MWs which is insufficient to cause fuel melting even with conservative assumptions. Although precise structural rearrangement of the core would create a potential hazard, it is simply not credible to assume that such an arrangement would result from the forces of an earthquake or other catastrophic event. Even damage to the fuel from falling debris or other objects is unlikely given the normal reactor structure. An explosion from a metal-water reaction could not occur because there is no credible source of sufficient energy to initiate the reaction. A graphite fire could conceivably create some damage to the reactor but not enough to melt any fuel or initiate a metal-water reaction. The only credible accident involving offsite doses was determined to be a fuel-handling accident which, given highly conservative assumptions, would produce a whole-body dose equivalent of 2 rem from noble gas immersion and a lifetime dose equivalent commitment to the thyroid of 43 rem from radioiodines.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-15

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

  6. FARO base case post-test analysis by COMETA code

    Energy Technology Data Exchange (ETDEWEB)

    Annunziato, A.; Addabbo, C. [Joint Research Centre, Ispra (Italy)

    1995-09-01

    The paper analyzes the COMETA (Core Melt Thermal-Hydraulic Analysis) post test calculations of FARO Test L-11, the so-called Base Case Test. The FARO Facility, located at JRC Ispra, is used to simulate the consequences of Severe Accidents in Nuclear Power Plants under a variety of conditions. The COMETA Code has a 6 equations two phase flow field and a 3 phases corium field: the jet, the droplets and the fused-debris bed. The analysis shown that the code is able to pick-up all the major phenomena occurring during the fuel-coolant interaction pre-mixing phase.

  7. Safety Analysis Results for Cryostat Ingress Accidents in ITER

    Science.gov (United States)

    Merrill, B. J.; Cadwallader, L. C.; Petti, D. A.

    1997-06-01

    Accidents involving the ingress of air, helium, or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.

  8. Accurate dose assessment system for an exposed person utilising radiation transport calculation codes in emergency response to a radiological accident.

    Science.gov (United States)

    Takahashi, F; Shigemori, Y; Seki, A

    2009-01-01

    A system has been developed to assess radiation dose distribution inside the body of exposed persons in a radiological accident by utilising radiation transport calculation codes-MCNP and MCNPX. The system consists mainly of two parts, pre-processor and post-processor of the radiation transport calculation. Programs for the pre-processor are used to set up a 'problem-dependent' input file, which defines the accident condition and dosimetric quantities to be estimated. The program developed for the post-processor part can effectively indicate dose information based upon the output file of the code. All of the programs in the dosimetry system can be executed with a generally used personal computer and accurately give the dose profile to an exposed person in a radiological accident without complicated procedures. An experiment using a physical phantom was carried out to verify the availability of the dosimetry system with the developed programs in a gamma ray irradiation field.

  9. Exposure rate response analysis of criticality accident dectector at Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Zino, J.F.

    1995-01-01

    This analysis investigated the exposure response behavior of gamma-ray ionization chambers used in the criticality accident systems at the Savannah River Site (SRS). The project consisted of performing exposure response measurements with a calibrated {sup 137}Cs source for benchmarking of the MCNP Monte Carlo code. MCNP was then used to extrapolate the ion chamber`s response to gamma-rays with energies outside the current domain of measured data for criticality fission sources.

  10. CFD Analysis of Migration Mechanism of Source Term Under Severe Accident

    Institute of Scientific and Technical Information of China (English)

    CHEN; Lin-lin; SUN; Xue-ting; JI; Song-tao

    2013-01-01

    The analysis of the migration of source term under severe accident is one of the important aspects of‘Studies on Migration Mechanism of the Source Term under Severe Accident’,which is a significant task of the National Large Advanced PWR Research Program.This research aims at building up a method for analyzing fission product behavior in the containment with CFD code.The effect of PCCS(Passive

  11. An Analysis of Syndrome Coding

    Science.gov (United States)

    Amiruzzaman, Md; Abdullah-Al-Wadud, M.; Chung, Yoojin

    In this paper a detail analysis is presented based on BCH syndrome coding for covert channel data hiding methods. The experimented technique is nothing but a syndrome coding algorithm with a coset based approach, analyzed results are showing that the examined method has more flexibility to choose coset, also providing less modification distortion caused by data hiding. Analyzed method presented by clear mathematical way. As it is mathematical equation dependent, hence analyzed results are showing that the analyzed method has fast computation ability and find perfect roots for modification.

  12. Analysis of FY79 Army Aircraft Accidents.

    Science.gov (United States)

    1980-04-01

    materiel failure/ malfunction 1. A accident. taskerrr (E) s Jb prforanc whch evited The acronym for the 3W approach to the Investigation, fro tht rquied y... lulli , It t , 1AZ" l iil~ i ..-... lI :.].,;.’,, - ,- S !5U" i_ ’ , o ’o , "% ", . #.,""" ".’ - "-’.’’’’-. "" " "’ ."- " " " II I S.7 44’ * U 11 Imem <I

  13. Neutronic analysis of LMFBRs during severe core disruptive accidents

    Energy Technology Data Exchange (ETDEWEB)

    Tomlinson, E.T.

    1979-01-01

    A number of numerical experiments were performed to assess the validity of diffusion theory and various perturbation methods for calculating the reactivity state of a severely disrupted liquid metal cooled fast breeder reactor (LMFBR). The disrupted configurations correspond, in general, to phases through which an LMFBR core could pass during a core disruptive accident (CDA). Two-reactor models were chosen for this study, the two zone, homogeneous Clinch River Breeder Reactor and the Large Heterogeneous Reactor Design Study Core. The various phases were chosen to approximate the CDA results predicted by the safety analysis code SAS3D. The calculational methods investigated in this study include the eigenvalue difference technique based on both discrete ordinate transport theory and diffusion theory, first-order perturbation theory, exact perturbation theory, and a new hybrid perturbation theory. Selected cases were analyzed using Monte Carlo methods. It was found that in all cases, diffusion theory and perturbation theory yielded results for the change in reactivity that significantly disagreed with both the discrete ordinate and Monte Carlo results. These differences were, in most cases, in a nonconservative direction.

  14. Room Heat-Up Analysis with GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Olza, J. M.

    2010-07-01

    The GOTHIC{sup T}M computer code is a state-of-the art program for modeling multiphase, multicomponent fluid flow. GOTHIC is rapidly becoming the industry-standard code for performing both containment design basis accident (DBA) analyses and analyses to support equipment qualification. GOTHIC has a flexible nodding structure that allows both lumped parameter and 3-D modeling capabilities. Multidimensional analysis capabilities greatly enhance the study of noncondensable gases and stratification and permit the calculation of flow field details within any given volume.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    1998-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-02-01

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

  17. Accident Sequence Evaluation Program: Human reliability analysis procedure

    Energy Technology Data Exchange (ETDEWEB)

    Swain, A.D.

    1987-02-01

    This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs.

  18. Detection and analysis of accident black spots with even small accident figures.

    NARCIS (Netherlands)

    Oppe, S.

    1982-01-01

    Accident black spots are usually defined as road locations with high accident potentials. In order to detect such hazardous locations we have to know the probability of an accident for a traffic situation of some kind, or the mean number of accidents for some unit of time. In almost all procedures

  19. Detection and analysis of accident black spots with even small accident figures.

    NARCIS (Netherlands)

    Oppe, S.

    1982-01-01

    Accident black spots are usually defined as road locations with high accident potentials. In order to detect such hazardous locations we have to know the probability of an accident for a traffic situation of some kind, or the mean number of accidents for some unit of time. In almost all procedures

  20. Hanford Waste Tank Bump Accident and Consequence Analysis

    Energy Technology Data Exchange (ETDEWEB)

    BRATZEL, D.R.

    2000-06-20

    This report provides a new evaluation of the Hanford tank bump accident analysis and consequences for incorporation into the Authorization Basis. The analysis scope is for the safe storage of waste in its current configuration in single-shell and double-shell tanks.

  1. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, main report

    Energy Technology Data Exchange (ETDEWEB)

    Harper, F.T.; Young, M.L.; Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States); Lui, C.H. [Nuclear Regulatory Commission, Washington, DC (United States); Goossens, L.H.J.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Paesler-Sauer, J. [Research Center, Karlsruhe (Germany); Helton, J.C. [and others

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The ultimate objective of the joint effort was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. Experts developed their distributions independently. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. To validate the distributions generated for the dispersion code input variables, samples from the distributions and propagated through the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the first of a three-volume document describing the project.

  2. MELCOR accident analysis for ARIES-ACT

    Energy Technology Data Exchange (ETDEWEB)

    Paul W. Humrickhouse; Brad J. Merrill

    2012-08-01

    We model a loss of flow accident (LOFA) in the ARIES-ACT1 tokamak design. ARIES-ACT1 features an advanced SiC blanket with LiPb as coolant and breeder, a helium cooled steel structural ring and tungsten divertors, a thin-walled, helium cooled vacuum vessel, and a room temperature water-cooled shield outside the vacuum vessel. The water heat transfer system is designed to remove heat by natural circulation during a LOFA. The MELCOR model uses time-dependent decay heats for each component determined by 1-D modeling. The MELCOR model shows that, despite periodic boiling of the water coolant, that structures are kept adequately cool by the passive safety system.

  3. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    Science.gov (United States)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  5. Probabilistic accident consequence uncertainty analysis -- Late health effects uncertain assessment. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Little, M.P.; Muirhead, C.R. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA late health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the expert panel on late health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  6. Probabilistic accident consequence uncertainty analysis -- Early health effects uncertainty assessment. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [Univ. of New Mexico, Albuquerque, NM (United States); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA early health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on early health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  7. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for deposited material and external doses. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Boardman, J. [AEA Technology (United Kingdom); Jones, J.A. [National Radiological Protection Board (United Kingdom); Harper, F.T.; Young, M.L. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA deposited material and external dose models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on deposited material and external doses, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  8. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for internal dosimetry. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harrison, J.D. [National Radiological Protection Board (United Kingdom); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1998-04-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA internal dosimetry models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on internal dosimetry, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  9. A preliminary uncertainty analysis of phenomenological inputs in TEXAS-V code

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. H.; Kim, H. D.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Uncertainty analysis is important step in safety analysis of nuclear power plants. The better estimate for the computer codes is on the increase instead of conservative codes. These efforts aim to get more precise evaluation of safety margins, and aim at determining the rate of change in the prediction of codes with one or more input parameters varies within its range of interest. From this point of view, a severe accident uncertainty analysis system, SAUNA, has been improved for TEXAS-V FCI uncertainty analysis. The main objective of this paper is to present the TEXAS FCI uncertainty analysis results implemented through the SAUNA code

  10. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report; Qualifizierung des Kernmodells DYN3D im Komplex mit dem Stoerfallcode ATHLET als fortgeschrittenes Werkzeug fuer die Stoerfallanalyse von WWER-Reaktoren. T. 1. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum

  11. Upgrading the safety toolkit: Initiatives of the accident analysis subgroup

    Energy Technology Data Exchange (ETDEWEB)

    O' Kula, K.R.; Chung, D.Y.

    1999-07-01

    Since its inception, the Accident Analysis Subgroup (AAS) of the Energy Facility Contractors Group (EFCOG) has been a leading organization promoting development and application of appropriate methodologies for safety analysis of US Department of Energy (DOE) installations. The AAS, one of seven chartered by the EFCOG Safety Analysis Working Group, has performed an oversight function and provided direction to several technical groups. These efforts have been instrumental toward formal evaluation of computer models, improving the pedigree on high-use computer models, and development of the user-friendly Accident Analysis Guidebook (AAG). All of these improvements have improved the analytical toolkit for best complying with DOE orders and standards shaping safety analysis reports (SARs) and related documentation. Major support for these objectives has been through DOE/DP-45.

  12. Analysis of Early Severe Accident Initiated by LBLOCA for Qinshan Phase II Nuclear Power Project

    Directory of Open Access Journals (Sweden)

    Shi Xing-Wei

    2013-07-01

    Full Text Available The purpose of this study is to simulate an early Severe Accident (SA scenario more detail through transferring the thermal-hydraulic status of the plant predicted by RELAP5 computer code to SA Program (SAP. Based on the criterion of date extract time, the RELAP5 thermal-hydraulic calculation data is extracted to form a file for SAP input card at 1477K of cladding surface. Relying on the thermal-hydraulic boundary parameters calculated by RELAP5 code, analysis of early SA initiated by the Large Break Loss-of-Coolant Accident (LBLOCA without mitigation measures for Qinshan Phase II Nuclear Power Plant (QSP-II performed by SAP through finding the key events of accident sequence, estimating the amount of hydrogen generation and oxidation behavior of the cladding and evaluating the relocation order of the materials collapsed in the central region of the core. The results of this study are expected to improve the SA analysis methodology more detail through analyzing early SA scenario.

  13. Radionuclides release possibility analysis of MSR at various accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Wie; Kim, Hee Reyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    There are some accidents which go beyond our expectation such as Fukushima Daiichi nuclear disaster and amounts of radionuclides release to environment, so more effort and research are conducted to prevent it. MSR (Molten Salt Reactor) is one of GEN-IV reactor types, and its coolant and fuel are mixtures of molten salt. MSR has a schematic like figure 1 and it has different features with the solid fuel reactor, but most important and interesting feature of MSR is its many safety systems. For example, MSR has a large negative void coefficient. Even though power increases, the reactor slows down soon. Radionuclides release possibility of MSR was analyzed at various accident conditions including Chernobyl and Fukushima ones. The MSR was understood to prevent the severe accident by the negative reactivity coefficient and the absence of explosive material such as water at the Chernobyl disaster condition. It was expected to contain fuel salts in the reactor building and not to release radionuclides into environment even if the primary system could be ruptured or broken and fuel salts would be leaked at the Fukushima Daiichi nuclear disaster condition of earthquake and tsunami. The MSR, which would not lead to the severe accident and therefore prevents the fuel release to the environment at many expected scenarios, was thought to have priority in the aspect of accidents. A quantitative analysis and a further research are needed to evaluate the possibility of radionuclide release to the environment at the various accident conditions based on the simple comparison of the safety feature between MSR and solid fuel reactor.

  14. Analysis of Crew Fatigue in AIA Guantanamo Bay Aviation Accident

    Science.gov (United States)

    Rosekind, Mark R.; Gregory, Kevin B.; Miller, Donna L.; Co, Elizabeth L.; Lebacqz, J. Victor; Statler, Irving C. (Technical Monitor)

    1994-01-01

    Flight operations can engender fatigue, which can affect flight crew performance, vigilance, and mood. The National Transportation Safety Board (NTSB) requested the NASA Fatigue Countermeasures Program to analyze crew fatigue factors in an aviation accident that occurred at Guantanamo Bay, Cuba. There are specific fatigue factors that can be considered in such investigations: cumulative sleep loss, continuous hours of wakefulness prior to the incident or accident, and the time of day at which the accident occurred. Data from the NTSB Human Performance Investigator's Factual Report, the Operations Group Chairman's Factual Report, and the Flight 808 Crew Statements were analyzed, using conservative estimates and averages to reconcile discrepancies among the sources. Analysis of these data determined the following: the entire crew displayed cumulative sleep loss, operated during an extended period of continuous wakefulness, and obtained sleep at times in opposition to the circadian disposition for sleep, and that the accident occurred in the afternoon window of physiological sleepiness. In addition to these findings, evidence that fatigue affected performance was suggested by the cockpit voice recorder (CVR) transcript as well as in the captain's testimony. Examples from the CVR showed degraded decision-making skills, fixation, and slowed responses, all of which can be affected by fatigue; also, the captain testified to feeling "lethargic and indifferent" just prior to the accident. Therefore, the sleep/wake history data supports the hypothesis that fatigue was a factor that affected crewmembers' performance. Furthermore, the examples from the CVR and the captain's testimony support the hypothesis that the fatigue had an impact on specific actions involved in the occurrence of the accident.

  15. Analysis of hot leg natural circulation under station blackout severe accident

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Under severe accidents, natural circulation flows are important to influence the accident progression and result in a pressurized water reactor (PWR). In a station blackout accident with no recovery of steam generator (SG) auxiliary feedwater (TMLB' severe accident scenario), the hot leg countercurrent natural circulation flow is analyzed by using a severe-accident code, to better understand its potential impacts on the creep-rupture timing among the surge line, the hot leg, and SG tubes. The results show that the natural circulation may delay the failure time of the hot leg.The recirculation ratio and the hot mixing factor are also calculated and discussed.

  16. Preliminary accident analysis of Flexblue® underwater reactor

    Directory of Open Access Journals (Sweden)

    Haratyk Geoffrey

    2015-01-01

    Full Text Available Flexblue® is a subsea-based, transportable, small modular reactor delivering 160 MWe. Immersion provides the reactor with an infinite heat sink – the ocean – around the metallic hull. The reference design includes a loop-type PWR with two horizontal steam generators. The safety systems are designed to operate passively; safety functions are fulfilled without operator action and external electrical input. Residual heat is removed through four natural circulation loops: two primary heat exchangers immersed in safety tanks cooled by seawater and two emergency condensers immersed in seawater. In case of a primary piping break, a two-train safety injection system is actuated. Each train includes a core makeup tank, an accumulator and a safety tank at low pressure. To assess the capability of these features to remove residual heat, the reactor and its safety systems have been modelled using thermal-hydraulics code ATHLET with conservative assumptions. The results of simulated transients for three typical PWR accidents are presented: a turbine trip with station blackout, a large break loss of coolant accident and a small break loss of coolant accident. The analyses show that the safety criteria are respected and that the reactor quickly reaches a safe shutdown state without operator action and external power.

  17. Advanced accident sequence precursor analysis level 2 models

    Energy Technology Data Exchange (ETDEWEB)

    Galyean, W.J.; Brownson, D.A.; Rempe, J.L. [and others

    1996-03-01

    The U.S. Nuclear Regulatory Commission Accident Sequence Precursor program pursues the ultimate objective of performing risk significant evaluations on operational events (precursors) occurring in commercial nuclear power plants. To achieve this objective, the Office of Nuclear Regulatory Research is supporting the development of simple probabilistic risk assessment models for all commercial nuclear power plants (NPP) in the U.S. Presently, only simple Level 1 plant models have been developed which estimate core damage frequencies. In order to provide a true risk perspective, the consequences associated with postulated core damage accidents also need to be considered. With the objective of performing risk evaluations in an integrated and consistent manner, a linked event tree approach which propagates the front end results to back end was developed. This approach utilizes simple plant models that analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude and timing of a radioactive release to the environment, and calculate the consequences for a given release. Detailed models and results from previous studies, such as the NUREG-1150 study, are used to quantify these simple models. These simple models are then linked to the existing Level 1 models, and are evaluated using the SAPHIRE code. To demonstrate the approach, prototypic models have been developed for a boiling water reactor, Peach Bottom, and a pressurized water reactor, Zion.

  18. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for deposited material and external doses. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Boardman, J. [AEA Technology (United Kingdom); Jones, J.A. [National Radiological Protection Board (United Kingdom); Harper, F.T.; Young, M.L. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA deposited material and external dose models.

  19. A preliminary uncertainty analysis of phenomenological inputs employed in MAAP code using the SAUNA system

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. H.; Park, S. Y.; Kim, K. R.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    Uncertainty analysis is an essential element of safety analysis of nuclear power plants, and especially on the increase as an essential methodology of safety assessment by computer codes. Recently, these efforts have been stepped up to apply the uncertainty methodology in severe accident analysis and PSA Level 2. From this point of view, a statistical sampling-based MAAP-specific platform for a severe accident uncertainty analysis, SAUNA, is being developed in KAERI. Its main purpose is to execute many simulations that are employed for uncertainty analysis. For its efficient implementation, the SAUNA system is composed of three related modules: Firstly, a module for preparing a statistical sampling matrix, secondly, a module for the dynamic linking between code and samples for code simulation, and thirdly, a postprocessing module for further analysis of the code simulation results. The main objective of this paper is to introduce the main functions of the SAUNA system and its example of implementation.

  20. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, appendices A and B

    Energy Technology Data Exchange (ETDEWEB)

    Harper, F.T.; Young, M.L.; Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States); Lui, C.H. [Nuclear Regulatory Commission, Washington, DC (United States); Goossens, L.H.J.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Paesler-Sauer, J. [Research Center, Karlsruhe (Germany); Helton, J.C. [and others

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the second of a three-volume document describing the project and contains two appendices describing the rationales for the dispersion and deposition data along with short biographies of the 16 experts who participated in the project.

  1. A System Supporting the Analysis of Motorway Traffic Accidents

    Directory of Open Access Journals (Sweden)

    Davide Anghinolfi

    2015-12-01

    Full Text Available This work presents a business intelligence tool for monitoring traffic accidents on motorways and supporting decisions relevant to road safety. The system manages information on road characteristics, traffic accidents and traffic volumes and produces reports for monitoring the evolution of key performance indicators for road safety, supporting decisions on actions for risk mitigation and safety improvements for road users. The paper illustrates the different types of analyses performed by the system. Pattern based analysis is used to evaluate safety performance indicators for the road sections matching defined patterns. Two different road segmentation algorithms, used to identify the most critical road sections according to various severity indicators, are presented and discussed. Differential analysis compares the value of selected severity indicators before and after the implementation of an intervention on a road. Finally, a graphical user interface allows the accident locations to be visualized and accidents with specific characteristics to be highlighted. The system was evaluated on the data collected between 2009 and 2011 for the A15 motorway in Italy, connecting Parma to La Spezia.

  2. Analysis of multiple tank car releases in train accidents.

    Science.gov (United States)

    Liu, Xiang; Liu, Chang; Hong, Yili

    2017-10-01

    There are annually over two million carloads of hazardous materials transported by rail in the United States. The American railroads use large blocks of tank cars to transport petroleum crude oil and other flammable liquids from production to consumption sites. Being different from roadway transport of hazardous materials, a train accident can potentially result in the derailment and release of multiple tank cars, which may result in significant consequences. The prior literature predominantly assumes that the occurrence of multiple tank car releases in a train accident is a series of independent Bernoulli processes, and thus uses the binomial distribution to estimate the total number of tank car releases given the number of tank cars derailing or damaged. This paper shows that the traditional binomial model can incorrectly estimate multiple tank car release probability by magnitudes in certain circumstances, thereby significantly affecting railroad safety and risk analysis. To bridge this knowledge gap, this paper proposes a novel, alternative Correlated Binomial (CB) model that accounts for the possible correlations of multiple tank car releases in the same train. We test three distinct correlation structures in the CB model, and find that they all outperform the conventional binomial model based on empirical tank car accident data. The analysis shows that considering tank car release correlations would result in a significantly improved fit of the empirical data than otherwise. Consequently, it is prudent to consider alternative modeling techniques when analyzing the probability of multiple tank car releases in railroad accidents. Copyright © 2017 Elsevier Ltd. All rights reserved.

  3. INDUSTRIAL/MILITARY ACTIVITY-INITIATED ACCIDENT SCREENING ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    D.A. Kalinich

    1999-09-27

    Impacts due to nearby installations and operations were determined in the Preliminary MGDS Hazards Analysis (CRWMS M&O 1996) to be potentially applicable to the proposed repository at Yucca Mountain. This determination was conservatively based on limited knowledge of the potential activities ongoing on or off the Nevada Test Site (NTS). It is intended that the Industrial/Military Activity-Initiated Accident Screening Analysis provided herein will meet the requirements of the ''Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants'' (NRC 1987) in establishing whether this external event can be screened from further consideration or must be included as a design basis event (DBE) in the development of accident scenarios for the Monitored Geologic Repository (MGR). This analysis only considers issues related to preclosure radiological safety. Issues important to waste isolation as related to impact from nearby installations will be covered in the MGR performance assessment.

  4. Cold Vacuum Drying facility design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.

    2000-08-08

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

  5. Experimental assessment of computer codes used for safety analysis of integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)

    1995-09-01

    Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.

  6. Aspects of using a best-estimate approach for VVER safety analysis in reactivity initiated accidents

    Energy Technology Data Exchange (ETDEWEB)

    Ovdiienko, Iurii; Bilodid, Yevgen; Ieremenko, Maksym [State Scientific and Technical Centre on Nuclear and Radiation, Safety (SSTC N and RS), Kyiv (Ukraine); Loetsch, Thomas [TUEV SUED Industrie Service GmbH, Energie und Systeme, Muenchen (Germany)

    2016-09-15

    At present time, Ukraine faces the problem of small margins of acceptance criteria in connection with the implementation of a conservative approach for safety evaluations. The problem is particularly topical conducting feasibility analysis of power up-rating for Ukrainian nuclear power plants. Such situation requires the implementation of a best-estimate approach on the basis of an uncertainty analysis. For some kind of accidents, such as loss-of-coolant accident (LOCA), the best estimate approach is, more or less, developed and established. However, for reactivity initiated accident (RIA) analysis an application of best estimate method could be problematical. A regulatory document in Ukraine defines a nomenclature of neutronics calculations and so called ''generic safety parameters'' which should be used as boundary conditions for all VVER-1000 (V-320) reactors in RIA analysis. In this paper the ideas of uncertainty evaluations of generic safety parameters in RIA analysis in connection with the use of the 3D neutron kinetic code DYN3D and the GRS SUSA approach are presented.

  7. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    . The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. NASA Accident Precursor Analysis Handbook, Version 1.0

    Science.gov (United States)

    Groen, Frank; Everett, Chris; Hall, Anthony; Insley, Scott

    2011-01-01

    Catastrophic accidents are usually preceded by precursory events that, although observable, are not recognized as harbingers of a tragedy until after the fact. In the nuclear industry, the Three Mile Island accident was preceded by at least two events portending the potential for severe consequences from an underappreciated causal mechanism. Anomalies whose failure mechanisms were integral to the losses of Space Transportation Systems (STS) Challenger and Columbia had been occurring within the STS fleet prior to those accidents. Both the Rogers Commission Report and the Columbia Accident Investigation Board report found that processes in place at the time did not respond to the prior anomalies in a way that shed light on their true risk implications. This includes the concern that, in the words of the NASA Aerospace Safety Advisory Panel (ASAP), "no process addresses the need to update a hazard analysis when anomalies occur" At a broader level, the ASAP noted in 2007 that NASA "could better gauge the likelihood of losses by developing leading indicators, rather than continue to depend on lagging indicators". These observations suggest a need to revalidate prior assumptions and conclusions of existing safety (and reliability) analyses, as well as to consider the potential for previously unrecognized accident scenarios, when unexpected or otherwise undesired behaviors of the system are observed. This need is also discussed in NASA's system safety handbook, which advocates a view of safety assurance as driving a program to take steps that are necessary to establish and maintain a valid and credible argument for the safety of its missions. It is the premise of this handbook that making cases for safety more experience-based allows NASA to be better informed about the safety performance of its systems, and will ultimately help it to manage safety in a more effective manner. The APA process described in this handbook provides a systematic means of analyzing candidate

  10. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    Science.gov (United States)

    Reyes, S.; Latkowski, J. F.; Gomez del Rio, J.; Sanz, J.

    2001-05-01

    Previous studies of the safety and environmental aspects of the HYLIFE-II inertial fusion energy power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work, computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) have been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here we consider a severe loss of coolant accident (LOCA) in conjunction with simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the confinement) and of the two barriers surrounding the chamber (inner shielding and confinement building itself). Even though confinement failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product transport and release. The results of these calculations show that the estimated off-site dose is less than 5 mSv (0.5 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  11. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, S; Gomez del Rio, J; Sanz, J

    2000-02-23

    Previous studies of the safety and environmental (S and E) aspects of the HYLIFE-II inertial fusion energy (IFE) power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work a set of computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) has been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here the authors consider a severe lost of coolant accident (LOCA) producing simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the containment) and of the two barriers surrounding the chamber (inner shielding and containment building it self). Even though containment failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product release and transport. The results of these calculations show that the estimated off-site dose is less than 6 mSv (0.6 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  12. Feasibility Analysis on Simulation of PLCS Malfunction Event using SPACE Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ung Soo; Lee, Cheol Shin; Sohn, Jong Joo [KEPCO-E and C, Daejeon (Korea, Republic of)

    2011-10-15

    A computer code named 'Safety and Performance Analysis Code (SPACE)' has been being developed in order to replace several existing computer codes used in designing nuclear power plant (NPP) in Korea. This SPACE code is a system code and should be able to simulate various plant events, needed for safety analysis of pressurized water reactors (PWRs), such as loss of coolant accident (LOCA), steam line break (SLB), feedwater line break (FLB), steam generator tube rupture (SGTR), and several anticipated operational occurrences (AOOs). Therefore, respective simulations of above events with the SPACE code should be verified and validated to utilize this code in the safety analysis. In this work, a feasibility analysis is performed for the simulation of pressurizer level control system (PLCS) malfunction event for the Shin-Kori units 3 and 4 (SKN 3 and 4)

  13. Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gilles Youinou; R. Sonat Sen

    2013-09-01

    The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy – UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

  14. Flow Analysis of Code Customizations

    DEFF Research Database (Denmark)

    Hessellund, Anders; Sestoft, Peter

    2008-01-01

    Inconsistency between metadata and code customizations is a major concern in modern, configurable enterprise systems. The increasing reliance on metadata, in the form of XML files, and code customizations, in the form of Java files, has led to a hybrid development platform. The expected consisten...

  15. ASTEC adaptation for PHWR limited core damage accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, P., E-mail: pmajum@barc.gov.in [Bhabha Atomic Research Centre, Reactor Safety Division, Mumbai 400085 (India); Chatterjee, B.; Lele, H.G. [Bhabha Atomic Research Centre, Reactor Safety Division, Mumbai 400085 (India); Guillard, G.; Fichot, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SAG, Cadarache, 13115 Saint-Paul-lez-Durance (France)

    2014-06-01

    Under limited core damage accidents (LCDAs) of Pressurized Heavy Water Reactor (PHWR), coolable geometry of the channel might be retained thanks to the presence of moderator heat sink. Indeed, the pressure tube is amenable to creep deformation at high temperature due to internal pressure and fuel bundles weight. Partial or complete circumferential contact between pressure tube and calandria tube aids heat dissipation to the moderator. A new module has been developed by Bhabha Atomic Research Centre (BARC) for simulating this phenomenon which is specific to horizontal-type of reactors. It requires additional calculation of pressure tube sagging/ballooning and temperature field in the circumferential direction. The module is well validated with available experimental results concerning pressure tube deformation and the associated heat transfer in the area of contact. It is then used in analysing typical LCDAs scenarios in Indian PHWR under low and medium internal pressure conditions. This module is implemented in the ASTEC IRSN-GRS severe accident code version under development and will thus be available in the next major version V2.1.

  16. An establishment of MELCOR code to generate source terms for off site consequence analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. H.; Han, S.; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Since the Fukushima accident, an effective approach to a source term analysis for off site consequence analyses has been needed. The MELCOR code has the capability to assess the source term characteristics for this kind of demand. A comprehensive effort is required to use the MELCOR code for a source term analysis effectively. For this purpose, the following works are required: - Review and assess the MELCOR model relevant to source term characterization - Generate input files for source term analysis - Utilize the source term parameters This paper shows an effort to establish the MELCOR code to generate source terms for an off site consequence analysis.

  17. Response Analysis on Electrical Pulses under Severe Nuclear Accident Temperature Conditions Using an Abnormal Signal Simulation Analysis Module

    Directory of Open Access Journals (Sweden)

    Kil-Mo Koo

    2012-01-01

    Full Text Available Unlike design basis accidents, some inherent uncertainties of the reliability of instrumentations are expected while subjected to harsh environments (e.g., high temperature and pressure, high humidity, and high radioactivity occurring in severe nuclear accident conditions. Even under such conditions, an electrical signal should be within its expected range so that some mitigating actions can be taken based on the signal in the control room. For example, an industrial process control standard requires that the normal signal level for pressure, flow, and resistance temperature detector sensors be in the range of 4~20 mA for most instruments. Whereas, in the case that an abnormal signal is expected from an instrument, such a signal should be refined through a signal validation process so that the refined signal could be available in the control room. For some abnormal signals expected under severe accident conditions, to date, diagnostics and response analysis have been evaluated with an equivalent circuit model of real instruments, which is regarded as the best method. The main objective of this paper is to introduce a program designed to implement a diagnostic and response analysis for equivalent circuit modeling. The program links signal analysis tool code to abnormal signal simulation engine code not only as a one body order system, but also as a part of functions of a PC-based ASSA (abnormal signal simulation analysis module developed to obtain a varying range of the R-C circuit elements in high temperature conditions. As a result, a special function for abnormal pulse signal patterns can be obtained through the program, which in turn makes it possible to analyze the abnormal output pulse signals through a response characteristic of a 4~20 mA circuit model and a range of the elements changing with temperature under an accident condition.

  18. Simulation of containment pressurization in a large break-loss of coolant accident using single-cell and multicell models and CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Kalkahoran, Omid Noori; Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

  19. Analysis of Three Mile Island-Unit 2 accident

    Energy Technology Data Exchange (ETDEWEB)

    1980-03-01

    The Nuclear Safety Analysis Center (NSAC) of the Electric Power Research Institute has analyzed the Three Mile Island-2 accident. Early results of this analysis were a brief narrative summary, issued in mid-May 1979 and an initial version of this report issued later in 1979 as noted in the Foreword. The present report is a revised version of the 1979 report, containing summaries, a highly detailed sequence of events, a comparison of that sequence of events with those from other sources, 25 appendices, references and a list of abbreviations and acronyms. A matrix of equipment and system actions is included as a folded insert.

  20. Traffic accident analysis using GIS: a case study of Kyrenia City

    Science.gov (United States)

    Kara, Can; Akçit, Nuhcan

    2015-06-01

    Traffic accidents are causing major deaths in urban environments, so analyzing locations of the traffic accidents and their reasons is crucial. In this manner, patterns of accidents and hotspot distribution are analyzed by using geographic information technology. Locations of the traffic accidents in the years 2011, 2012 and 2013 are combined to generate the kernel distribution map of Kyrenia City. This analysis aims to find high dense intersections and segments within the city. Additionally, spatial autocorrelation methods Local Morans I and Getis-Ord Gi are employed . The results are discussed in detail for further analysis. Finally, required changes for numerous intersections are suggested to decrease potential risks of high dense accident locations.

  1. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

  2. [Hospital information system performance for road traffic accidents analysis in a hospital recruitment based area].

    Science.gov (United States)

    Jannot, A-S; Fauconnier, J

    2013-06-01

    Road traffic accidents in France are mainly analyzed through reports completed by the security forces (police and gendarmerie). But the hospital information systems can also identify road traffic accidents via specific documentary codes of the International Classification of Diseases (ICD-10). The aim of this study was therefore to determine whether hospital stays consecutive to road traffic accident were truly identified by these documentary codes in a facility that collects data routinely and to study the consistency of results from hospital information systems and from security forces during the 2002-2008 period. We retrieved all patients for whom a documentary code for road traffic accident was entered in 2002-2008. We manually checked the concordance of documentary code for road traffic accident and trauma origin in 350 patient files. The number of accidents in the Grenoble area was then inferred by combining with hospitalization regional data and compared to the number of persons injured by traffic accidents declared by the security force. These hospital information systems successfully report road traffic accidents with 96% sensitivity (95%CI: [92%, 100%]) and 97% specificity (95%CI: [95%, 99%]). The decrease in road traffic accidents observed was significantly less than that observed was significantly lower than that observed in the data from the security force (45% for security force data against 27% for hospital data). Overall, this study shows that hospital information systems are a powerful tool for studying road traffic accidents morbidity in hospital and are complementary to security force data. Copyright © 2013 Elsevier Masson SAS. All rights reserved.

  3. KSTAR Severe Accident Analysis using MELCOR : Ex-vessel Coolant Pipe Break with Failure of Fusion Power Termination System

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-10-15

    To investigate the consequence of severe accidents in fusion reactor, a number of thermal hydraulics simulation codes were used (ECART, INTRA, ATHENA/RELAP and so on). MELCOR is chosen as the thermal hydraulics code to simulate the consequence of radioactive material release from accident in preliminary safety report. Capability of the simulation code for fusion reactor severe accident analysis is ability to simulate the hydraulic system in ITER and the transport phenomenon of radionuclides. MELCOR is a fully integrated code that models the accidents in Light Water Reactor (LWR). There are three kinds of radioactive materials in fusion reactor; tritium (or Tiritiated water: HTO), activation products (AP) of divertor or first-wall and activated corrosion products(ACP). In generic Site Safety Report (GSSR), the release guidelines for tritium and activation products are listed for normal operation, incidents, and accidents. And this guidelines presented in Table 1. Not only ITER, the KSTAR (Korea Superconducting Tokamak Advanced Research) is also developing fusion research reactor. The scale of facility is smaller than ITER but this small scale of facility offers the experimental flexibility to develop fusion technology. The major differences between KSTAR and ITER systems are presented in Table 2. Fusion source difference between KSTAR and ITER is D-D fusion reaction (Deuterium-Deuterium fusion reaction) and D-T fusion reaction (Deuterium-Tritium fusion reaction). This D-D fusion makes one tritium by 50 percent chance. The radioactivity of tritium is small to consider compared to radioactive materials in nuclear fission reactor. This reaction is presented in equation (1) In the present work, conservatively estimated tritium inventory amount in KSTAR is used with one of the most severe accident in ITER; Ex-vessel pipe break with Fusion Power Termination System (FPTS). The MELCOR KSTAR input is made by scaling down the ITER input deck. So, the detail system is not same

  4. Station Blackout Severe Accident Analysis of Spent Fuel Pool of 600 MWe NPP by Using MELCOR Code%用 MELCOR 程序分析600 MWe 核电厂乏燃料水池失去厂内外电源严重事故

    Institute of Scientific and Technical Information of China (English)

    张应超; 季松涛; 魏严凇; 史晓磊; 许倩

    2016-01-01

    Using MELCOR code ,the spent fuel pool (SFP) of 600 MWe nuclear power plant (NPP) was modeled ,and the station blackout severe accidents were calculated when the SFP was under normal condition ,refuelling condition and the reactor accident condition .The calculation results show that fuel assemblies will melt down and hydro‐gen will generate ,due to zirconium‐water reaction ,after the half height of fuel assem‐blies is uncovered .The influence of injection or spray on SFP accidents was analysed , and the results show that SFP accidents will be terminated and the water level of SFP will return up before fuel cladding damage if water is injected or sprayed into the SFP with the boiling evaporation mass rate .%利用MELCOR程序建立了600 MWe核电厂乏燃料水池计算模型,分别计算了在正常储存、正常换料和反应堆事故工况下,乏燃料水池失去厂内外电源严重事故序列。计算结果表明,燃料组件大约裸露一半后,锆水反应导致燃料熔化并产生大量氢气。分析了喷淋和注水对乏燃料水池事故的影响,分析结果表明,在燃料包壳失效前,以沸腾蒸发速率注水或喷淋能中止事故发展,并能使乏燃料水池水位缓慢回升。

  5. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Perin, Y.; Velkov, K. [Gesellschaft flier Anlagen- und Reaktorsicherheit - GRS mbH, Boltzmannstasse 14, 85748 Garching bei Muenchen (Germany)

    2013-07-01

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  6. Source term analysis for a nuclear submarine accident

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B.J.; Hugron, J.J.M.R. [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    1999-07-01

    A source term analysis has been conducted to determine the activity release into the environment as a result of a large-break loss-of-coolant accident aboard a visiting nuclear-powered submarine to a Canadian port. This best-estimate analysis considers the fractional release from the core, and fission product transport in the primary heat transport system, primary containment (i.e. reactor compartment) and submarine hull. Physical removal mechanisms such as vapour and aerosol deposition are treated in the calculation. Since a thermalhydraulic analysis indicated that the integrity of the reactor compartment is maintained, release from the reactor compartment will only occur by leakage; however, it is conservatively assumed that the secondary containment is not isolated for a 24-h period where release occurs through an open hatch in the submarine hull. Consequently, during this period, the activity release into the atmosphere is estimated as 4.6 TBq, leading to a maximum individual dose equivalent of 0.5 mSv at 800 metres from the berthing location. This activity release is comparable to that obtained in the BEREX TSA study (for a similar accident scenario) but is four orders of magnitude less than that reported in the earlier Davis study where, unrealistically, no credit had been taken for the containment system or for any physical removal processes. (author)

  7. Statistical analysis of sudden chemical leakage accidents reported in China between 2006 and 2011.

    Science.gov (United States)

    Li, Yang; Ping, Hua; Ma, Zhi-Hong; Pan, Li-Gang

    2014-04-01

    According to the data from authoritative sources, 1,400 sudden leakage accidents occurred in China during 2006 to 2011 were investigated, in which, 666 accidents were used for statistical characteristic abstracted with no or little damage. The research results were as follows: (1) Time fluctuation: the yearly number of sudden leakage accidents is shown to be decreasing from 2006 to 2010, and a slightly increase in 2011. Sudden leakage accidents occur mainly in summer, and more than half of the accidents occur from May to September. (2) Regional distribution: the accidents are highly concentrated in the coastal area, in which accidents result from small and medium-sized enterprises more easily than that of the larger ones. (3) Pollutants: hazardous chemicals are up to 95 % of sudden leakage accidents. (4) Steps: transportation represents almost half of the accidents, followed by production, usage, storage, and discard. (5) Pollution and casualties: it is easy to cause environmental pollution and casualties. (6) Causes: more than half of the cases were caused by human factor, followed by management reason, and equipment failure. However, sudden chemical leakage may also be caused by high temperature, rain, wet road, and terrain. (7) The results of principal component analysis: five factors are extracted by the principal component analysis, including pollution, casualties, regional distribution, steps, and month. According to the analysis of the accident, the characteristics, causes, and damages of the sudden leakage accident will be investigated. Therefore, advices for prevention and rescue should be acquired.

  8. Estimating the causes of traffic accidents using logistic regression and discriminant analysis.

    Science.gov (United States)

    Karacasu, Murat; Ergül, Barış; Altin Yavuz, Arzu

    2014-01-01

    Factors that affect traffic accidents have been analysed in various ways. In this study, we use the methods of logistic regression and discriminant analysis to determine the damages due to injury and non-injury accidents in the Eskisehir Province. Data were obtained from the accident reports of the General Directorate of Security in Eskisehir; 2552 traffic accidents between January and December 2009 were investigated regarding whether they resulted in injury. According to the results, the effects of traffic accidents were reflected in the variables. These results provide a wealth of information that may aid future measures toward the prevention of undesired results.

  9. Development of RBMK-1500 Model for BDBA Analysis Using RELAP/SCDAPSIM Code

    Science.gov (United States)

    Uspuras, Eugenijus; Kaliatka, Algirdas

    This article discusses the specificity of RBMK (channel type, boiling water, graphite moderated) reactors and problems of Reactor Cooling System modelling employing computer codes. The article presents, how the RELAP/SCDAPSIM code, which is originally designed for modelling of accidents in vessel type reactors, is fit to simulate the phenomena in the RBMK reactor core and RCS in case of Beyond Design Basis Accidents. For this reason, use of two RELAP/SCDAPSIM models is recommended. First model with described complete geometry of RCS is recommended for analysis of initial phase of accident. The calculations results, received using this model, are used as boundary conditions in simplified model for simulation of later phases of severe accidents. The simplified model was tested comparing results of simulation performed using RELAP5 and RELAP/SCDAPSIM codes. As the typical example of BDBA, large break LOCA in reactor cooling system with failure of emergency core cooling system was analyzed. Use of developed models allows to receive behaviour of thermal-hydraulic parameters, temperatures of core components, amount of generated hydrogen due to steam-zirconium reaction. These parameters will be used as input for other special codes, designed for analysis of processes in reactor containment.

  10. [Spatial exploratory analysis of road accidents in Ciudad Juarez, Mexico].

    Science.gov (United States)

    Hernández Hernández, Vladimir

    2012-05-01

    Prepare a tool for the exploratory study of road accidents in Ciudad Juarez, Chihuahua, Mexico, that exclusively applies the spatial geographical variable (location). Observational and cross-sectional study that uses a Geographic Information System to explore the spatial nature of 13 305 road accidents recorded during 2008 and 2009 in Ciudad Juarez. Indicators were constructed that approximated the transit flow and included two variables: indices of the level of urbanization and population density. The value of the global spatial autocorrelation was positive, indicating the presence of groupings that were identified through the spatial association indicators. There are road risk clusters located in areas with a high level of urbanization, low population density, and a high transit flow level. The exploratory analysis of spatial data is a phase that precedes the use of multivariate techniques with a broader scope. The application of exploratory analysis techniques in itself makes it possible to standardize spatial groupings, identify global autocorrelation, and indicate the direction of the variables under study.

  11. Comprehensive analysis of atmospheric radionuclides just after the Fukushima accident

    Science.gov (United States)

    Tsuruta, Haruo; Oura, Yasuji; Ebihara, Mitsuru; Ohara, Toshimasa; Moriguchi, Yuichi; Nakajima, Teruyuki

    2017-04-01

    Even six years passed after the Fukushima Daiichi Nuclear Power Plant (FD1NPP) accident, we still have large uncertainty for atmospheric transport and deposition models, the estimate of release rate of source terms and of internal exposure from inhalation. For our better understanding and to reduce the uncertainty, we thoroughly analyzed all the published data of radionuclides such as Cs-137, I-131 and Xe-133, and of radiation dose rates at many monitoring sites in eastern Japan. We also retrieved the spatio-temporal distributions of Cs-137 just after the accident by using the unique dataset of hourly radionuclides in atmospheric aerosols collected on the used filter-tapes installed in the suspended particulate matter (SPM) monitors operated at more than 100 stations in the air pollution monitoring network of Japan. The most important findings are summarized as follows. Analyzing the hourly Cs-137 concentrations at two SPM stations located within 20 km from the FD1NPP, we revealed the complicated behavior of plumes and atmospheric radionuclides near the FD1NPP just after the accident. The transport pathways to the northwestern and northern areas from the FD1NPP are clarified especially on March 12-21, 2011. Analysis of the published data clearly shows that atmospheric ratio of I-131/Cs-137 (=R) was mainly divided into two groups, one (R≦10) is for the plumes before March 21, 2011, and the other (R>100) is after that day. These two groups are consistent in all the measured sites, whether the sites are in the Fukushima prefecture or in the Tokyo Metropolitan area. These results are expected partially to identify the source term for each plume.

  12. Analysis of surface powered haulage accidents, January 1990--July 1996

    Energy Technology Data Exchange (ETDEWEB)

    Fesak, G.M.; Breland, R.M.; Spadaro, J. [Dept. of Labor, Arlington, VA (United States)

    1996-12-31

    This report addresses surface haulage accidents that occurred between January 1990 and July 1996 involving haulage trucks (including over-the-road trucks), front-end-loaders, scrapers, utility trucks, water trucks, and other mobile haulage equipment. The study includes quarries, open pits and surface coal mines utilizing self-propelled mobile equipment to transport personnel, supplies, rock, overburden material, ore, mine waste, or coal for processing. A total of 4,397 accidents were considered. This report summarizes the major factors that led to the accidents and recommends accident prevention methods to reduce the frequency of these accidents.

  13. An analysis on the severe accident progression with operator recovery actions

    Energy Technology Data Exchange (ETDEWEB)

    Vo, T.H. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejon 305-353 (Korea, Republic of); Korea University of Science and Technology (UST), 217 Gajeong-ro, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Song, J.H., E-mail: dosa@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejon 305-353 (Korea, Republic of); Korea University of Science and Technology (UST), 217 Gajeong-ro, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Kim, T.W.; Kim, D.H. [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejon 305-353 (Korea, Republic of)

    2014-12-15

    Highlights: • Severe accident progression for the station blackout and SBLOCA accident. • Analyses on APR1400 using MELCOR. • Operator recovery actions for decay heat removal and inventory make up. • Determine the time allowed for the operator to prevent reactor vessel failure. • Insight for the operator recovery actions for the severe accident management. - Abstract: Analyses on the severe accident progressions for the station blackout (SBO) accident and small break LOCA (SBLOCA) initiated severe accident were performed for APR1400 by using MELCOR computer code. Operator recovery actions for decay heat removal and inventory make up using a depressurization system and safety injection pump were simulated in parallel with a simulation of the severe accident progression. Sensitivity studies on the operator actions were performed to investigate the changes in the timing of the reactor vessel failure and to determine the time allowed for the operator to prevent reactor vessel failure. Sensitivity analyses on the effect of major modeling parameters were performed additionally to quantify the uncertainties in timing. It is found that the operator has about 2 h for the recovery actions after the indication of core damage by the signal of core exit thermocouple (CET) for the SBLOCA initiated severe accident, while the operator has to take immediate actions after the indication of core damage by CET for the SBO accident.

  14. Improvement of severe accident analysis method for KSNP

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Cho, Song Won; Cho, Youn Soo [Korea Radiation Technology Institute Co., Taejon (Korea, Republic of)

    2002-03-15

    The objective of this study is preparation of MELCOR 1.8.5 input deck for KSNP and simulation of some major severe accidents. The contents of this project are preparation of MELCOR 1.8.5 base input deck for KSNP to understand severe accident phenomena and to assess severe accident strategy, preparation of 20 cell containment input deck to simulate the distribution of hydrogen and fission products in containment, simulation of some major severe accident scenarios such as TLOFW, SBO, SBLOCA, MBLOCA, and LBLOCA. The method for MELCOR 1.8.5 input deck preparation can be used to prepare the input deck for domestic PWRs and to simulate severe accident experiments such as ISP-46. Information gained from analyses of severe accidents may be helpful to set up the severe accident management strategy and to develop regulatory guidance.

  15. Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramsthaler, J. A.; Lime, J. F.; Sahota, M. S.

    1978-12-01

    A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A.

  16. Analysis of simulation results of damaged nuclear fuel accidents at NPPs with shell-type nuclear reactors

    Directory of Open Access Journals (Sweden)

    Igor L. Kozlov

    2015-03-01

    Full Text Available Lessons from the accident at the Fukushima Daiichi NPP made it necessary to reevaluate and intensificate the work on modeling and analyzing various scenarios of severe accidents with damage to the nuclear fuel in the reactor, containment and spent nuclear fuel storage pool with the expansion of the primary initiating event causes group listing. Further development of computational tools for modeling the explosion prevention criteria as to steam and gas mixtures, considering the specific thermal-hydrodynamic conditions and mechanisms of explosive situations arrival at different stages of a severe accident development, is substantiated. Based on the analysis of the known shell-type nuclear reactors accidents results the explosion safety thermodynamic criteria are presented, the parameters defining the steam and gas explosions conditions are found, the need to perform the further verification and validation of deterministic codes serving to simulate general accident processes behavior as well as phase-to-phase interaction calculated dependencies is established. The main parameters controlling and defining the criteria explosion safety effective regulation areas and their optimization conditions are found.

  17. An Accident Precursor Analysis Process Tailored for NASA Space Systems

    Science.gov (United States)

    Groen, Frank; Stamatelatos, Michael; Dezfuli, Homayoon; Maggio, Gaspare

    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 and which may differ in frequency or type from those in the various models. These discrepancies between the models (perceived risk) and the system (actual risk) provide the leading indication of an underappreciated risk. This paper presents an APA process developed specifically for NASA Earth-to-Orbit space systems. The purpose of the process is to identify and characterize potential sources of system risk as evidenced by anomalous events which, although not necessarily presenting an immediate safety impact, may indicate that an unknown or insufficiently understood risk-significant condition exists in the system. Such anomalous events are considered accident precursors because they signal the potential for severe consequences that may occur in the future, due to causes that are discernible from their occurrence today. Their early identification allows them to be integrated into the overall system risk model used to intbrm decisions relating to safety.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-04-01

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

  19. Expert system interaction with existing analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Ransom, V.H.; Fink, R.K.; Bertch, W.J.; Callow, R.A.

    1986-01-01

    Coupling expert systems with existing engineering analysis codes is a promising area in the field of artificial intelligence. The added intelligence can provide for easier and less costly use of the code and also reduce the potential for code misuse. This paper will discuss the methods available to allow interaction between an expert system and a large analysis code running on a mainframe. Concluding remarks will identify potential areas of expert system application with specific areas that are being considered in a current research program. The difficulty of interaction between an analysis code and an expert system is due to the incompatibility between the FORTRAN environment used for the analysis code and the AI environment used for the expert system. Three methods, excluding file transfer techniques, are discussed to help overcome this incompatibility. The first method is linking the FORTRAN routines to the LISP environment on the same computer. Various LISP dialects available on mainframes and their interlanguage communication capabilities are discussed. The second method involves network interaction between a LISP machine and a mainframe computer. Comparisons between the linking method and networking are noted. The third method involves the use of an expert system tool that is campatible with a FORTRAN environment. Several available tools are discussed. With the interaction methods identified, several potential application areas are considered. Selection of the specific areas that will be developed for the pilot project and applied to a thermal-hydraulic energy analysis code are noted.

  20. Stability Analysis of the EBR-I Mark-II Core Meltdown Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Jae-Yong; Kang, Chang Mu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The purpose of this paper is to analyze the stability of the EBR-I core meltdown accident using the NuSTAB code. The result of NuSTAB analysis is compared with previous stability analysis by Sandmeier using the root locus method. The Experimental Breeder Reactor I (EBR-1) at Argonne National Laboratory was designed to demonstrate fast reactor breeding and to prove the use of liquid-metal coolant for power production and reached criticality in August 1951. The EBR-I reactor was undergoing a series of physics experiments and the Mark-II core was melted accidentally on Nov. 29, 1955. The experiment was going to increase core temperature to 500C to see if the reactor loses reactivity, and scram when the power reached 1500 kW or doubling of fission rate per second. However the operator scrammed with a slow moving control and missed the shutdown by two seconds and caused the core meltdown. The NuSTAB code has an advantage of analyzing space-dependent fast reactors and predicting regional oscillations compared to the point kinetics. Also, NuSTAB can be useful when the coupled neutronic-thermal-hydraulic codes cannot be used for stability analysis. Future work includes analyses of the PGSFR for various operating conditions as well as further validation of the NuSTAB calculations against SFR stability experiments when such experiments become available.

  1. Fluid-structure interaction analysis of a hypothetical core disruptive accident in LMFBRs

    Energy Technology Data Exchange (ETDEWEB)

    Liu Chuang [Department of Engineering Mechanics, Tsinghua University, Beijing 100084 (China)]. E-mail: lch98@mails.tsinghua.edu.cn; Zhang Xiong [Department of Engineering Mechanics, Tsinghua University, Beijing 100084 (China); Lu Mingwan [Department of Engineering Mechanics, Tsinghua University, Beijing 100084 (China)

    2005-03-01

    To ensure safety, it is necessary to assess the integrity of a reactor vessel of liquid-metal fast breeder reactor (LMFBR) under HCDA. Several important problems for a fluid-structural interaction analysis of HCDA are discussed in the present paper. Various loading models of hypothetical core disruptive accident (HCDA) are compared and the polytropic processes of idea gas (PPIG) law is recommended. In order to define a limited total energy release, a '5% truncation criterion' is suggested. The relationship of initial pressure of gas bubble and the total energy release is given. To track the moving interfaces and to avoid the severe mesh distortion an arbitrary Lagrangrian-Eulerian (ALE) approach is adopted in the finite element modeling (FEM) analysis. Liquid separation and splash from a free surface are discussed. By using an elasticity solution under locally uniform pressure, two simplified analytical solutions for 3D and axi-symmetric case of the liquid impact pressure on roof slab are derived. An axi-symmetric finite elements code FRHCDA for fluid-structure interaction analysis of hypothetical core disruptive accident in LMFBR is developed. The CONT benchmark problem is calculated. The numerical results agree well with those from published papers.

  2. Simulation of Containment Pressurization in a Large Break-Loss of Coolant Accident Using Single-Cell and Multicell Models and CONTAIN Code

    Directory of Open Access Journals (Sweden)

    Omid Noori-Kalkhoran

    2016-10-01

    Full Text Available Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model. In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code’s results.

  3. Spectral Analysis Code: PARAS SPEC

    CERN Document Server

    Chaturvedi, Priyanka; Anandarao, B G

    2016-01-01

    The light emitted from the stellar photosphere serves as a unique signature for the nature of stars. The behaviour of these stellar lines depend upon the surface temperature, mass, evolutionary status and chemical composition of the star. With the advent of high-resolution spectrographs coupled with medium to large aperture telescopes around the globe, there is plenty of high-resolution and high signal-to-noise ratio data available to the astronomy community. Apart from radial velocity (RV) studies, such data offer us the unique opportunity to study chemical composition and atmospheric properties of the star. The procedure used to derive these parameters must be automated and well adaptable to data available from any high-resolution spectrograph. We hereby present an IDL code, PARAS SPEC, which was primary designed to handle high-resolution spectroscopy data from PARAS spectrograph coupled with the 1.2~m telescope at Mt. Abu, India. This code is designed to adapt with data from other spectrographs as well. Th...

  4. An analysis of accident data for franchised public buses in Hong Kong.

    Science.gov (United States)

    Evans, W A; Courtney, A J

    1985-10-01

    This paper analyses data on accidents involving franchised public buses operating in Hong Kong. The data were obtained from the Royal Hong Kong Police, the Hong Kong Government Transport Department, the two major franchised bus operators and international sources. The analysis includes an international comparison of accidents with emphasis on the situation in Hong Kong compared to urban areas in the United Kingdom. An attempt has been made to identify the characteristics of bus accidents; accident incidence has been related to time of day, day of the week, time of year, weather conditions, driver's age and experience, hours on duty and policy-reported cause. The results indicate that Hong Kong has a high accident rate compared to Japan, the U.K. and the U.S.A., with particularly high pedestrian involvement rates. Bus accidents peak at around 9:00 AM and 4:00 PM but the accident rate is high throughout the day. Monday and Saturday appear to have a higher than average accident rate. The variability of accident rate throughout the year does not seem to be significant and the accident rate does not appear to be influenced by weather conditions. Older, more experienced drivers generally have a safer driving record than their younger, less experienced colleagues. Accident occurrence is related to the time the driver has been on duty. The paper questions the reliability of police-reported accident causation data and suggests improvements in the design of the accident report form and in the training of police investigators. The relevance of the Hong Kong study for accident research in general is also discussed.

  5. Comparison of Management Oversight and Risk Tree and Tripod-Beta in Excavation Accident Analysis

    Directory of Open Access Journals (Sweden)

    Mohamadfam

    2015-01-01

    Full Text Available Background Accident investigation programs are a necessary part in identification of risks and management of the business process. Objectives One of the most important features of such programs is the analysis technique for identifying the root causes of accidents in order to prevent their recurrences. Analytical Hierarchy Process (AHP was used to compare management oversight and risk tree (MORT with Tripod-Beta in order to determine the superior technique for analysis of fatal excavation accidents in construction industries. Materials and Methods MORT and Tripod-Beta techniques were used for analyzing two major accidents with three main steps. First, these techniques were applied to find out the causal factors of the accidents. Second, a number of criteria were developed for the comparison of the techniques and third, using AHP, the techniques were prioritized in terms of the criteria for choosing the superior one. Results The Tripod-Beta investigation showed 41 preconditions and 81 latent causes involved in the accidents. Additionally, 27 root causes of accidents were identified by the MORT analysis. Analytical hierarchy process (AHP investigation revealed that MORT had higher priorities only in two criteria than Tripod-Beta. Conclusions Our findings indicate that Tripod-Beta with a total priority of 0.664 is superior to MORT with the total priority of 0.33. It is recommended for future research to compare the available accident analysis techniques based on proper criteria to select the best for accident analysis.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-04-01

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

  7. Exploring the potential of data mining techniques for the analysis of accident patterns

    DEFF Research Database (Denmark)

    Prato, Carlo Giacomo; Bekhor, Shlomo; Galtzur, Ayelet

    2010-01-01

    Research in road safety faces major challenges: individuation of the most significant determinants of traffic accidents, recognition of the most recurrent accident patterns, and allocation of resources necessary to address the most relevant issues. This paper intends to comprehend which data mining...... and association rules) data mining techniques are implemented for the analysis of traffic accidents occurred in Israel between 2001 and 2004. Results show that descriptive techniques are useful to classify the large amount of analyzed accidents, even though introduce problems with respect to the clear...... importance of input and intermediate neurons, and the relative importance of hundreds of association rules. Further research should investigate whether limiting the analysis to fatal accidents would simplify the task of data mining techniques in recognizing accident patterns without the “noise” probably...

  8. The accident analysis of mobile mine machinery in Indian opencast coal mines.

    Science.gov (United States)

    Kumar, R; Ghosh, A K

    2014-01-01

    This paper presents the analysis of large mining machinery related accidents in Indian opencast coal mines. The trends of coal production, share of mining methods in production, machinery deployment in open cast mines, size and population of machinery, accidents due to machinery, types and causes of accidents have been analysed from the year 1995 to 2008. The scrutiny of accidents during this period reveals that most of the responsible factors are machine reversal, haul road design, human fault, operator's fault, machine fault, visibility and dump design. Considering the types of machines, namely, dumpers, excavators, dozers and loaders together the maximum number of fatal accidents has been caused by operator's faults and human faults jointly during the period from 1995 to 2008. The novel finding of this analysis is that large machines with state-of-the-art safety system did not reduce the fatal accidents in Indian opencast coal mines.

  9. Analysis of LMFBR containment response to an HCDA using a multifield Eulerian code. [MICE code

    Energy Technology Data Exchange (ETDEWEB)

    Chu, H.Y.; Chang, Y.W.

    1977-01-01

    During a hypothetical core disruptive accident (HCDA), a core meltdown may cause the fuel cladding to rupture and the fuel fragments to penetrate into the sodium coolant. The heat in the molten fuel may cause the liquid sodium to boil, changing its phase. The interactions between materials are so complicated that a single-material model with homogenized material properties is not adequate. In order to analyze the above phenomena more realistically, a Multifield Implicit Continuous-Fluid Eulerian containment code (MICE) is being developed at Argonne National Laboratory (ANL) to solve the multifield fluid-flow problems in which the interpenetrations of materials, heat transfer, and phase changes are considered in the analysis. The hydrodynamics of the MICE code is based upon the implicit multifield (IMF) method developed by Harlow and Amsden. A partial donor-cell formulation is used for the calculation of the convective fluxes to minimize the truncation errors, while the Newton-Raphson method is used for the numerical iterations. An implicit treatment of the mass convection together with the equation of state for each material enables the method to be applicable to both compressible and incompressible flows. A partial implicit treatment of the momentum-exchange functions allows the coupling drag forces between two material fields to range from very weak to those strong enough to tie the fields completely. The differential equations and exchange functions used in the MICE code, and the treatment of the fluid and structure interactions as well as the numerical procedure are described. Two sample calculations are given to illustrate the present capability of the MICE code.

  10. Aircraft Accident Prevention: Loss-of-Control Analysis

    Science.gov (United States)

    Kwatny, Harry G.; Dongmo, Jean-Etienne T.; Chang, Bor-Chin; Bajpai, Guarav; Yasar, Murat; Belcastro, Christine M.

    2009-01-01

    The majority of fatal aircraft accidents are associated with loss-of-control . Yet the notion of loss-of-control is not well-defined in terms suitable for rigorous control systems analysis. Loss-of-control is generally associated with flight outside of the normal flight envelope, with nonlinear influences, and with an inability of the pilot to control the aircraft. The two primary sources of nonlinearity are the intrinsic nonlinear dynamics of the aircraft and the state and control constraints within which the aircraft must operate. In this paper we examine how these nonlinearities affect the ability to control the aircraft and how they may contribute to loss-of-control. Examples are provided using NASA s Generic Transport Model.

  11. Radionuclide analysis on bamboos following the Fukushima nuclear accident.

    Directory of Open Access Journals (Sweden)

    Takumi Higaki

    Full Text Available In response to contamination from the recent Fukushima nuclear accident, we conducted radionuclide analysis on bamboos sampled from six sites within a 25 to 980 km radius of the Fukushima Daiichi nuclear power plant. Maximum activity concentrations of radiocesium (134Cs and (137Cs in samples from Fukushima city, 65 km away from the Fukushima Daiichi plant, were in excess of 71 and 79 kBq/kg, dry weight (DW, respectively. In Kashiwa city, 195 km away from the Fukushima Daiichi, the sample concentrations were in excess of 3.4 and 4.3 kBq/kg DW, respectively. In Toyohashi city, 440 km away from the Fukushima Daiichi, the concentrations were below the measurable limits of up to 4.5 Bq/kg DW. In the radiocesium contaminated samples, the radiocesium activity was higher in mature and fallen leaves than in young leaves, branches and culms.

  12. Analysis on relation between safety input and accidents

    Institute of Scientific and Technical Information of China (English)

    YAO Qing-guo; ZHANG Xue-mu; LI Chun-hui

    2007-01-01

    The number of safety input directly determines the level of safety, and there exists dialectical and unified relations between safety input and accidents. Based on the field investigation and reliable data, this paper deeply studied the dialectical relationship between safety input and accidents, and acquired the conclusions. The security situation of the coal enterprises was related to the security input rate, being effected little by the security input scale, and build the relationship model between safety input and accidents on this basis, that is the accident model.

  13. Analysis of Waste Leak and Toxic Chemical Release Accidents from Waste Feed Delivery (WFD) Diluent System

    Energy Technology Data Exchange (ETDEWEB)

    WILLIAMS, J.C.

    2000-09-15

    Radiological and toxicological consequences are calculated for 4 postulated accidents involving the Waste Feed Delivery (WFD) diluent addition systems. Consequences for the onsite and offsite receptor are calculated. This analysis contains technical information used to determine the accident consequences for the River Protection Project (RPP) Final Safety Analysis Report (FSAR).

  14. Development of a post accident analysis model for KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Chang, W. P.; Ha, G. S.; Jeong, H. Y.; Kwon, Y. M.; Heo, S.; Lee, Y. B. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    An ultimate safety measure of the KALIMER depends on the inherent safety, which have the core maintain a negative reactivity during any accident periods. In order to secure the integrity of a fuel rod, the void reactivity feedback under sodium boiling must be analyzed. Even though the KALIMER design might not allow boiling at any circumstance, sodium boiling would be possible under HCDA (Hypothetical Core Disruptive Accident) initiating events which are represented by UTOP (Unprotected Transient Over Power), ULOF (Unprotected Loss Of Flow), ULOHS (Unprotected Loss Of Heat Sink), or sudden flow channel blockage, due to power excursion caused by the reactivity feedback. The slug and annular flow regimes tend to prevail for the boiling of a liquid-metal coolant such as sodium near the atmospheric pressure. In contrast, the bubbly flow is typical under a high pressure in light water reactors. This phenomenon difference brings to develop the present model, especially, at the onset of boiling. A few models had been developed for the sodium boiling analysis. The models such as those in the HOMSEP-2 and SAS series are classified into relatively detailed models. Both models are usually called a multiple-bubble slug ejection model. Some simpler models are also introduced to evade either parameter sensitivities or a mathematical complexity associated with those rigorous models. The present model based on the multiple-bubble slug ejection model. It allows a finite number (N) of bubbles, separated by liquid slugs, in a channel. Boiling occurs at a user specified superheat, and a generated vapor is modeled to fill the whole cross section of the coolant channel except for a static liquid film left on the cladding or/and structure surfaces. The model also assumes a vapor with one uniform pressure. The present analysis is focused on the behavior of early sodium boiling after ULOHS.

  15. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment. Volume 3, Appendices C, D, E, F, and G

    Energy Technology Data Exchange (ETDEWEB)

    Harper, F.T.; Young, M.L.; Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States)] [and others

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the third of a three-volume document describing the project and contains descriptions of the probability assessment principles; the expert identification and selection process; the weighting methods used; the inverse modeling methods; case structures; and summaries of the consequence codes.

  16. Electromagnetic Gun Circuit Analysis Code (EGCAC)

    Science.gov (United States)

    Rolader, Glenn E.; Thornhill, Lindsey D.; Batteh, Jad H.; Scanlon, James J., III

    1993-01-01

    This paper describes a system engineering code that simulates the performance of a railgun/power supply system. The code, named EGCAC (Electromagnetic Gun Circuit Analysis Code), accounts for many performance degrading effects including viscous drag on the armature, viscous drag on the gas being pushed in front of the projectile, entrained gas that must be accelerated in front of the projectile, time-dependent rail resistance, armature resistance, system resistance, and ablation drag. EGCAC has been utilized to predict railgun performance up to a velocity of approximately 4 km/s for experiments at several laboratories. In this paper, the theory of EGCAC is described, and sample calculations are presented.

  17. Cause Analysis of Wuhan Tianheng Building Pile Accident

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The geological condition and the original structure feature and foundation design of Wuhan Tianheng building are described. The accident appearance of pile foundation in the construction execution of work is illustrated. The generating source of this pile foundation accident is analyzed in great details.``

  18. Analysis of Workplace Accidents in Automotive Repair Workshops in Spain

    Directory of Open Access Journals (Sweden)

    Antonio López-Arquillos

    2016-09-01

    Conclusion: Health and safety strategies and accident prevention measures should be individualized and adapted to the type of worker most likely to be injured in each type of accident. Occupational health and safety training courses designed according to worker profile, and improving the participation of the workers in small firms creating regional or roving safety representatives would improve working conditions.

  19. Accident Analysis and Barrier Function (AEB) Method. Manual for Incident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Svenson, Ola [Stockholm Univ. (Sweden). Dept. of Psychology

    2000-02-01

    The Accident Analysis and Barrier Function (AEB) Method models an accident or incident as a series of interactions between human and technical systems. In the sequence of human and technical errors leading to an accident there is, in principle, a possibility to arrest the development between each two successive errors. This can be done by a barrier function which, for example, can stop an operator from making an error. A barrier function can be performed by one or several barrier function systems. To illustrate, a mechanical system, a computer system or another operator can all perform a given barrier function to stop an operator from making an error. The barrier function analysis consists of analysis of suggested improvements, the effectiveness of the improvements, the costs of implementation, probability of implementation, the cost of maintaining the barrier function, the probability that maintenance will be kept up to standards and the generalizability of the suggested improvement. The AEB method is similar to the US method called HPES, but differs from that method in different ways. To exemplify, the AEB method has more emphasis on technical errors than HPES. In contrast to HPES that describes a series of events, the AEB method models only errors. This gives a more focused analysis making it well suited for checking other HPES-type accident analyses. However, the AEB method is a generic and stand-alone method that has been applied in other fields than nuclear power, such as, in traffic accident analyses.

  20. Injury patterns of seniors in traffic accidents: A technical and medical analysis

    Science.gov (United States)

    Brand, Stephan; Otte, Dietmar; Mueller, Christian Walter; Petri, Maximilian; Haas, Philipp; Stuebig, Timo; Krettek, Christian; Haasper, Carl

    2012-01-01

    AIM: To investigate the actual injury situation of seniors in traffic accidents and to evaluate the different injury patterns. METHODS: Injury data, environmental circumstances and crash circumstances of accidents were collected shortly after the accident event at the scene. With these data, a technical and medical analysis was performed, including Injury Severity Score, Abbreviated Injury Scale and Maximum Abbreviated Injury Scale. The method of data collection is named the German In-Depth Accident Study and can be seen as representative. RESULTS: A total of 4430 injured seniors in traffic accidents were evaluated. The incidence of sustaining severe injuries to extremities, head and maxillofacial region was significantly higher in the group of elderly people compared to a younger age (P accident-related injuries was higher in the group of seniors compared to other groups. CONCLUSION: Seniors are more likely to be involved in traffic injuries and to sustain serious to severe injuries compared to other groups. PMID:23173111

  1. Progress in accident analysis of the HYLIFE-II inertial fusion energy power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, S; Latkowski, J F; Gomez del Rio, J; Sanz, J

    2000-10-11

    The present work continues our effort to perform an integrated safety analysis for the HYLIFE-II inertial fusion energy (IFE) power plant design. Recently we developed a base case for a severe accident scenario in order to calculate accident doses for HYLIFE-II. It consisted of a total loss of coolant accident (LOCA) in which all the liquid flibe (Li{sub 2}BeF{sub 4}) was lost at the beginning of the accident. Results showed that the off-site dose was below the limit given by the DOE Fusion Safety Standards for public protection in case of accident, and that his dose was dominated by the tritium released during the accident.

  2. Road Accident Analysis: A Case Study of Federal Route FT024 Yong Peng- Parit Sulong

    Directory of Open Access Journals (Sweden)

    Mohd Masirin Mohd Idrus

    2016-01-01

    Full Text Available Traffic accidents are considered as an unplanned and unfortunate event which is a serious concern to the community as well as the authority. An accident-counter measure can reduce the rate of road accidents by initially identifying critical locations. The total of road accidents along FT 024 between 2009 and 2012 is 907 cases. Road accidents during the same period range between 24 % and 26 % each year. These accidents killed 34 people and injuring another 101 people. This research aims to identify factors that may contribute to the cause of accidents and to study the effects of speed, volume and road geometry on road accidents. In this study, the locations labelled as km 1, km 2, km 4, km 8, km 5 and km 14 of Federal Route FT 024 Yong Peng – Parit Sulong were selected as the study-case sections based on Accident Point Weightage of ranking, in which each location has different road characteristics. Speed study was carried out at selected road sections to evaluate the influence of speed upon road accidents; and traffic volume count was conducted at the same selected road sections to determine the existing condition of the route. Besides, road geometry observations and measurements were also conducted at selected sections, they were also studied to evaluate influence of road designs upon road accidents. The extracted data were analyzed by using regression analysis on different variables to evaluate the relationship between accident Weightage point and other dependent variables that were considered to have considerable effects upon road accidents such as mean speed, volume, shoulder width, lane width and access point. P value below 0.05 was considered as statistically significant. After conducting data analysis, this study showed that the number of road accidents increases with the increment of speed and access point. On the other hand, volume has no strong relationship to road accidents which means that it may not have an effect on accidents

  3. Development of MAAP5.0.3 Spent Fuel Pool Model for Severe Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Mi Ro [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    After the Fukushima accident, the severe accident phenomena in the Spent Fuel Pool (SFP) have been the great issues in the nuclear industry. Generally, during full power operation status, the decay heat of the spent fuel in the SFP is not high enough to cause the severe accident that is the say, the melting of fuel and fuel rack. In addition to this, the SFP of the PWR is not isolated within the containment like the SFP of the old BWR plant, there are so many possible measures to prevent and mitigate severe accidents in the SFP. On the other hand, in the low power shutdown status (fuel refueling), all the core is transferred into the SFP during the refueling period. At this period, if some accidents happen such as the loss of SFP cooling and the failure of SFP integrity then the accidents may be developed into severe accident because the decay heat is high enough. So, the analysis of severe accidents in the SFP during low power shutdown state is greatly affected to the establishment of the major strategies in the severe accident management guideline (SAMG). However, the status of the domestic technical background for those analyses is very weak. it is known that the decay heat of the spent fuel in the SFP is not high enough to cause the severe accident qualitatively. However, there are some possibilities that can cause the severe accidents in the SFP if the loss of SFP cooling and integrity happens simultaneously. The severe accident phenomena in SFP themselves are not much different from those in the containment. However, since the structure of SFP cannot be isolated during the accidents like the containment, the consequence can be extremely significant. So, in terms of the establishment of the severe accident management strategy, it is necessary that the quantitative analysis for the severe accident progression in the SFP should be performed. In this study, the general behavior which can be appeared during the severe accidents in the SFP was analyzed using the

  4. Development of the SCHAMBETA code for scoping analysis of HCDA

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Hahn, D. H

    2000-06-01

    A computer code, SCHAMBETA(Scoping Code for HCDA Analysis using Modified Bethe-Tait Method), is developed to investigate the core disassembly process following a meltdown accident in the framework of a mofified Bethe-Tait method as part of the scoping analysis work to demonstrate the inherent safety of conceptual designs of Korea Advanced Liquid Metal Reactor(KALIMER), A 150 Mwe pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. The methodologies adopted in the code ared particularly useful to perform various parametric studies for better understanding of core disassembly process of liquid metal fast reactors as well as to estimate upper-limit values of the energy release resulting from a power excursion. In the SCHAMBETA code, the core kinetics and hydraulic behavior of the KALIMER is followed over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion, starting at the time that the sodium-voided core reaches the melting temperature of the metallic fuels. For this purpose, the equations of state of pressure-energy density relationship are derived for the saturated-vapor as well as the solid liquid of metallic uranium fuel, and implemenmted into the formulations of the disassembly reactivity. Mathematical formulations are then developed, in the framework of Modified Bethe-Tait method, in a form relevant to utilize the improved equations of state as well as to consider Doppler effects, for scoping analysis of the super-prompt-critical power excursions driven by a specified rate of reactivity insertion.

  5. Development of Auditing Technology for Accident Analysis of SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, Y. J.; Jeong, J. J.; Kim, H. C.; Chung, Y. J.; Bae, K. H

    2006-02-15

    The objective of this project is to develop thermal hydraulic models of the regulatory auditing codes for the application of SMART-P integrated reactor. At initial period, PIRT has been performed to identify the model deficiencies and determine the priority of model improvements. The identified thermal hydraulic models has been implemented to RELAP5/MOD3.3 auditing code according to the PIRT ranking. The input model for SMART-P has been developed with consistent to the current design status documents and checked by independent reviewer as Q/A procedure.The evaluation of experimental availabilities and code collapsible has been done by expert group and summarized as validation matrix forms. The experimental data of VISTA, which is the only integral effect test facility, were used to validate the improved model. The safety analysis has been demonstrated for the essential accident scenario. The validation and demonstration show that the developed models are applicable to utilize in reliable and independent auditing for SMART design certification.

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

    Directory of Open Access Journals (Sweden)

    Cheewapattananuwong Weeradej

    2016-01-01

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

  7. A numerical study on a lumped-parameter model and a CFD code coupling for the analysis of the loss of coolant accident in a reactor containment; Etude numerique 0D-multiD pour l'analyse de perte de refrigerant dans une enceinte de confinement d'un reacteur nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Y.J.

    2005-12-15

    In the case of PWR severe accident (Loss of Coolant Accident, LOCA), the inner containment ambient properties such as temperature, pressure and gas species concentrations due to the released steam condensation are the main factors that determine the risk. For this reason, their distributions should be known accurately, but the complexity of the geometry and the computational costs are strong limitations to conduct full three-dimensional numerical simulations. An alternative approach is presented in this thesis, namely, the coupling between a lumped-parameter model and a CFD. The coupling is based on the introduction of a 'heat transfer function' between both models and it is expected that large decreases in the CPU-costs may be achieved. First of all, wall condensation models, such as the Uchida or the Chilton-Colburn models which are implemented in the code CAST3M/TONUS, are investigated. They are examined through steady-state calculations by using the code TONUS-0D, based on lumped parameter models. The temperature and the pressure within the inner containment are compared with those reported in the archival literature. In order to build the 'heat transfer function', natural convection heat transfer is then studied by using the code CAST3M for a partitioned cavity which represents a simplified geometry of the reactor containment. At a first step, two-dimensional natural convection heat transfer without condensation is investigated only. Either the incompressible-Boussinesq fluid flow model or the asymptotic low Mach model are considered for solving the time dependent conservation equations. The SUPG finite element method and the implicit scheme are applied for the numerical discretization. The computed results are qualified by the second-order Richardson extrapolation method which allows obtaining the so-called 'Exact values', i.e. grid size independent values. The computations are also validated through non-partitioned cavity case

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Tae Beom

    2004-02-15

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

  9. Report on preliminary analysis of state of nuclear criticality accident at JCO at Tokaimura, Ibaraki, Japan (I)

    Energy Technology Data Exchange (ETDEWEB)

    Ha, J.J.; Park, J.H.; Chang, J.H. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    This preliminary report was prepared by the Special Task Force Team of KAERI in order to analysis status of nuclear criticality accident broken out at 10:35 September 30, 1999 at JCO nuclear conversion test facility located at Tokaimura, Ibaraki, Japan. The report was consisted of accident summary of cause of accident summary of cause of accident and response by relevant organizations, and preliminary technical analysis of radiation exposure of JCO workers, analysis of cause of accident, and accident assessment and preventive actions against criticality accident. It is expected that JCO accident, Japan's first nuclear criticality accident, would make significant effects to Japan nuclear policy and would be also a good example to Korea future actions to be taken in use and development of nuclear energy. 63 refs., 3 figs., 1 tab. (Author)

  10. An Evidential Reasoning-Based CREAM to Human Reliability Analysis in Maritime Accident Process.

    Science.gov (United States)

    Wu, Bing; Yan, Xinping; Wang, Yang; Soares, C Guedes

    2017-01-09

    This article proposes a modified cognitive reliability and error analysis method (CREAM) for estimating the human error probability in the maritime accident process on the basis of an evidential reasoning approach. This modified CREAM is developed to precisely quantify the linguistic variables of the common performance conditions and to overcome the problem of ignoring the uncertainty caused by incomplete information in the existing CREAM models. Moreover, this article views maritime accident development from the sequential perspective, where a scenario- and barrier-based framework is proposed to describe the maritime accident process. This evidential reasoning-based CREAM approach together with the proposed accident development framework are applied to human reliability analysis of a ship capsizing accident. It will facilitate subjective human reliability analysis in different engineering systems where uncertainty exists in practice.

  11. Analysis of Construction Accidents in Turkey and Responsible Parties

    Science.gov (United States)

    GÜRCANLI, G. Emre; MÜNGEN, Uğur

    2013-01-01

    Construction is one of the world’s biggest industry that includes jobs as diverse as building, civil engineering, demolition, renovation, repair and maintenance. Construction workers are exposed to a wide variety of hazards. This study analyzes 1,117 expert witness reports which were submitted to criminal and labour courts. These reports are from all regions of the country and cover the period 1972–2008. Accidents were classified by the consequence of the incident, time and main causes of the accident, construction type, occupation of the victim, activity at time of the accident and party responsible for the accident. Falls (54.1%), struck by thrown/falling object (12.9%), structural collapses (9.9%) and electrocutions (7.5%) rank first four places. The accidents were most likely between the hours 15:00 and 17:00 (22.6%), 10:00–12:00 (18.7%) and just after the lunchtime (9.9%). Additionally, the most common accidents were further divided into sub-types. Expert-witness assessments were used to identify the parties at fault and what acts of negligence typically lead to accidents. Nearly two thirds of the faulty and negligent acts are carried out by the employers and employees are responsible for almost one third of all cases. PMID:24077446

  12. Control rod ejection accident analysis for a PWR with thorium fuel loading

    Energy Technology Data Exchange (ETDEWEB)

    Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2010-07-01

    This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kress, T. S. [comp.

    1985-04-01

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

  14. Superimposed Code Theorectic Analysis of DNA Codes and DNA Computing

    Science.gov (United States)

    2010-03-01

    Bounds for DNA Codes Based on Fibonacci Ensembles of DNA Sequences ”, 2008 IEEE Proceedings of International Symposium on Information Theory, pp. 2292...5, June 2008, pp. 525-34. 32 28. A. Macula, et al., “Random Coding Bounds for DNA Codes Based on Fibonacci Ensembles of DNA Sequences ”, 2008...combinatorial method of bio-memory design and detection that encodes item or process information as numerical sequences represented in DNA. ComDMem is a

  15. The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-01-01

    The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.

  16. The analysis of pressurizer safety valve stuck open accident for low power and shutdown PSA

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Ho Gon; Park, Jin Hee; Jang, Seong Chul; Kim, Tae Woon

    2005-01-01

    The PSV (Pressurizer Safety Valve) popping test carried out practically in the early phase of a refueling outage has a little possibility of triggering a test-induced LOCA due to a PSV not fully closed or stuck open. According to a KSNP (Korea Standard Nuclear Power Plant) low power and shutdown PSA (Probabilistic Safety Assessment), the failure of a HPSI (High Pressure Safety Injection) following a PSV stuck open was identified as a dominant accident sequence with a significant contribution to low power and shutdown risks. In this study, we aim to investigate the consequences of the NPP for the various accident sequences following the PSV stuck open as an initiating event through the thermal-hydraulic system code calculations. Also, we search the accident mitigation method for the sequence of HPSI failure, then, the applicability of the method is verified by the simulations using T/H system code.

  17. Analysis of accidents with organic material in health workers.

    Science.gov (United States)

    Vieira, Mariana; Padilha, Maria Itayra; Pinheiro, Regina Dal Castel

    2011-01-01

    This retrospective and descriptive study with a quantitative design aimed to evaluate occupational accidents with exposure to biological material, as well as the profile of workers, based on reporting forms sent to the Regional Reference Center of Occupational Health in Florianópolis/SC. Data collection was carried out through a survey of 118 reporting forms in 2007. Data were analyzed electronically. The occurrence of accidents was predominantly among nursing technicians, women and the mean age was 34.5 years. 73% of accidents involved percutaneous exposure, 78% had blood and fluid with blood, 44.91% resulted from invasive procedures. It was concluded that strategies to prevent the occurrence of accidents with biological material should include joint activities between workers and service management and should be directed at improving work conditions and organization.

  18. Spatial Analysis of Accident Spots Using Weighted Severity Index ...

    African Journals Online (AJOL)

    ADOWIE PERE

    Density-based Clustering for Traffic Accident Risk (DBCTAR) was carried out to assist in ascertaining the distribution of ... least one road vehicle, occurring on a road open to ... Road Safety Agency (FRSC), the Lagos State Traffic. Management ...

  19. 3D analysis of the reactivity insertion accident in VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Abdullayev, A. M.; Zhukov, A. I.; Slyeptsov, S. M. [NSC Kharkov Inst. for Physics and Technology, 1, Akademicheskaya Str., Kharkov 61108 (Ukraine)

    2012-07-01

    Fuel parameters such as peak enthalpy and temperature during rod ejection accident are calculated. The calculations are performed by 3D neutron kinetics code NESTLE and 3D thermal-hydraulic code VIPRE-W. Both hot zero power and hot full power cases were studied for an equilibrium cycle with Westinghouse hex fuel in VVER-1000. It is shown that the use of 3D methodology can significantly increase safety margins for current criteria and met future criteria. (authors)

  20. Accidents at work and costs analysis: a field study in a large Italian company.

    Science.gov (United States)

    Battaglia, Massimo; Frey, Marco; Passetti, Emilio

    2014-01-01

    Accidents at work are still a heavy burden in social and economic terms, and action to improve health and safety standards at work offers great potential gains not only to employers, but also to individuals and society as a whole. However, companies often are not interested to measure the costs of accidents even if cost information may facilitate preventive occupational health and safety management initiatives. The field study, carried out in a large Italian company, illustrates technical and organisational aspects associated with the implementation of an accident costs analysis tool. The results indicate that the implementation (and the use) of the tool requires a considerable commitment by the company, that accident costs analysis should serve to reinforce the importance of health and safety prevention and that the economic dimension of accidents is substantial. The study also suggests practical ways to facilitate the implementation and the moral acceptance of the accounting technology.

  1. Uncertainty analysis of accident notification time and emergency medical service response time in work zone traffic accidents.

    Science.gov (United States)

    Meng, Qiang; Weng, Jinxian

    2013-01-01

    Taking into account the uncertainty caused by exogenous factors, the accident notification time (ANT) and emergency medical service (EMS) response time were modeled as 2 random variables following the lognormal distribution. Their mean values and standard deviations were respectively formulated as the functions of environmental variables including crash time, road type, weekend, holiday, light condition, weather, and work zone type. Work zone traffic accident data from the Fatality Analysis Report System between 2002 and 2009 were utilized to determine the distributions of the ANT and the EMS arrival time in the United States. A mixed logistic regression model, taking into account the uncertainty associated with the ANT and the EMS response time, was developed to estimate the risk of death. The results showed that the uncertainty of the ANT was primarily influenced by crash time and road type, whereas the uncertainty of EMS response time is greatly affected by road type, weather, and light conditions. In addition, work zone accidents occurring during a holiday and in poor light conditions were found to be statistically associated with a longer mean ANT and longer EMS response time. The results also show that shortening the ANT was a more effective approach in reducing the risk of death than the EMS response time in work zones. To shorten the ANT and the EMS response time, work zone activities are suggested to be undertaken during non-holidays, during the daytime, and in good weather and light conditions.

  2. Traffic Accident Analysis Using Decision Trees and Neural Networks

    OpenAIRE

    Chong, Miao M.; Abraham, Ajith; Paprzycki, Marcin

    2004-01-01

    The costs of fatalities and injuries due to traffic accident have a great impact on society. This paper presents our research to model the severity of injury resulting from traffic accidents using artificial neural networks and decision trees. We have applied them to an actual data set obtained from the National Automotive Sampling System (NASS) General Estimates System (GES). Experiment results reveal that in all the cases the decision tree outperforms the neural network. Our research analys...

  3. Offsite Radiological Consequence Analysis for the Bounding Flammable Gas Accident

    CERN Document Server

    Carro, C A

    2003-01-01

    This document quantifies the offsite radiological consequences of the bounding flammable gas accident for comparison with the 25 rem Evaluation Guideline established in DOE-STD-3009, Appendix A. The bounding flammable gas accident is a detonation in a single-shell tank The calculation applies reasonably conservation input parameters in accordance with DOE-STD-3009, Appendix A, guidance. Revision 1 incorporates comments received from Office of River Protection.

  4. Bayesian optimization analysis of containment-venting operation in a boiling water reactor severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Xiaoyu; Ishikawa, Jun; Sugiyama, Tomoyuki; Maryyama, Yu [Nuclear Safety Research Center, Japan Atomic Energy Agency, Ibaraki (Japan)

    2017-03-15

    Containment venting is one of several essential measures to protect the integrity of the final barrier of a nuclear reactor during severe accidents, by which the uncontrollable release of fission products can be avoided. The authors seek to develop an optimization approach to venting operations, from a simulation-based perspective, using an integrated severe accident code, THALES2/KICHE. The effectiveness of the containment-venting strategies needs to be verified via numerical simulations based on various settings of the venting conditions. The number of iterations, however, needs to be controlled to avoid cumbersome computational burden of integrated codes. Bayesian optimization is an efficient global optimization approach. By using a Gaussian process regression, a surrogate model of the “black-box” code is constructed. It can be updated simultaneously whenever new simulation results are acquired. With predictions via the surrogate model, upcoming locations of the most probable optimum can be revealed. The sampling procedure is adaptive. Compared with the case of pure random searches, the number of code queries is largely reduced for the optimum finding. One typical severe accident scenario of a boiling water reactor is chosen as an example. The research demonstrates the applicability of the Bayesian optimization approach to the design and establishment of containment-venting strategies during severe accidents.

  5. Development of statistical analysis code for meteorological data (W-View)

    Energy Technology Data Exchange (ETDEWEB)

    Tachibana, Haruo; Sekita, Tsutomu; Yamaguchi, Takenori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    A computer code (W-View: Weather View) was developed to analyze the meteorological data statistically based on 'the guideline of meteorological statistics for the safety analysis of nuclear power reactor' (Nuclear Safety Commission on January 28, 1982; revised on March 29, 2001). The code gives statistical meteorological data to assess the public dose in case of normal operation and severe accident to get the license of nuclear reactor operation. This code was revised from the original code used in a large office computer code to enable a personal computer user to analyze the meteorological data simply and conveniently and to make the statistical data tables and figures of meteorology. (author)

  6. New Codes for Ambient Seismic Noise Analysis

    Science.gov (United States)

    Duret, F.; Mooney, W. D.; Detweiler, S.

    2007-12-01

    In order to determine a velocity model of the crust, scientists generally use earthquakes recorded by seismic stations. However earthquakes do not occur continuously and most are too weak to be useful. When no event is recorded, a waveform is generally considered to be noise. This noise, however, is not useless and carries a wealth of information. Thus, ambient seismic noise analysis is an inverse method of investigating the Earth's interior. Until recently, this technique was quite difficult to apply, as it requires significant computing capacities. In early 2007, however, a team led by Gregory Benson and Mike Ritzwoller from UC Boulder published a paper describing a new method for extracting group and phase velocities from those waveforms. The analysis consisting of recovering Green functions between a pair of stations, is composed of four steps: 1) single station data preparation, 2) cross-correlation and stacking, 3) quality control and data selection and 4) dispersion measurements. At the USGS, we developed a set of ready-to-use computing codes for analyzing waveforms to run the ambient noise analysis of Benson et al. (2007). Our main contribution to the analysis technique was to fully automate the process. The computation codes were written in Fortran 90 and the automation scripts were written in Perl. Furthermore, some operations were run with SAC. Our choices of programming language offer an opportunity to adapt our codes to the major platforms. The codes were developed under Linux but are meant to be adapted to Mac OS X and Windows platforms. The codes have been tested on Southern California data and our results compare nicely with those from the UC Boulder team. Next, we plan to apply our codes to Indonesian data, so that we might take advantage of newly upgraded seismic stations in that region.

  7. Risk analysis of emergent water pollution accidents based on a Bayesian Network.

    Science.gov (United States)

    Tang, Caihong; Yi, Yujun; Yang, Zhifeng; Sun, Jie

    2016-01-01

    To guarantee the security of water quality in water transfer channels, especially in open channels, analysis of potential emergent pollution sources in the water transfer process is critical. It is also indispensable for forewarnings and protection from emergent pollution accidents. Bridges above open channels with large amounts of truck traffic are the main locations where emergent accidents could occur. A Bayesian Network model, which consists of six root nodes and three middle layer nodes, was developed in this paper, and was employed to identify the possibility of potential pollution risk. Dianbei Bridge is reviewed as a typical bridge on an open channel of the Middle Route of the South to North Water Transfer Project where emergent traffic accidents could occur. Risk of water pollutions caused by leakage of pollutants into water is focused in this study. The risk for potential traffic accidents at the Dianbei Bridge implies a risk for water pollution in the canal. Based on survey data, statistical analysis, and domain specialist knowledge, a Bayesian Network model was established. The human factor of emergent accidents has been considered in this model. Additionally, this model has been employed to describe the probability of accidents and the risk level. The sensitive reasons for pollution accidents have been deduced. The case has also been simulated that sensitive factors are in a state of most likely to lead to accidents.

  8. Static Code Analysis with Gitlab-CI

    CERN Document Server

    Datko, Szymon Tomasz

    2016-01-01

    Static Code Analysis is a simple but efficient way to ensure that application’s source code is free from known flaws and security vulnerabilities. Although such analysis tools are often coming with more advanced code editors, there are a lot of people who prefer less complicated environments. The easiest solution would involve education – where to get and how to use the aforementioned tools. However, counting on the manual usage of such tools still does not guarantee their actual usage. On the other hand, reducing the required effort, according to the idea “setup once, use anytime without sweat” seems like a more promising approach. In this paper, the approach to automate code scanning, within the existing CERN’s Gitlab installation, is described. For realization of that project, the Gitlab-CI service (the “CI” stands for "Continuous Integration"), with Docker assistance, was employed to provide a variety of static code analysers for different programming languages. This document covers the gene...

  9. Analysis of the 1957-58 Soviet nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Trabalka, J.R.; Eyman, L.D.; Auerbach, S.I.

    1979-12-01

    The occurrence of a Soviet accident in the winter of 1957-58, involving the atmospheric release of reprocessed fission wastes (cooling time approximately 1-2 yrs.), appears to have been confirmed, primarily by an analysis of the USSR radioecology literature. Due to the high population density in the affected region (Cheliabinsk Province in the highly industrialized Urals Region) and the reported level of /sup 90/Sr contamination , the event probably resulted in the evacuation and/or resettlement of the human population from a significant area (100-1000 km/sup 2/). The resulting contamination zone is estimated to have contained approximately 10/sup 6/ Ci of /sup 90/Sr (reference radionuclide); a relatively small fraction of the total may have been dispersed as an aerosol. Although a plausible explanation for the incident exists (i.e., use of now-obsolete waste storage-/sup 137/Cs isotope separation techniques), it is not yet possible, based on the limited information presently available, to completely dismiss this phenomenon as a purely historical event. It seems imperative that we have a complete explanation of the causes and consequences of this incident. Soviet experience gained in application of corrective measures would be invaluable to the rest of the world nuclear community.

  10. Multi-physics modelling in the frame of the DRACCAR code development and its application to spent-fuel pool draining accidents

    Energy Technology Data Exchange (ETDEWEB)

    Jacq, F.; Luze, O. de; Guillard, G.; Bascou, S. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Fontenay-aux-Roses (France)

    2013-07-01

    To meet the simulation needs of its LOCA R and D program, the IRSN is developing a multi-pin computational tool named DRACCAR. In order to realistically describe the behavior of the reactor core during a Loss Of Coolant Accident (LOCA), modeling has to take into account many coupled phenomena such as thermics (heat generation, radiation, convection and conduction), hydraulics (multi dimensional 1-3 phase flow, shrinkage), mechanics (thermal dilatation, creep, embrittlement) and chemistry (oxidation, oxygen diffusion, hydriding,..). This paper presents several aspects of the DRACCAR code abilities: investigation of the bundle rods strain during a LOCA transient, checking of the thermalhydraulics during reflooding of a partially ballooned bundle, and application to spent-fuel-pool draining accidents in the case of a propagation of the burn front in a typical non axis-symmetrical situation for the thermal heat exchanges which are driving the accident. (orig.)

  11. Analysis of flammability in the attached buildings to containment under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, J.C. de la, E-mail: juan-carlos.de-la-rosa-blul@ec.europa.eu [European Commission Joint Research Centre (Netherlands); Fornós, Joan, E-mail: jfornosh@anacnv.com [Asociación Nuclear Ascó-Vandellós (Spain)

    2016-11-15

    Highlights: • Analysis of flammability conditions in buildings outside containment. • Stepwise approach easily applicable for any kind of containment and attached buildings layout. • Detailed application for real plant conditions has been included. - Abstract: Right after the events unfolded in Fukushima Daiichi, the European Union countries agreed in subjecting Nuclear Power Plants to Stress Tests as developed by WENRA and ENSREG organizations. One of the results as implemented in many European countries derived from such tests consisted of mandatory technical instructions issued by nuclear regulatory bodies on the analysis of potential risk of flammable gases in attached buildings to containment. The current study addresses the key aspects of the analysis of flammable gases leaking to auxiliary buildings attached to Westinghouse large-dry PWR containment for the specific situation where mitigating systems to prevent flammable gases to grow up inside containment are available, and containment integrity is preserved – hence avoiding isolation system failure. It also provides a full practical exercise where lessons learned derived from the current study – hence limited to the imposed boundary conditions – are applied. The leakage of gas from the containment to the support buildings is based on separate calculations using the EPRI-owned Modular Accident Analysis Program, MAAP4.07. The FATE™ code (facility Flow, Aerosol, Thermal, and Explosion) was used to model the transport and distribution of leaked flammable gas (H{sub 2} and CO) in the penetration buildings. FATE models the significant mixing (dilution) which occurs as the released buoyant gas rises and entrains air. Also, FATE accounts for the condensation of steam on room surfaces, an effect which acts to concentrate flammable gas. The results of the analysis show that during a severe accident, flammable conditions are unlikely to occur in compartmentalized buildings such as the one used in the

  12. Analysis of Loss-of-Coolant Accidents in the NBSR

    Energy Technology Data Exchange (ETDEWEB)

    Baek J. S.; Cheng L.; Diamond, D.

    2014-05-23

    This report documents calculations of the fuel cladding temperature during loss-of-coolant accidents in the NBSR. The probability of a pipe failure is small and procedures exist to minimize the loss of water and assure emergency cooling water flows into the reactor core during such an event. Analysis in the past has shown that the emergency cooling water would provide adequate cooling if the water filled the flow channels within the fuel elements. The present analysis is to determine if there is adequate cooling if the water drains from the flow channels. Based on photographs of how the emergency water flows into the fuel elements from the distribution pan, it can be assumed that this water does not distribute uniformly across the flow channels but rather results in a liquid film flowing downward on the inside of one of the side plates in each fuel element and only wets the edges of the fuel plates. An analysis of guillotine breaks shows the cladding temperature remains below the blister temperature in fuel plates in the upper section of the fuel element. In the lower section, the fuel plates are also cooled by water outside the element that is present due to the hold-up pan and temperatures are lower than in the upper section. For small breaks, the simulation results show that the fuel elements are always cooled on the outside even in the upper section and the cladding temperature cannot be higher than the blister temperature. The above results are predicated on assumptions that are examined in the study to see their influence on fuel temperature.

  13. Calculation of Departure from Nucleate Boiling Ratio (DNBR) minimum for accident analysis of main steam line break at Angra-1; Calculo do minimo DNBR para analise do acidente de ruptura da linha principal de vapor em Angra-1

    Energy Technology Data Exchange (ETDEWEB)

    Machado, Marcio Dornellas [ELETROBRAS Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil). E-mail: mdorne@eletronuclear.gov.br

    2000-07-01

    The maintenance costs, the operational problems and the failures possibilities of the boron injection system, composed by pumps, valves, heated lines and the boron injection tank, make this tank removal or the boron concentration reduction advisable for Angra 1 Power Plant. The main accident from chapter XV of the final safety analysis report affected by this modification is the main steam line break. It is necessary the interaction of the areas of Accidents and Transients Analysis (RETRAN 02/Mod 5.1 code), Neutronics (APA System) and Thermohydraulics (COBRA IIIC/MIT) to analyse this accident. The present Angra 1 boron concentration is 20000 ppm and it could be reduced to 2000 ppm as a result of the present study. The Departure from Nucleate Boiling Ratio (DNBR) is the restrictive parameter of this accident, which is calculated from the initials and boundary conditions obtained from the Transients and Accidents Analysis and Neutronics areas. (author)

  14. Energy Analysis of Road Accidents Based on Close-Range Photogrammetry

    Directory of Open Access Journals (Sweden)

    Alejandro Morales

    2015-11-01

    Full Text Available This paper presents an efficient and low-cost approach for energy analysis of road accidents using images obtained using consumer-grade digital cameras and smartphones. The developed method could be used by security forces in order to improve the qualitative and quantitative analysis of traffic accidents. This role of the security forces is crucial to settle arguments; consequently, the remote and non-invasive collection of accident related data before the scene is modified proves to be essential. These data, taken in situ, are the basis to perform the necessary calculations, basically the energy analysis of the road accident, for the corresponding expert reports and the reconstruction of the accident itself, especially in those accidents with important damages and consequences. Therefore, the method presented in this paper provides the security forces with an accurate, three-dimensional, and scaled reconstruction of a road accident, so that it may be considered as a support tool for the energy analysis. This method has been validated and tested with a real crash scene simulated by the local police in the Academy of Public Safety of Extremadura, Spain.

  15. Exploring the potential of data mining techniques for the analysis of accident patterns

    DEFF Research Database (Denmark)

    Prato, Carlo Giacomo; Bekhor, Shlomo; Galtzur, Ayelet

    2010-01-01

    Research in road safety faces major challenges: individuation of the most significant determinants of traffic accidents, recognition of the most recurrent accident patterns, and allocation of resources necessary to address the most relevant issues. This paper intends to comprehend which data mining...... and association rules) data mining techniques are implemented for the analysis of traffic accidents occurred in Israel between 2001 and 2004. Results show that descriptive techniques are useful to classify the large amount of analyzed accidents, even though introduce problems with respect to the clear......-cut definition of the clusters and the triviality of the description of the main accident characteristics. Results also show that prediction techniques present problems with respect to the large number of rules produced by decision trees, the interpretation of neural network results in terms of relative...

  16. Transient Ejector Analysis (TEA) code user's guide

    Science.gov (United States)

    Drummond, Colin K.

    1993-01-01

    A FORTRAN computer program for the semi analytic prediction of unsteady thrust augmenting ejector performance has been developed, based on a theoretical analysis for ejectors. That analysis blends classic self-similar turbulent jet descriptions with control-volume mixing region elements. Division of the ejector into an inlet, diffuser, and mixing region allowed flexibility in the modeling of the physics for each region. In particular, the inlet and diffuser analyses are simplified by a quasi-steady-analysis, justified by the assumption that pressure is the forcing function in those regions. Only the mixing region is assumed to be dominated by viscous effects. The present work provides an overview of the code structure, a description of the required input and output data file formats, and the results for a test case. Since there are limitations to the code for applications outside the bounds of the test case, the user should consider TEA as a research code (not as a production code), designed specifically as an implementation of the proposed ejector theory. Program error flags are discussed, and some diagnostic routines are presented.

  17. Sandia National Laboratories analysis code data base

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, C.W.

    1994-11-01

    Sandia National Laboratories, mission is to solve important problems in the areas of national defense, energy security, environmental integrity, and industrial technology. The Laboratories` strategy for accomplishing this mission is to conduct research to provide an understanding of the important physical phenomena underlying any problem, and then to construct validated computational models of the phenomena which can be used as tools to solve the problem. In the course of implementing this strategy, Sandia`s technical staff has produced a wide variety of numerical problem-solving tools which they use regularly in the design, analysis, performance prediction, and optimization of Sandia components, systems and manufacturing processes. This report provides the relevant technical and accessibility data on the numerical codes used at Sandia, including information on the technical competency or capability area that each code addresses, code ``ownership`` and release status, and references describing the physical models and numerical implementation.

  18. Development of Tritium Permeation Analysis Code (TPAC)

    Energy Technology Data Exchange (ETDEWEB)

    Eung S. Kim; Chang H. Oh; Mike Patterson

    2010-10-01

    Idaho National Laboratory developed the Tritium Permeation Analysis Code (TPAC) for tritium permeation in the Very High Temperature Gas Cooled Reactor (VHTR). All the component models in the VHTR were developed and were embedded into the MATHLAB SIMULINK package with a Graphic User Interface. The governing equations of the nuclear ternary reaction and thermal neutron capture reactions from impurities in helium and graphite core, reflector, and control rods were implemented. The TPAC code was verified using analytical solutions for the tritium birth rate from the ternary fission, the birth rate from 3He, and the birth rate from 10B. This paper also provides comparisons of the TPAC with the existing other codes. A VHTR reference design was selected for tritium permeation study from the reference design to the nuclear-assisted hydrogen production plant and some sensitivity study results are presented based on the HTGR outlet temperature of 750 degrees C.

  19. A Hybrid Algorithm of Traffic Accident Data Mining on Cause Analysis

    Directory of Open Access Journals (Sweden)

    Jianfeng Xi

    2013-01-01

    Full Text Available Road traffic accident databases provide the basis for road traffic accident analysis, the data inside which usually has a radial, multidimensional, and multilayered structure. Traditional data mining algorithms such as association rules, when applied alone, often yield uncertain and unreliable results. An improved association rule algorithm based on Particle Swarm Optimization (PSO put forward by this paper can be used to analyze the correlation between accident attributes and causes. The new algorithm focuses on characteristics of the hyperstereo structure of road traffic accident data, and the association rules of accident causes can be calculated more accurately and in higher rates. A new concept of Association Entropy is also defined to help compare the importance between different accident attributes. T-test model and Delphi method were deployed to test and verify the accuracy of the improved algorithm, the result of which was a ten times faster speed for random traffic accident data sampling analyses on average. In the paper, the algorithms were tested on a sample database of more than twenty thousand items, each with 56 accident attributes. And the final result proves that the improved algorithm was accurate and stable.

  20. [Paragliding accidents--a prospective analysis in Swiss mountain regions].

    Science.gov (United States)

    Lautenschlager, S; Karli, U; Matter, P

    1993-01-01

    During the period from 1.1 to 31.12.90, 86 injuries associated with paragliding were analysed in a prospective study in 12 different Swiss hospitals with reference to causes, patterns, and frequencies. Spine injuries (36%) and lesions of the lower extremities (35%) were diagnosed most frequently. Surprisingly no neurological complications occurred, which is possibly explained by the solitary axial trauma. In 15 cases very severe malleolar fractures required surgical intervention. One accident was fatal due to a lung rupture. 60% of all accidents happened during the landing phase, 26% at launching and 14% at flight. Half of the pilots were affected in their primary training course. Most accidents were due to an in-flight error of judgement, such as incorrect estimation of wind conditions and a choice of unfavourable landing sites. In contrast to early reports of hang-gliding injuries, only one accident was due to an equipment failure, namely a ruptured steering line. In more than a third of all accidents, the used paraglider was not in correct correlation with the pilot's weight and experience. Inspired by the desire for a long flight, gliders of too large surface-areas were often used, leading to a more unstable flight. To reduce the frequency of paragliding injuries, an accurate choice of equipment and increased attention to environmental factors is mandatory. Furthermore education-programs should focus more on intensifying the pilot's mental and practical skills.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-04-15

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

  2. Development of analytical code for core disruptive accident of metallic fuel core and evaluation of upper limitation of coolant void reactivity

    Energy Technology Data Exchange (ETDEWEB)

    Ueda, Nobuyuki [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab

    2001-06-01

    In the event of HCDA (Hypothetical Core Damage Accident) the safety of the metallic core must be ensured. Coolant void reactivity, which produces a positive reactivity effect, must be restrained to exclude 'energetic sequence' (uncontrollable power generation). As this reactivity and the core performance are in a state of confrontation, an upper limit must be determined through safety analyses. The larger the coolant void reactivity becomes, the more the core performance improves. A core dynamics analysis code named CANIS has been developed to analysis metallic fuel properties, irradiation behavior, and transient behavior in the initiating phase of HCDA. ULOF (Unprotected Loss Of Flow) events, which are considered major HCDA initiators, are analyzed with various parameters of cladding failure conditions and molten fuel dispersion velocities. The results indicate, the cladding failure condition is the most sensitive safety parameter. The coolant void reactivity of 12$ is acceptable in the condition that cladding fails at 1,000 degC. And the upper value is limited to 8$ for the conservative condition of a 1,200 degC failure criteria. (author)

  3. Accident Precursor Analysis and Management: Reducing Technological Risk Through Diligence

    Science.gov (United States)

    Phimister, James R. (Editor); Bier, Vicki M. (Editor); Kunreuther, Howard C. (Editor)

    2004-01-01

    Almost every year there is at least one technological disaster that highlights the challenge of managing technological risk. On February 1, 2003, the space shuttle Columbia and her crew were lost during reentry into the atmosphere. In the summer of 2003, there was a blackout that left millions of people in the northeast United States without electricity. Forensic analyses, congressional hearings, investigations by scientific boards and panels, and journalistic and academic research have yielded a wealth of information about the events that led up to each disaster, and questions have arisen. Why were the events that led to the accident not recognized as harbingers? Why were risk-reducing steps not taken? This line of questioning is based on the assumption that signals before an accident can and should be recognized. To examine the validity of this assumption, the National Academy of Engineering (NAE) undertook the Accident Precursors Project in February 2003. The project was overseen by a committee of experts from the safety and risk-sciences communities. Rather than examining a single accident or incident, the committee decided to investigate how different organizations anticipate and assess the likelihood of accidents from accident precursors. The project culminated in a workshop held in Washington, D.C., in July 2003. This report includes the papers presented at the workshop, as well as findings and recommendations based on the workshop results and committee discussions. The papers describe precursor strategies in aviation, the chemical industry, health care, nuclear power and security operations. In addition to current practices, they also address some areas for future research.

  4. Sensitivity analysis for CORSOR models simulating fission product release in LOFT-LP-FP-2 severe accident experiment

    Energy Technology Data Exchange (ETDEWEB)

    Hoseyni, Seyed Mohsen [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Basic Sciences; Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Young Researchers and Elite Club; Pourgol-Mohammad, Mohammad [Sahand Univ. of Technology, Tabriz (Iran, Islamic Republic of). Dept. of Mechanical Engineering; Yousefpour, Faramarz [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of)

    2017-03-15

    This paper deals with simulation, sensitivity and uncertainty analysis of LP-FP-2 experiment of LOFT test facility. The test facility simulates the major components and system response of a pressurized water reactor during a LOCA. MELCOR code is used for predicting the fission product release from the core fuel elements in LOFT LP-FP-2 experiment. Moreover, sensitivity and uncertainty analysis is performed for different CORSOR models simulating release of fission products in severe accident calculations for nuclear power plants. The calculated values for the fission product release are compared under different modeling options to the experimental data available from the experiment. In conclusion, the performance of 8 CORSOR modeling options is assessed for available modeling alternatives in the code structure.

  5. Analysis of traffic accidents on rural highways using Latent Class Clustering and Bayesian Networks.

    Science.gov (United States)

    de Oña, Juan; López, Griselda; Mujalli, Randa; Calvo, Francisco J

    2013-03-01

    One of the principal objectives of traffic accident analyses is to identify key factors that affect the severity of an accident. However, with the presence of heterogeneity in the raw data used, the analysis of traffic accidents becomes difficult. In this paper, Latent Class Cluster (LCC) is used as a preliminary tool for segmentation of 3229 accidents on rural highways in Granada (Spain) between 2005 and 2008. Next, Bayesian Networks (BNs) are used to identify the main factors involved in accident severity for both, the entire database (EDB) and the clusters previously obtained by LCC. The results of these cluster-based analyses are compared with the results of a full-data analysis. The results show that the combined use of both techniques is very interesting as it reveals further information that would not have been obtained without prior segmentation of the data. BN inference is used to obtain the variables that best identify accidents with killed or seriously injured. Accident type and sight distance have been identify in all the cases analysed; other variables such as time, occupant involved or age are identified in EDB and only in one cluster; whereas variables vehicles involved, number of injuries, atmospheric factors, pavement markings and pavement width are identified only in one cluster. Copyright © 2012 Elsevier Ltd. All rights reserved.

  6. Explorative spatial analysis of traffic accident statistics and road mortality among the provinces of Turkey.

    Science.gov (United States)

    Erdogan, Saffet

    2009-10-01

    The aim of the study is to describe the inter-province differences in traffic accidents and mortality on roads of Turkey. Two different risk indicators were used to evaluate the road safety performance of the provinces in Turkey. These indicators are the ratios between the number of persons killed in road traffic accidents (1) and the number of accidents (2) (nominators) and their exposure to traffic risk (denominator). Population and the number of registered motor vehicles in the provinces were used as denominators individually. Spatial analyses were performed to the mean annual rate of deaths and to the number of fatal accidents that were calculated for the period of 2001-2006. Empirical Bayes smoothing was used to remove background noise from the raw death and accident rates because of the sparsely populated provinces and small number of accident and death rates of provinces. Global and local spatial autocorrelation analyses were performed to show whether the provinces with high rates of deaths-accidents show clustering or are located closer by chance. The spatial distribution of provinces with high rates of deaths and accidents was nonrandom and detected as clustered with significance of Paccidents and deaths were located in the provinces that contain the roads connecting the Istanbul, Ankara, and Antalya provinces. Accident and death rates were also modeled with some independent variables such as number of motor vehicles, length of roads, and so forth using geographically weighted regression analysis with forward step-wise elimination. The level of statistical significance was taken as Paccidents according to denominators in the provinces. The geographically weighted regression analyses did significantly better predictions for both accident rates and death rates than did ordinary least regressions, as indicated by adjusted R(2) values. Geographically weighted regression provided values of 0.89-0.99 adjusted R(2) for death and accident rates, compared with 0

  7. Analysis of the optimality of the standard genetic code.

    Science.gov (United States)

    Kumar, Balaji; Saini, Supreet

    2016-07-19

    Many theories have been proposed attempting to explain the origin of the genetic code. While strong reasons remain to believe that the genetic code evolved as a frozen accident, at least for the first few amino acids, other theories remain viable. In this work, we test the optimality of the standard genetic code against approximately 17 million genetic codes, and locate 29 which outperform the standard genetic code at the following three criteria: (a) robustness to point mutation; (b) robustness to frameshift mutation; and (c) ability to encode additional information in the coding region. We use a genetic algorithm to generate and score codes from different parts of the associated landscape, which are, as a result, presumably more representative of the entire landscape. Our results show that while the genetic code is sub-optimal for robustness to frameshift mutation and the ability to encode additional information in the coding region, it is very strongly selected for robustness to point mutation. This coupled with the observation that the different performance indicator scores for a particular genetic code are negatively correlated makes the standard genetic code nearly optimal for the three criteria tested in this work.

  8. Analysis of Workplace Accidents in Automotive Repair Workshops in Spain.

    Science.gov (United States)

    López-Arquillos, Antonio; Rubio-Romero, Juan Carlos

    2016-09-01

    To analyze the effects of the factors associated with different types of injury (superficial wounds, dislocations and sprains, bone fractures, concussion and internal injuries, burns scalding and freezing) caused by occupational accidents in automotive repair workshops. Study of a sample consisting of 89,954 industry accidents reported from 2003 to 2008. Odds ratios were calculated with a 95% confidence interval. Belonging to a small company is a risk factor for suffering three of the five types of injury studied. Women are less likely to suffer burns and superficial wounds, and more likely to suffer dislocations or sprains. Foreign workers are more likely to suffer concussion and internal injuries. Health and safety strategies and accident prevention measures should be individualized and adapted to the type of worker most likely to be injured in each type of accident. Occupational health and safety training courses designed according to worker profile, and improving the participation of the workers in small firms creating regional or roving safety representatives would improve working conditions.

  9. Analysis of Criticality Accident Transients of Uranium Solution System

    Institute of Scientific and Technical Information of China (English)

    DUAN; Ming-hui; DU; Kai-wen; LIU; Zhen-hua

    2012-01-01

    <正>In the nuclear fuel cycle, fissile materials are often dissolved in water. Criticality accidents are likely to happen in the uranium solution system and release a large amount of energy and radioactive materials. Therefore, the criticality safety of uranium solution system is very important in the nuclear safety technology research.

  10. Fractal methods in image analysis and coding

    OpenAIRE

    Neary, David

    2001-01-01

    In this thesis we present an overview of image processing techniques which use fractal methods in some way. We show how these fields relate to each other, and examine various aspects of fractal methods in each area. The three principal fields of image processing and analysis th a t we examine are texture classification, image segmentation and image coding. In the area of texture classification, we examine fractal dimension estimators, comparing these methods to other methods in use, a...

  11. Analysis of Shahid Rajaee hospital administrative data on injuries resulting from car accidents in Shiraz, Iran: 2011-2014 data.

    Science.gov (United States)

    Yadollahi, Mahnaz; Ghiassee, Aida; Anvar, Mehrdad; Ghaem, Hale; Farahmand, Mohammad

    2017-02-01

    The administrative data from trauma centers could serve as potential sources of invaluable information while studying epidemiologic features of car accidents. In this cross-sectional analysis of Shahid Rajaee hospital administrative data, we aimed to evaluate patients injured in car accidents in terms of age, gender, injury severity, injured body regions and hospitalization outcome in the recent four years (2011-2014). The hospital registry was accessed at Shiraz Trauma Research Center (Shiraz, Iran) and the admission's unit data were merged with the information gathered upon discharge. A total number of 27,222 car accident patients aged over 15 years with International Classification of Diseases 10th revision (ICD-10) external causes of injury codes (V40.9-V49.9) were analyzed. Injury severity score and injured body regions were determined based on converting ICD-10 injury codes to Abbreviated Injury Scale (AIS-98) severity codes using a domestically developed electronic algorithm. A binary logistic regression model was applied to the data to examine the contribution of all independent variables to in-hospital mortality. Men accounted for 68.9% of the injuries and the male to female ratio was 2.2:1. The age of the studied population was (34 ± 15) years, with more than 77.2% of the population located in the 15-45 years old age group. Head and neck was the most commonly injured body region (39.0%) followed by extremities (27.2%). Injury severity score (ISS) was calculated for 13,152 (48.3%) patients, of whom, 80.9% had severity scores less than 9. There were 332 patients (1.2%) admitted to the intensive care units and 422 in-hospital fatalities (1.5%) were recorded during the study period. Age above 65 years [OR = 7.4, 95% CI (5.0-10.9)], ISS above 16 [OR = 9.1, 95% CI (5.5-14.9)], sustaining a thoracic injury [OR = 7.4, 95% CI (4.6-11.9)] and head injury [OR = 4.9, 95% CI (3.1-7.6)] were the most important independent predictors of death following car

  12. Analysis of Traffic Accident-Related Facial Trauma.

    Science.gov (United States)

    Choi, Su Hyun; Gu, Ja Hea; Kang, Dong Hee

    2016-10-01

    The consequences of facial trauma remain of great significance both functionally and esthetically. Traffic accident-related facial trauma is a frequent and significant cause of maxillofacial injury. The purpose of this study was to determine the natural history of traffic accident-related facial injuries in 846 patients over a 10-year period at a regional emergency center. The authors report a retrospective study of 846 patients with facial trauma from traffic accidents. The medical records of these patients were reviewed and analyzed to determine clinical characteristics, treatments, and outcomes of traffic accident-related facial trauma. In total, 687 of the 846 patients (81.2%) had fractures of the face, and nasal bone fractures were the most common followed by zygomatic complex fractures, blow-out fractures, and maxilla fractures. About 51.2% patients had open wounds on the face, such as lacerations, abrasions, skin or soft tissue defects, and friction burns. Only 7.4% of patients were treated conservatively and the others underwent repair or closed and open reduction. The complication rate was 46.3%, and scars were the most common followed by nose-related complication, hypoesthesia, and eyelid deformities. In addition, 47.6% of complication patients underwent secondary operations. Almost 15% of drivers were drunk, and about 8.7% were confessed drowsy during drive. Almost 30% of pedestrians were hit when they jaywalked across the street. Epidemiological data regarding traffic accident-related facial trauma are important and useful not only for decisions about patient care and developing optimal treatment regimens but to develop new methods to prevent injuries.

  13. Test and validation of CFD codes for the simulation of accident-typical phenomena in the reactor containment; Erprobung und Validierung von CFD-Codes fuer die Simulation von unfalltypischen Phaenomenen im Sicherheitseinschluss

    Energy Technology Data Exchange (ETDEWEB)

    Schramm, Berthold; Stewering, Joern; Sonnenkalb, Martin

    2014-03-15

    CFD (Computational Fluid Dynamic) simulation techniques have a growing relevance for the simulation and assessment of accidents in nuclear reactor containments. Some fluid dynamic problems like the calculation of the flow resistances in a complex geometry, turbulence calculations or the calculation of deflagrations could only be solved exactly for very simple cases. These fluid dynamic problems could not be represented by lumped parameter models and must be approximated numerically. Therefore CFD techniques are discussed by a growing international community in conferences like the CFD4NRS-conference. Also the number of articles with a CFD topic is increasing in professional journals like Nuclear Engineering and Design. CFD tools like GASFLOW or GOTHIC are already in use in European nuclear site licensing processes for future nuclear power plants like EPR or AP1000 and the results of these CFD tools are accepted by the authorities. For these reasons it seems to be necessary to build up national competences in the field of CFD techniques and it is important to validate and assess the existing CFD tools. GRS continues the work for the validation and assessment of CFD codes for the simulation of accident scenarios in a nuclear reactor containment within the framework of the BMWi sponsored project RS1500. The focus of this report is on the following topics: - Further validation of condensation models from GRS, FZJ and ANSYS and development of a new condensate model. - Validation of a new turbulence model which was developed by the University of Stuttgart in cooperation with ANSYS. - The formation and dissolution of light gas stratifications are analyzed by large scale experiments. These experiments were simulated by GRS. - The AREVA correlations for hydrogen recombiners (PARs) could be improved by GRS after the analysis of experimental data. Relevant experiments were simulated with this improved recombiner correlation. - Analyses on the simulation of H{sub 2

  14. Validation of a plant dynamics code for 4S - Test analysis of natural circulation behavior

    Energy Technology Data Exchange (ETDEWEB)

    Sebe, F.; Horie, H.; Matsumiya, H. [Toshiba Corporation, 8 Shinsugita-Cho, Isogo-Ku, Yokohama, 235-8523 (Japan); Fanning, T. H. [Argonne National Laboratory, 9700 S Cass Ave, Argonne, IL 60439 (United States)

    2012-07-01

    A plant transient dynamics code for a sodium-cooled fast reactor was developed by Toshiba. The code is used to evaluate the safety performance of Super-Safe, Small, and Simple reactor (4S) for Anticipated Operational Occurrences (AOOs), Design Basis Accident (DBA) and Beyond DBA (BDBA). The code is currently undergoing verification and validation (V and V). As one of the validation, test analysis of the Shutdown Heat Removal Test (SHRT)-17 performed in the Experimental Breeder Reactor (EBR)-II was conducted. The SHRT-17 is protected loss of flow test. The purpose of this validation is to confirm capability of the code to simulate natural circulation behavior of the plant. As a result, good agreements are shown between the analytical results and the measured data which were available from instrumented subassembly. The detailed validation result of the natural circulation behavior is described in this paper. (authors)

  15. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    Brown, J. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)] [and others

    1997-06-01

    This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  16. Analysis of Fukushima unit 2 accident considering the operating conditions of RCIC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il, E-mail: sikim@kaeri.re.kr; Park, Jong Hwa; Ha, Kwang Soon; Cho, Song-Won; Song, JinHo

    2016-03-15

    Highlights: • Fukushima unit 2 accident was analyzed using MELCOR 1.8.6. • RCIC operating conditions were assumed and best case was selected. • Effect of RCIC operating condition on accident scenario was found. - Abstract: A severe accident in Fukushima occurred on March 11, 2011 and units 1, 2 and 3 were damaged severely. A tsunami following an earthquake made the supply of electricity power stop, and the safety systems, which use AC or DC power in plants could not operate properly. It is supposed that the degree of core degradation of unit 2 is less serious than in the other plants, and it was estimated that the operation of reactor core isolation cooling (RCIC) system at the initial stage of the accident minimized the core damage through decay heat removal. Although the operating conditions of the RCIC system are not known clearly, it can be important to analyze the accident scenario of unit 2. In this study, best case of the Fukushima unit 2 accident was presented considering the operating conditions of the RCIC system. The effects of operating condition on core degradation and fission product release rate to environment were also examined. In addition, importance of torus room flooding level in the accident analysis was discussed. MELCOR 1.8.6 was used in this research, and the geometries of plant and operating conditions of safety system were obtained from TEPCO through OECD/NEA BSAF Project.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-11-01

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

  18. The sensitivity analysis for APR1400 nodalization under Large Break LOCA condition based on mars code

    Directory of Open Access Journals (Sweden)

    Jang Hyung-Wook

    2017-01-01

    Full Text Available The phenomena of loss of coolant accident have been investigated for long time and the result of experiment shows that the flow condition in the downcomer during the end-of-blowdown were highly multi-dimensional at full-scale. However, the downcomer nodalization of input deck for large break loss of coolant accident used in advanced power reactor 1400 analyses are made up with 1-D model and improperly designed to describe realistic coolant phenomena during loss of coolant accident analysis. In this paper, the authors modified the nodalization of MARS code LBLOCA input deck and performed LBLOCA analysis with new input deck. From original LBLOCA input deck file, the nodalization of downcomer and junction connections with 4 cold legs and direct vessel injection lines are modified for reflecting the realistic cross-flow effect and real downcomer structure. The analysis results show that the peak cladding temperature of new input deck decreases more rapidly than previous result and that the drop of peak cladding temperature was advanced by application of momentum flux term in cross-flow. Additionally, the authors developed a new input deck with multi-dimensional downcomer model and ran MARS code with multi-dimensional input deck as well. By using the modified input deck, the Emergency core cooling system by-pass flow phenomena is better characterized and found to be consistent with both experimental report and regulatory guide.

  19. Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mattie, Patrick D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-01-01

    Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this study was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.

  20. Scram discharge volume break studies accident sequence analysis

    Energy Technology Data Exchange (ETDEWEB)

    Harrington, R.M.; Hodge, S.A.

    1982-01-01

    This paper is a summary of a report describing the predicted response of Unit 1 at the Tennessee Valley Authority (TVA) Browns Ferry Nuclear Plant to a hypothetical small break loss of coolant accident (SBLOCA) outside of containment. The accident studied would be initiated by a break in the scram discharge volume (SDV) piping when it is pressurized to full reactor vessel pressure as a normal consequence of a reactor scram. If the scram could be reset, the scram outlet valves would close to isolate the SDV and the piping break from the reactor vessel. However, reset is possible only if the conditions that caused the scram have cleared; it has been assumed in this study that the scram signal remains in effect over a long period of time.

  1. Finite element based stress analysis of BWR internals exposed to accident loads

    Energy Technology Data Exchange (ETDEWEB)

    Altstadt, E.; Weiss, F.P.; Werner, M.; Willschuetz, H.G.

    1998-10-01

    During a hypothetical accident the reactor pressure vessel internals of boiling water reactors can be exposed to considerable loads resulting from temperature gradients and pressure waves. Three dimensional FE models were developed for the core shroud, the upper and the lower core supporting structure, the steam separator pipes and the feed water distributor. The models of core shroud, upper core structure and lower core structure were coupled by means of the substructure technique. All FE models can be used for thermal and for structural mechanical analyses. As an example the FE analysis for the case of a station black-out scenario (loss of power supply for the main circulating pumps) with subsequent emergency core cooling is demonstrated. The transient temperature distributions within the core shroud and within the steam dryer pipes as well were calculated based on the fluid temperatures and the heat transfer coefficients provided by thermo-hydraulic codes. At the maximum temperature gradients in the core shroud, the mechanical stress distribution was computed in a static analysis with the actual temperature field being the load. (orig.)

  2. Accident sequence precursor analysis level 2/3 model development

    Energy Technology Data Exchange (ETDEWEB)

    Lui, C.H. [Nuclear Regulatory Commission, Washington, DC (United States); Galyean, W.J.; Brownson, D.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1997-02-01

    The US Nuclear Regulatory Commission`s Accident Sequence Precursor (ASP) program currently uses simple Level 1 models to assess the conditional core damage probability for operational events occurring in commercial nuclear power plants (NPP). Since not all accident sequences leading to core damage will result in the same radiological consequences, it is necessary to develop simple Level 2/3 models that can be used to analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude of the resulting radioactive releases to the environment, and calculate the consequences associated with these releases. The simple Level 2/3 model development work was initiated in 1995, and several prototype models have been completed. Once developed, these simple Level 2/3 models are linked to the simple Level 1 models to provide risk perspectives for operational events. This paper describes the methods implemented for the development of these simple Level 2/3 ASP models, and the linkage process to the existing Level 1 models.

  3. Simulations of argon accident scenarios in the ATLAS experimental cavern a safety analysis

    CERN Document Server

    Balda, F

    2002-01-01

    Some characteristic accidents in the ATLAS experimental cavern (UX15) are simulated by means of STAR-CD, a code using the "Finite-Volume" method. These accidents involve different liquid argon leaks from the barrel cryostat of the detector, thus causing the dispersion of the argon into the Muon Chamber region and the evaporation of the liquid. The subsequent temperature gradients and distribution of argon concentrations, as well as their evolution in time are simulated and discussed, with the purpose of analysing the dangers related to asphyxiation and to contact with cryogenic fluids for the working personnel. A summary of the theory that stands behind the code is also given. In order to validate the models, an experimental test on a liquid argon spill performed earlier is simulated, showing that the program is able to output reliable results. At the end, some safety-related recommendations are listed.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

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

  5. Development of Methodology for Spent Fuel Pool Severe Accident Analysis Using MELCOR Program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won-Tae; Shin, Jae-Uk [RETech. Co. LTD., Yongin (Korea, Republic of); Ahn, Kwang-Il [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The general reason why SFP severe accident analysis has to be considered is that there is a potential great risk due to the huge number of fuel assemblies and no containment in a SFP building. In most cases, the SFP building is vulnerable to external damage or attack. In contrary, low decay heat of fuel assemblies may make the accident processes slow compared to the accident in reactor core because of a great deal of water. In short, its severity of consequence cannot exclude the consideration of SFP risk management. The U.S. Nuclear Regulatory Commission has performed the consequence studies of postulated spent fuel pool accident. The Fukushima-Daiichi accident has accelerated the needs for the consequence studies of postulated spent fuel pool accidents, causing the nuclear industry and regulatory bodies to reexamine several assumptions concerning beyond-design basis events such as a station blackout. The tsunami brought about the loss of coolant accident, leading to the explosion of hydrogen in the SFP building. Analyses of SFP accident processes in the case of a loss of coolant with no heat removal have studied. Few studies however have focused on a long term process of SFP severe accident under no mitigation action such as a water makeup to SFP. USNRC and OECD have co-worked to examine the behavior of PWR fuel assemblies under severe accident conditions in a spent fuel rack. In support of the investigation, several new features of MELCOR model have been added to simulate both BWR fuel assembly and PWR 17 x 17 assembly in a spent fuel pool rack undergoing severe accident conditions. The purpose of the study in this paper is to develop a methodology of the long-term analysis for the plant level SFP severe accident by using the new-featured MELCOR program in the OPR-1000 Nuclear Power Plant. The study is to investigate the ability of MELCOR in predicting an entire process of SFP severe accident phenomena including the molten corium and concrete reaction. The

  6. Understanding Code Patterns - Analysis, Interpretation & Measurement

    CERN Document Server

    Dundas, Jitesh

    2011-01-01

    This research paper aims to find, analyze and understand code patterns in any software system and measure its quality by defining standards and proposing a formula for the same. Every code that is written can be divided into different code segments, each having its own impact on the overall system. We can analyze these code segments to get the code quality. The measures used in this paper include Lines of Code, Number of calls made by a module, Execution time, the system knowledge of user and developers, the use of generalization, inheritance, reusability and other object-oriented concepts. The entire software code is divided into code snippets, based on the logic that they implement. Each of these code snippets has an impact. This measure is called Impact Factor and is valued by the software developer and/or other system stakeholders. Efficiency = (Code Area / Execution Time) * Qr

  7. Analysis of Core Degradation in Fukushima Unit 1 Accident with MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il; Kim, Tae Woon; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    In this study, an accident analysis of Fukushima Daiichi Unit 1 was performed using MELCOR 1.8.6. The behavior of the initial stage of the accident was focused during 30 hours after the reactor scram, because it was predicted that the vessel failure (severe accident) occurred before 20 hours. A hydrogen explosion also occurred at about 24 hours after the accident, and thus the phenomenon of core degradation before 30 hours was highlighted. Moreover, the effect of the amount of fresh water injection on the core degradation was performed by changing the amount of injection water. It was expected that a large portion of the injection water could not reach the core because of leakage. Thus, core damage was observed according to the amount of water that reached the core. The plant geometries and operating conditions were obtained from TEPCO (Tokyo Electric Power Company) through the OECD/NEA BSAF (Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station) Project. An analysis of the Fukushima accident was also performed by Sandia National Laboratories, but several conditions were revised and added in this study. First, the flow rate of the steam into turbine and water into downcomer were considered at the initial stage of accident. The water level followed well the measured data by adding this mechanism. Second, SRV stuck open was included in this calculation. SRV stuck open can occur due to high temperature and frequent operation, and was modeled in calculation. The timing of SRV stuck open was closed to the timing of MSL failure, and thus the depressurization of RPV could have originated from both of MSL failure and SRV stuck open. The effect of injection water was observed, and it was found that the proper water injection can prevent a severe accident at the initial stage of the accident. In conclusion, an analysis of the severe accident occurring in Fukushima Unit 1 was conducted by using MELCOR. The analysis results were consistent with the

  8. Analysis of swimming pool accidents resulting in spinal cord injury.

    Science.gov (United States)

    Green, B A; Gabrielsen, M A; Hall, W J; O'Heir, J

    1980-04-01

    This paper is a summary of a study of 72 cases of swimming-pool accidents resulting in serious injuries with the potential of permanent disability. Sixty-four of the 72 cases resulted in spinal cord injuries, 57 of which involved quadriplegic lesions. The authors observed that the majority of these injuries resulted from a lack of good judgement and common sense rather than from intoxication or pool structural deficiencies. Also of note was the lack of appropriate first-aid and extrication rendered, as well as the absence of uniform treatment and care received by the majority of the patients.

  9. MELCOR code source term characteristics for fast SBO scenario of OPR1000 plant

    Energy Technology Data Exchange (ETDEWEB)

    Han, Seok Jung; Kim, Tae Woon; Park, Sun Hee; Ahn, Kwang Il [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Off site consequence analysis in Level 3 PSA is mainly affected by source terms release characteristics of nuclear plant. The severe accidents analysis codes for quantifying the source terms release characteristics, such as MELCOR or MAAP, could be available to provide the detailed information of these characteristics to assess offsite consequence. To utilize these characteristics from severe accident analysis codes, MELCOR code was used in a specific severe accident scenario, i.e., fast station black out (SBO) for OPR1000 plant.

  10. Risk-based Analysis of Construction Accidents in Iran During 2007-2011-Meta Analyze Study.

    Science.gov (United States)

    Amiri, Mehran; Ardeshir, Abdollah; Fazel Zarandi, Mohammad Hossein

    2014-04-01

    The present study aimed to investigate the characteristics of occupational accidents and frequency and severity of work related accidents in the construction industry among Iranian insured workers during the years 20072011. The Iranian Social Security Organization (ISSO) accident database containing 21,864 cases between the years 2007-2011 was applied in this study. In the next step, Total Accident Rate (TRA), Total Severity Index (TSI), and Risk Factor (RF) were defined. The core of this work is devoted to analyzing the data from different perspectives such as age of workers, occupation and construction phase, day of the week, time of the day, seasonal analysis, regional considerations, type of accident, and body parts affected. Workers between 15-19 years old (TAR=13.4%) are almost six times more exposed to risk of accident than the average of all ages (TAR=2.51%). Laborers and structural workers (TAR=66.6%) and those working at heights (TAR=47.2%) experience more accidents than other groups of workers. Moreover, older workers over 65 years old (TSI=1.97%> average TSI=1.60%), work supervisors (TSI=12.20% >average TSI=9.09%), and night shift workers (TSI=1.89% >average TSI=1.47%) are more prone to severe accidents. It is recommended that laborers, young workers, weekend and night shift workers be supervised more carefully in the workplace. Use of Personal Protective Equipment (PPE) should be compulsory in working environments, and special attention should be undertaken to people working outdoors and at heights. It is also suggested that policymakers pay more attention to the improvement of safety conditions in deprived and cold western regions.

  11. Risk-based Analysis of Construction Accidents in Iran During 2007-2011-Meta Analyze Study.

    Directory of Open Access Journals (Sweden)

    Mehran Amiri

    2015-04-01

    Full Text Available The present study aimed to investigate the characteristics of occupational accidents and frequency and severity of work related accidents in the construction industry among Iranian insured workers during the years 20072011.The Iranian Social Security Organization (ISSO accident database containing 21,864 cases between the years 2007-2011 was applied in this study. In the next step, Total Accident Rate (TRA, Total Severity Index (TSI, and Risk Factor (RF were defined. The core of this work is devoted to analyzing the data from different perspectives such as age of workers, occupation and construction phase, day of the week, time of the day, seasonal analysis, regional considerations, type of accident, and body parts affected.Workers between 15-19 years old (TAR=13.4% are almost six times more exposed to risk of accident than the average of all ages (TAR=2.51%. Laborers and structural workers (TAR=66.6% and those working at heights (TAR=47.2% experience more accidents than other groups of workers. Moreover, older workers over 65 years old (TSI=1.97%> average TSI=1.60%, work supervisors (TSI=12.20% >average TSI=9.09%, and night shift workers (TSI=1.89% >average TSI=1.47% are more prone to severe accidents.It is recommended that laborers, young workers, weekend and night shift workers be supervised more carefully in the workplace. Use of Personal Protective Equipment (PPE should be compulsory in working environments, and special attention should be undertaken to people working outdoors and at heights. It is also suggested that policymakers pay more attention to the improvement of safety conditions in deprived and cold western regions.

  12. Digital Image Analysis for Detechip Code Determination

    Directory of Open Access Journals (Sweden)

    Marcus Lyon

    2012-08-01

    Full Text Available DETECHIP® is a molecular sensing array used for identification of a large variety of substances. Previous methodology for the analysis of DETECHIP® used human vision to distinguish color changes induced by the presence of the analyte of interest. This paper describes several analysis techniques using digital images of DETECHIP® . Both a digital camera and flatbed desktop photo scanner were used to obtain Jpeg images. Color information within these digital images was obtained through the measurement of redgreen-blue (RGB values using software such as GIMP, Photoshop and ImageJ. Several different techniques were used to evaluate these color changes. It was determined that the flatbed scanner produced in the clearest and more reproducible images. Furthermore, codes obtained using a macro written for use within ImageJ showed improved consistency versus pervious methods.

  13. Development of PIRT and Assessment Matrix for Verification and Validation of Sodium Fire Analysis Codes

    Science.gov (United States)

    Ohno, Shuji; Ohshima, Hiroyuki; Tajima, Yuji; Ohki, Hiroshi

    Thermodynamic consequence in liquid sodium leak and fire accident is one of the important issues to be evaluated when considering the safety aspect of fast reactor plant building. The authors are therefore initiating systematic verification and validation (V&V) activity to assure and demonstrate the reliability of numerical simulation tool for sodium fire analysis. The V&V activity is in progress with the main focuses on already developed sodium fire analysis codes SPHINCS and AQUA-SF. The events to be evaluated are hypothetical sodium spray, pool, or combined fire accidents followed by thermodynamic behaviors postulated in a plant building. The present paper describes that the ‘Phenomena Identification and Ranking Table (PIRT)’ is developed at first for clarifying the important validation points in the sodium fire analysis codes, and that an ‘Assessment Matrix’ is proposed which summarizes both separate effect tests and integral effect tests for validating the computational models or whole code for important phenomena. Furthermore, the paper shows a practical validation with a separate effect test in which the spray droplet combustion model of SPHINCS and AQUA-SF predicts the burned amount of a falling sodium droplet with the error mostly less than 30%.

  14. Analysis, simulation and modeling of atmospheric stratification erosion with lumped parameter codes; Analyse, Simulation und Modellierung der Erosion atmosphaerischer Schichtungen mit Lumped Parameter-Codes

    Energy Technology Data Exchange (ETDEWEB)

    Burkhardt, Joerg

    2013-07-01

    The courses and consequences of severe accidents in nuclear power plants are usually simulated with the help of so called Lumped Parameter-Codes which are especially designed for this purpose. These codes are able to simulate complex physical phenomena within short computing times since they are based on a simplified zone principle. Furthermore they are provided with a simplified flow model basis. This dissertation aims at the ability of the German Containment Code System (COCOSYS) to simulate local accumulations of hydrogen. During severe accidents with a melting reactor core (as in Harrisburg or Fukushima) hydrogen can be generated and then be released to the containment. In case of a local accumulation a detonation can occur that endangers the buildings integrity. The results show that the development and the erosion of these hydrogen accumulations based on bouant flows are qualitatively well simulated. From a systematic grid study general rules concerning the simulation of the stratification erosion have been derivated. Those have been applied and confirmed by several blind code-benchmarks. A detailed analysis has shown that the simulated erosion rate and the resistance of simulated hydrogen accumulations are directly related to the grid discretisation chosen by the user. Based upon this analysis a model concept has been developed, which is able to detect hydrogen accumulations and to determine their intensity of interaction with impinging flows by non-dimensional numbers. The erosion flow is controlled by adjusting local grid effects. The model is in the development phase.

  15. Error-correction coding and decoding bounds, codes, decoders, analysis and applications

    CERN Document Server

    Tomlinson, Martin; Ambroze, Marcel A; Ahmed, Mohammed; Jibril, Mubarak

    2017-01-01

    This book discusses both the theory and practical applications of self-correcting data, commonly known as error-correcting codes. The applications included demonstrate the importance of these codes in a wide range of everyday technologies, from smartphones to secure communications and transactions. Written in a readily understandable style, the book presents the authors’ twenty-five years of research organized into five parts: Part I is concerned with the theoretical performance attainable by using error correcting codes to achieve communications efficiency in digital communications systems. Part II explores the construction of error-correcting codes and explains the different families of codes and how they are designed. Techniques are described for producing the very best codes. Part III addresses the analysis of low-density parity-check (LDPC) codes, primarily to calculate their stopping sets and low-weight codeword spectrum which determines the performance of these codes. Part IV deals with decoders desi...

  16. Analysis of occupational accidents: prevention through the use of additional technical safety measures for machinery.

    Science.gov (United States)

    Dźwiarek, Marek; Latała, Agata

    2016-01-01

    This article presents an analysis of results of 1035 serious and 341 minor accidents recorded by Poland's National Labour Inspectorate (PIP) in 2005-2011, in view of their prevention by means of additional safety measures applied by machinery users. Since the analysis aimed at formulating principles for the application of technical safety measures, the analysed accidents should bear additional attributes: the type of machine operation, technical safety measures and the type of events causing injuries. The analysis proved that the executed tasks and injury-causing events were closely connected and there was a relation between casualty events and technical safety measures. In the case of tasks consisting of manual feeding and collecting materials, the injuries usually occur because of the rotating motion of tools or crushing due to a closing motion. Numerous accidents also happened in the course of supporting actions, like removing pollutants, correcting material position, cleaning, etc.

  17. [Model of Analysis and Prevention of Accidents - MAPA: tool for operational health surveillance].

    Science.gov (United States)

    de Almeida, Ildeberto Muniz; Vilela, Rodolfo Andrade de Gouveia; da Silva, Alessandro José Nunes; Beltran, Sandra Lorena

    2014-12-01

    The analysis of work-related accidents is important for accident surveillance and prevention. Current methods of analysis seek to overcome reductionist views that see these occurrences as simple events explained by operator error. The objective of this paper is to analyze the Model of Analysis and Prevention of Accidents (MAPA) and its use in monitoring interventions, duly highlighting aspects experienced in the use of the tool. The descriptive analytical method was used, introducing the steps of the model. To illustrate contributions and or difficulties, cases where the tool was used in the context of service were selected. MAPA integrates theoretical approaches that have already been tried in studies of accidents by providing useful conceptual support from the data collection stage until conclusion and intervention stages. Besides revealing weaknesses of the traditional approach, it helps identify organizational determinants, such as management failings, system design and safety management involved in the accident. The main challenges lie in the grasp of concepts by users, in exploring organizational aspects upstream in the chain of decisions or at higher levels of the hierarchy, as well as the intervention to change the determinants of these events.

  18. Analysis of Human Errors in Industrial Incidents and Accidents for Improvement of Work Safety

    DEFF Research Database (Denmark)

    Leplat, J.; Rasmussen, Jens

    1984-01-01

    Methods for the analysis of work accidents are discussed, and a description is given of the use of a causal situation analysis in terms of a 'variation tree' in order to explain the course of events of the individual cases and to identify possible improvements. The difficulties in identifying 'ca...

  19. Human error and the problem of causality in analysis of accidents

    DEFF Research Database (Denmark)

    Rasmussen, Jens

    1990-01-01

    , designers or managers have played a major role. There are, however, several basic problems in analysis of accidents and identification of human error. This paper addresses the nature of causal explanations and the ambiguity of the rules applied for identification of the events to include in analysis......Present technology is characterized by complexity, rapid change and growing size of technical systems. This has caused increasing concern with the human involvement in system safety. Analyses of the major accidents during recent decades have concluded that human errors on part of operators...

  20. SiC MODIFICATIONS TO MELCOR FOR SEVERE ACCIDENT ANALYSIS APPLICATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Brad J. Merrill; Shannon M Bragg-Sitton

    2013-09-01

    The Department of Energy (DOE) Office of Nuclear Energy (NE) Light Water Reactor (LWR) Sustainability Program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. The Fuels Pathway within this program focuses on fuel system components outside of the fuel pellet, allowing for alteration of the existing zirconium-based clad system through coatings, addition of ceramic sleeves, or complete replacement (e.g. fully ceramic cladding). The DOE-NE Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC) is also conducting research on materials for advanced, accident tolerant fuels and cladding for application in operating LWRs. To aide in this assessment, a silicon carbide (SiC) version of the MELCOR code was developed by substituting SiC in place of Zircaloy in MELCOR’s reactor core oxidation and material property routines. The purpose of this development effort is to provide a numerical capability for estimating the safety advantages of replacing Zr-alloy components in LWRs with SiC components. This modified version of the MELCOR code was applied to the Three Mile Island (TMI-2) plant accident. While the results are considered preliminary, SiC cladding showed a dramatic safety advantage over Zircaloy cladding during this accident.

  1. Narrative Text Analysis of Accident Reports with Tractors, Self-Propelled Harvesting Machinery and Materials Handling Machinery in Austrian Agriculture from 2008 to 2010 – A Comparison

    OpenAIRE

    Hannes Mayrhofer; Elisabeth Quendler; Josef Boxberger

    2014-01-01

    The aim of this study was the identification of accident scenarios and causes by analysing existing accident reports of recognized agricultural occupational accidents with tractors, self-propelled harvesting machinery and materials handling machinery from 2008 to 2010. As a result of a literature-based evaluation of past accident analyses, the narrative text analysis was chosen as an appropriate method. A narrative analysis of the text fields of accident reports that farmers used to report ac...

  2. Recent development and application of a new safety analysis code for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J., E-mail: Brad.Merrill@inl.gov; Humrickhouse, Paul W.; Shimada, Masashi

    2016-11-01

    Highlights: • This paper presents recent code development activities for the MELCOR for fusion and Tritium Migration Analysis Program computer codes at the Idaho National Engineering Laboratory. • The capabilities of these computer codes are being merged into a single safety analysis tool for fusion reactor accidents. • The result of benchmarking these codes against previous code versions is presented by the authors of this paper. • This new capability is applied to study the tritium inventory and permeation rate for a water cold tungsten divertor that has neutron damage at 0.3 dpa. - Abstract: This paper describes the recent progress made in the development of two codes for fusion reactor safety assessments at the Idaho National Laboratory (INL): MELCOR for fusion and the Tritium Migration Analysis Program (TMAP). During the ITER engineering design activity (EDA), the INL Fusion Safety Program (FSP) modified the MELCOR 1.8.2 code for fusion applications to perform ITER thermal hydraulic safety analyses. Because MELCOR has undergone many improvements at SNL-NM since version 1.8.2 was released, the INL FSP recently imported these same fusion modifications into the MELCOR 1.8.6 code, along with the multiple fluids modifications of MELCOR 1.8.5 for fusion used in US advanced fusion reactor design studies. TMAP has also been under development for several decades at the INL by the FSP. TMAP treats multi-specie surface absorption and diffusion in composite materials with dislocation traps, plus the movement of these species from room to room by fluid flow within a given facility. Recently, TMAP was updated to consider multiple trap site types to allow the simulation of experimental data from neutron irradiated tungsten. The natural development path for both of these codes is to merge their capabilities into one computer code to provide a more comprehensive safety tool for analyzing accidents in fusion reactors. In this paper we detail recent developments in this

  3. Analysis of quantum error-correcting codes: Symplectic lattice codes and toric codes

    Science.gov (United States)

    Harrington, James William

    Quantum information theory is concerned with identifying how quantum mechanical resources (such as entangled quantum states) can be utilized for a number of information processing tasks, including data storage, computation, communication, and cryptography. Efficient quantum algorithms and protocols have been developed for performing some tasks (e.g. , factoring large numbers, securely communicating over a public channel, and simulating quantum mechanical systems) that appear to be very difficult with just classical resources. In addition to identifying the separation between classical and quantum computational power, much of the theoretical focus in this field over the last decade has been concerned with finding novel ways of encoding quantum information that are robust against errors, which is an important step toward building practical quantum information processing devices. In this thesis I present some results on the quantum error-correcting properties of oscillator codes (also described as symplectic lattice codes) and toric codes. Any harmonic oscillator system (such as a mode of light) can be encoded with quantum information via symplectic lattice codes that are robust against shifts in the system's continuous quantum variables. I show the existence of lattice codes whose achievable rates match the one-shot coherent information over the Gaussian quantum channel. Also, I construct a family of symplectic self-dual lattices and search for optimal encodings of quantum information distributed between several oscillators. Toric codes provide encodings of quantum information into two-dimensional spin lattices that are robust against local clusters of errors and which require only local quantum operations for error correction. Numerical simulations of this system under various error models provide a calculation of the accuracy threshold for quantum memory using toric codes, which can be related to phase transitions in certain condensed matter models. I also present

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

    Energy Technology Data Exchange (ETDEWEB)

    Tveten, U. (Institutt for Energiteknikk, Kjeller (Norway))

    1990-06-01

    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.

  5. SINFAC - SYSTEMS IMPROVED NUMERICAL FLUIDS ANALYSIS CODE

    Science.gov (United States)

    Costello, F. A.

    1994-01-01

    The Systems Improved Numerical Fluids Analysis Code, SINFAC, consists of additional routines added to the April 1983 revision of SINDA, a general thermal analyzer program. The purpose of the additional routines is to allow for the modeling of active heat transfer loops. The modeler can simulate the steady-state and pseudo-transient operations of 16 different heat transfer loop components including radiators, evaporators, condensers, mechanical pumps, reservoirs and many types of valves and fittings. In addition, the program contains a property analysis routine that can be used to compute the thermodynamic properties of 20 different refrigerants. SINFAC can simulate the response to transient boundary conditions. SINFAC was first developed as a method for computing the steady-state performance of two phase systems. It was then modified using CNFRWD, SINDA's explicit time-integration scheme, to accommodate transient thermal models. However, SINFAC cannot simulate pressure drops due to time-dependent fluid acceleration, transient boil-out, or transient fill-up, except in the accumulator. SINFAC also requires the user to be familiar with SINDA. The solution procedure used by SINFAC is similar to that which an engineer would use to solve a system manually. The solution to a system requires the determination of all of the outlet conditions of each component such as the flow rate, pressure, and enthalpy. To obtain these values, the user first estimates the inlet conditions to the first component of the system, then computes the outlet conditions from the data supplied by the manufacturer of the first component. The user then estimates the temperature at the outlet of the third component and computes the corresponding flow resistance of the second component. With the flow resistance of the second component, the user computes the conditions down stream, namely the inlet conditions of the third. The computations follow for the rest of the system, back to the first component

  6. Analysis on the nitrogen drilling accident of Well Qionglai 1 (I: Major inducement events of the accident

    Directory of Open Access Journals (Sweden)

    Yingfeng Meng

    2015-12-01

    Full Text Available Nitrogen drilling in poor tight gas sandstone should be safe because of very low gas production. But a serious accident of fire blowout occurred during nitrogen drilling of Well Qionglai 1. This is the first nitrogen drilling accident in China, which was beyond people's knowledge about the safety of nitrogen drilling and brought negative effects on the development of gas drilling technology still in start-up phase and resulted in dramatic reduction in application of gas drilling. In order to form a correct understanding, the accident was systematically analyzed, the major events resulting in this accident were inferred. It is discovered for the first time that violent ejection of rock clasts and natural gas occurred due to the sudden burst of downhole rock when the fractured tight gas zone was penetrated during nitrogen drilling, which has been named as “rock burst and blowout by gas bomb”, short for “rock burst”. Then all the induced events related to the rock burst are as following: upthrust force on drilling string from rock burst, bridging-off formed and destructed repeatedly at bit and centralizer, and so on. However, the most direct important event of the accident turns out to be the blockage in the blooie pipe from rock burst clasts and the resulted high pressure at the wellhead. The high pressure at the wellhead causes the blooie pipe to crack and trigged blowout and deflagration of natural gas, which is the direct presentation of the accident.

  7. Empirical Risk Analysis of Severe Reactor Accidents in Nuclear Power Plants after Fukushima

    Directory of Open Access Journals (Sweden)

    Jan Christian Kaiser

    2012-01-01

    Full Text Available Many countries are reexamining the risks connected with nuclear power generation after the Fukushima accidents. To provide updated information for the corresponding discussion a simple empirical approach is applied for risk quantification of severe reactor accidents with International Nuclear and Radiological Event Scale (INES level ≥5. The analysis is based on worldwide data of commercial nuclear facilities. An empirical hazard of 21 (95% confidence intervals (CI 4; 62 severe accidents among the world’s reactors in 100,000 years of operation has been estimated. This result is compatible with the frequency estimate of a probabilistic safety assessment for a typical pressurised power reactor in Germany. It is used in scenario calculations concerning the development in numbers of reactors in the next twenty years. For the base scenario with constant reactor numbers the time to the next accident among the world's 441 reactors, which were connected to the grid in 2010, is estimated to 11 (95% CI 3.7; 52 years. In two other scenarios a moderate increase or decrease in reactor numbers have negligible influence on the results. The time to the next accident can be extended well above the lifetime of reactors by retiring a sizeable number of less secure ones and by safety improvements for the rest.

  8. Study on coal mines accidents based on the grey relational analysis

    Institute of Scientific and Technical Information of China (English)

    WANG Shuai; ZHANG Jin-long

    2008-01-01

    The subject investigated the system of people-machine-environment in coal mines. The coal mines working process was researched and the theory of grey system was applied to analyze coal mines safety accidents and those relevant factors. This re-search reveals that this analysis method is easy and highly available and the result is of great credibility, which can not only provide theoretical supports to the quantitative study of coal mines safety accident, but offer basis for coal mines companies' safety management.

  9. The Cost of Company Occupational Accidents: An Activity Based Analysis using the SACA Method

    DEFF Research Database (Denmark)

    Rikhardsson, Pall M.; Impgaard, Martin

    The Systematic Accident Cost Analysis (SACA) project is a research project carried out during 2001 by The Aarhus School of Business and PricewaterhouseCoopers Denmark with financial support from The Danish National Working Environment Authority. It empirically tested a method - the SACA method...... - for evaluating the visible and hidden costs of corporate occupational accidents. It also focused on whether the registration, processing and reporting of these costs could be integrated in the corporate accounting information system. The project was based on case studies in 9 Danish companies within 3 different...

  10. The Importance of Bloodstain Pattern Analysis in the Investigation of Road Traffic Accidents: A Case Report

    Directory of Open Access Journals (Sweden)

    Younis M. Albalooshi

    2015-12-01

    Full Text Available Bloodstain pattern analysis has become a field of specialization in Forensic sciences and plays an important role in the reconstruction of events at a crime scene. Research, books, and articles have been published on the analysis and interpretation of bloodstain patterns We present a case study of a road traffic accident in which bloodstain pattern analysis helped us to solve the discrepancy between reports produced by forensic examiners and by the forensic biology department. The case was of a 22-year-old man who died immediately and a 31- year-old woman who survived a road traffic accident. They were both found outside their overturned car and it was impossible to ascertain from initial observations which of the victims was driving the car at the time of the accident. An external examination of the man revealed multiple injuries, and the cause of his death was severe brain injury. The woman survived with a fracture of the forearm, dislocated clavicle bone, and other minor injuries. After initial examination of the car and based on the pattern of injuries the deceased received, forensic examiner concluded that the man was the driving the car at the time of accident. On the other hand, the forensic DNA analysis of bloodstains obtained from the driver's seat matched that of the woman, suggesting that she was the driver. This apparent discrepancy directed the forensic examiner to carry out a bloodstain pattern analysis on the driver's seat. The bloodstain pattern analysis helped resolve the discrepancy and enabled the investigators to identify the driver correctly. This case report emphasizes the importance of bloodstain pattern analysis in the reconstruction of cases involving road traffic accidents.

  11. Final safety analysis report for the Galileo Mission: Volume 2, Book 2: Accident model document: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    1988-12-15

    This section of the Accident Model Document (AMD) presents the appendices which describe the various analyses that have been conducted for use in the Galileo Final Safety Analysis Report II, Volume II. Included in these appendices are the approaches, techniques, conditions and assumptions used in the development of the analytical models plus the detailed results of the analyses. Also included in these appendices are summaries of the accidents and their associated probabilities and environment models taken from the Shuttle Data Book (NSTS-08116), plus summaries of the several segments of the recent GPHS safety test program. The information presented in these appendices is used in Section 3.0 of the AMD to develop the Failure/Abort Sequence Trees (FASTs) and to determine the fuel releases (source terms) resulting from the potential Space Shuttle/IUS accidents throughout the missions.

  12. Analysis of Reactivity Induced Accident for Control Rods Ejection with Loss of Cooling

    CERN Document Server

    Saad, Hend Mohammed El Sayed; Wahab, Moustafa Aziz Abd El

    2013-01-01

    Understanding of the time-dependent behavior of the neutron population in nuclear reactor in response to either a planned or unplanned change in the reactor conditions, is a great importance to the safe and reliable operation of the reactor. In the present work, the point kinetics equations are solved numerically using stiffness confinement method (SCM). The solution is applied to the kinetics equations in the presence of different types of reactivities and is compared with different analytical solutions. This method is also used to analyze reactivity induced accidents in two reactors. The first reactor is fueled by uranium and the second is fueled by plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor. Both power and temperature pulse following the reactivity- initiated accidents are calculated. The results are compared with previous works and...

  13. Overhead analysis of universal concatenated quantum codes

    Science.gov (United States)

    Chamberland, Christopher; Jochym-O'Connor, Tomas; Laflamme, Raymond

    2017-02-01

    We analyze the resource overhead of recently proposed methods for universal fault-tolerant quantum computation using concatenated codes. Namely, we examine the concatenation of the 7-qubit Steane code with the 15-qubit Reed-Muller code, which allows for the construction of the 49- and 105-qubit codes that do not require the need for magic state distillation for universality. We compute a lower bound for the adversarial noise threshold of the 105-qubit code and find it to be 8.33 ×10-6. We obtain a depolarizing noise threshold for the 49-qubit code of 9.69 ×10-4 which is competitive with the 105-qubit threshold result of 1.28 ×10-3 . We then provide lower bounds on the resource requirements of the 49- and 105-qubit codes and compare them with the surface code implementation of a logical T gate using magic state distillation. For the sampled input error rates and noise model, we find that the surface code achieves a smaller overhead compared to our concatenated schemes.

  14. Chemical Analysis of Tire Traces in Traffic Accidents Investigation

    Directory of Open Access Journals (Sweden)

    Line Gueissaz

    2015-01-01

    Full Text Available The aim of the forensic investigation of traffic accidents is to help establish the nature and/or the circumstances of the event. This might be done with the purpose of determining the legal responsibilities of each person involved or to provide families, with a reconstruction of the events, to help understand why their relatives were injured or killed. A methodology for the comparison of chemical profiles of tire traces and tire tread samples obtained by pyrolysis-gas chromatography/mass spectrometry has been developed. Chemical profiles are represented by relative abundances of 86 compounds. The variability of the tread within and between 12 tires was assessed. Considering the level of the source as "brand and model" the intra-variability was found to be smaller than the inter-variability, leading to the complete discrimination of the 12 tires of the sample set. Braking tests were carried out on a racetrack in order to produce tire traces which origin was known. The results obtained with a supervised classification method showed that more than 94% of the replicates of the traces were correctly assigned to the class membership (i.e., brand and model of the tire at their origin. These results support that the chemical profile of one trace does not differ from the chemical profile of the tire at its origin but differs generally from the other chemical profiles of the sample set.

  15. A systemic analysis of South Korea Sewol ferry accident - Striking a balance between learning and accountability.

    Science.gov (United States)

    Kee, Dohyung; Jun, Gyuchan Thomas; Waterson, Patrick; Haslam, Roger

    2017-03-01

    The South Korea Sewol ferry accident in April 2014 claimed the lives of over 300 passengers and led to criminal charges of 399 personnel concerned including imprisonment of 154 of them as of Oct 2014. Blame and punishment culture can be prevalent in a more hierarchical society like South Korea as shown in the aftermath of this disaster. This study aims to analyse the South Korea ferry accident using Rasmussen's risk management framework and the associated AcciMap technique and to propose recommendations drawn from an AcciMap-based focus group with systems safety experts. The data for the accident analysis were collected mainly from an interim investigation report by the Board of Audit and Inspection of Korea and major South Korean and foreign newspapers. The analysis showed that the accident was attributed to many contributing factors arising from front-line operators, management, regulators and government. It also showed how the multiple factors including economic, social and political pressures and individual workload contributed to the accident and how they affected each other. This AcciMap was presented to 27 safety researchers and experts at 'the legacy of Jens Rasmussen' symposium adjunct to ODAM2014. Their recommendations were captured through a focus group. The four main recommendations include forgive (no blame and punishment on individuals), analyse (socio-technical system-based), learn (from why things do not go wrong) and change (bottom-up safety culture and safety system management). The findings offer important insights into how this type of accident should be understood, analysed and the subsequent response. Copyright © 2016 Elsevier Ltd. All rights reserved.

  16. Preliminary Analysis of Aircraft Loss of Control Accidents: Worst Case Precursor Combinations and Temporal Sequencing

    Science.gov (United States)

    Belcastro, Christine M.; Groff, Loren; Newman, Richard L.; Foster, John V.; Crider, Dennis H.; Klyde, David H.; Huston, A. McCall

    2014-01-01

    Aircraft loss of control (LOC) is a leading cause of fatal accidents across all transport airplane and operational classes, and can result from a wide spectrum of hazards, often occurring in combination. Technologies developed for LOC prevention and recovery must therefore be effective under a wide variety of conditions and uncertainties, including multiple hazards, and their validation must provide a means of assessing system effectiveness and coverage of these hazards. This requires the definition of a comprehensive set of LOC test scenarios based on accident and incident data as well as future risks. This paper defines a comprehensive set of accidents and incidents over a recent 15 year period, and presents preliminary analysis results to identify worst-case combinations of causal and contributing factors (i.e., accident precursors) and how they sequence in time. Such analyses can provide insight in developing effective solutions for LOC, and form the basis for developing test scenarios that can be used in evaluating them. Preliminary findings based on the results of this paper indicate that system failures or malfunctions, crew actions or inactions, vehicle impairment conditions, and vehicle upsets contributed the most to accidents and fatalities, followed by inclement weather or atmospheric disturbances and poor visibility. Follow-on research will include finalizing the analysis through a team consensus process, defining future risks, and developing a comprehensive set of test scenarios with correlation to the accidents, incidents, and future risks. Since enhanced engineering simulations are required for batch and piloted evaluations under realistic LOC precursor conditions, these test scenarios can also serve as a high-level requirement for defining the engineering simulation enhancements needed for generating them.

  17. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    Energy Technology Data Exchange (ETDEWEB)

    Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  18. Resolve! Version 2.5: Flammable Gas Accident Analysis Tool Acceptance Test Plan and Test Results

    Energy Technology Data Exchange (ETDEWEB)

    LAVENDER, J.C.

    2000-10-17

    RESOLVE! Version 2 .5 is designed to quantify the risk and uncertainty of combustion accidents in double-shell tanks (DSTs) and single-shell tanks (SSTs). The purpose of the acceptance testing is to ensure that all of the options and features of the computer code run; to verify that the calculated results are consistent with each other; and to evaluate the effects of the changes to the parameter values on the frequency and consequence trends associated with flammable gas deflagrations or detonations.

  19. ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    WILLIAMS, J.C.

    2003-11-15

    This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.

  20. ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    WILLIAMS, J.C.

    2003-11-15

    This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.

  1. System thermalhydraulics for design basis accident analysis and simulation: Status of tools and methods and direction for future R&D

    Energy Technology Data Exchange (ETDEWEB)

    Bestion, D., E-mail: dominique.bestion@cea.fr

    2017-02-15

    Highlights: • A state of the art on system code application is presented. • Requirements for demonstration of code up-scaling capabilities are proposed. • The role of multi-scale analysis in safety analysis is explained. • Uncertainty quantifications methodologies for system codes and CFD codes are compared and discussed. - Abstract: System thermalhydraulic investigations of Design Basis Accident require several tools and methods including the Process Identification and Ranking Table, the scaling, experiment analysis, modelling, code development, code Validation and Verification, and Uncertainty Quantification. This paper intends to give an overview of these methods and tools showing what the state of the art is, and presenting some recent advances. Recommendations are made with future direction for R&D including the need of new advanced experiments and instrumentation, and the future role of CFD and multi-scale analyses. For many people it is not clear what current system codes are, and what they can be. Then the main characteristics of these codes are recalled and propositions are made to clarify the code capabilities and limitations and to improve the knowledge of the conditions for a correct application of the codes for safety in a Best Estimate Plus Uncertainty approach. Also, the on-going developments of 3-field models and Transport of Interfacial Area are summarized and associated experimental needs are identified. The growing role of 3D modelling of reactor core and Pressure Vessel requires additional experimental data for a proper validation. CFD in open medium also contributes to investigations when 3D geometrical aspects play an important role. Recent activities performed in the OECD-NEA Working Group for Analysis and Management of Accidents is summarized and recent applications of two-phase CFD to boiling flows and two-phase PTS scenarios are reported. The role of a multi-scale approach for safety issues is illustrated with the LOCA transients

  2. Analysis of Adolescent Awareness of Radiation: Marking the First Anniversary of the Fukushima Nuclear Accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Bang Ju [Korean Science Reporters Association, Seoul (Korea, Republic of)

    2012-06-15

    Marking the first anniversary of the Fukushima nuclear accident, which took place on March 11th, 2011, the level of adolescent awareness and understanding of radiation was surveyed, and the results were then compared with those for adults with the same questionnaires conducted at similar times. A qualitative survey and frequency analysis were made for the design of the study methodology. Those surveyed were limited to 3rd grade middle school students, 15 years of age, who are the future generation. The questionnaire, which is a survey tool, was directly distributed to the students and 2,217 answers were analysed. The questionnaires were composed of 40 questions, and it was found that Cronbach's coefficient was high with 'self awareness of radiation' at 0.494, 'risk of radiation' at 0.843, 'benefit of radiation' at 0.748, 'radiological safety control' at 0.692, 'information sources of radiation' at 0.819, and 'impacts of Fukushima accident'. The results of the survey analysis showed that the students' knowledge of radiation was not very high with 67.4 points (69.5 points for adults) calculated on a maximum scale of 100 points (converted points). The impacts of the Fukushima nuclear accident were found to be less significant to adolescents than adults, and the rate of answer of 'so' or ' very so' in the following questions demonstrates this well. It was also shown that the impacts of the Fukushima accident to adolescents were comparatively low with 27.0% (38.9% for adults) on the question of 'attitude changed against nuclear power due to the Fukushima accident,' 65.7%(86.6% for adults) on the question of 'the damages from the Fukushima accident was immeasurably huge,' and 65.0% (86.3% for adults) on 'the Fukushima accident contributed to raising awareness on the safety of nuclear power plants'. The adolescents had a high rate of &apos

  3. Analysis of dental materials as an aid to identification in aircraft accidents

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.S.; Cruickshanks-Boyd, D.W.

    1982-04-01

    The failure to achieve positive identification of aircrew following an aircraft accident need not prevent a full autopsy and toxicological examination to ascertain possible medical factors involved in the accident. Energy-dispersive electron microprobe analysis provides morphological, qualitative, and accurate quantitative analysis of the composition of dental amalgam. Wet chemical analysis can be used to determine the elemental composition of crowns, bridges and partial dentures. Unfilled resin can be analyzed by infrared spectroscopy. Detailed analysis of filled composite restorative resins has not yet been achieved in the as-set condition to permit discrimination between manufacturers' products. Future work will involve filler studies and pyrolysis of the composite resins by thermogravimetric analysis to determine percentage weight loss when the sample examined is subjected to a controlled heating regime. With these available techniques, corroborative evidence achieved from the scientific study of materials can augment standard forensic dental results to obtain a positive identification.

  4. LPNORM: A linear programming normative analysis code

    Science.gov (United States)

    de Caritat, Patrice; Bloch, John; Hutcheon, Ian

    1994-04-01

    The computer code LPNORM implements the mathematical method of linear programming to calculate the mineralogical makeup of mineral mixtures, such as rock, sediment, or soil samples, from their bulk geochemical composition and from the mineralogical (or geochemical) composition of the contained minerals. This method simultaneously solves the set of linear equations governing the distribution of oxides into these minerals, subject to an objective function and a set of basic constraints. LPNORM allows the user to specify what minerals will be considered for normative analysis, what their composition is (in terms of mineral formula or geochemical composition), and whether to maximize mineral abundances, minimize slack variables (oxides that can not be accounted for), or do both at once in the objective function. Independent knowledge about the abundance of one or several of the minerals in the sample can be entered as additional equality or inequality constraints. Trial-and-error approach enables the user to "optimize" the composition of one or a few of the contained minerals. Results of comparative tests, highlighting the efficiency, as well as the shortcomings, of LPNORM are presented.

  5. Development and Verification of Smoothed Particle Hydrodynamics Code for Analysis of Tsunami near NPP

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Young Beom; Kim, Eung Soo [Seoul National Univ., Seoul (Korea, Republic of)

    2014-10-15

    It becomes more complicated when considering the shape and phase of the ground below the seawater. Therefore, some different attempts are required to precisely analyze the behavior of tsunami. This paper introduces an on-going activities on code development in SNU based on an unconventional mesh-free fluid analysis method called Smoothed Particle Hydrodynamics (SPH) and its verification work with some practice simulations. This paper summarizes the on-going development and verification activities on Lagrangian mesh-free SPH code in SNU. The newly developed code can cover equation of motions and heat conduction equation so far, and verification of each models is completed. In addition, parallel computation using GPU is now possible, and GUI is also prepared. If users change input geometry or input values, they can simulate for various conditions geometries. A SPH method has large advantages and potential in modeling of free surface, highly deformable geometry and multi-phase problems that traditional grid-based code has difficulties in analysis. Therefore, by incorporating more complex physical models such as turbulent flow, phase change, two-phase flow, and even solid mechanics, application of the current SPH code is expected to be much more extended including molten fuel behaviors in the sever accident.

  6. OPR1000 RCP Flow Coastdown Analysis using SPACE Code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong-Hyuk; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Korean nuclear industry developed a thermal-hydraulic analysis code for the safety analysis of PWRs, named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). Current loss of flow transient analysis of OPR1000 uses COAST code to calculate transient RCS(Reactor Coolant System) flow. The COAST code calculates RCS loop flow using pump performance curves and RCP(Reactor Coolant Pump) inertia. In this paper, SPACE code is used to reproduce RCS flowrates calculated by COAST code. The loss of flow transient is transient initiated by reduction of forced reactor coolant circulation. Typical loss of flow transients are complete loss of flow(CLOF) and locked rotor(LR). OPR1000 RCP flow coastdown analysis was performed using SPACE using simplified nodalization. Complete loss of flow(4 RCP trip) was analyzed. The results show good agreement with those from COAST code, which is CE code for calculating RCS flow during loss of flow transients. Through this study, we confirmed that SPACE code can be used instead of COAST code for RCP flow coastdown analysis.

  7. Parallelization of Subchannel Analysis Code MATRA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seongjin; Hwang, Daehyun; Kwon, Hyouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    A stand-alone calculation of MATRA code used up pertinent computing time for the thermal margin calculations while a relatively considerable time is needed to solve the whole core pin-by-pin problems. In addition, it is strongly required to improve the computation speed of the MATRA code to satisfy the overall performance of the multi-physics coupling calculations. Therefore, a parallel approach to improve and optimize the computability of the MATRA code is proposed and verified in this study. The parallel algorithm is embodied in the MATRA code using the MPI communication method and the modification of the previous code structure was minimized. An improvement is confirmed by comparing the results between the single and multiple processor algorithms. The speedup and efficiency are also evaluated when increasing the number of processors. The parallel algorithm was implemented to the subchannel code MATRA using the MPI. The performance of the parallel algorithm was verified by comparing the results with those from the MATRA with the single processor. It is also noticed that the performance of the MATRA code was greatly improved by implementing the parallel algorithm for the 1/8 core and whole core problems.

  8. Development of a three-dimensional model and calculation code for the packed bed simulation for safety analyses of severe reactor accidents; Entwicklung eines dreidimensionalen Modells und Rechencodes zur Simulation von Schuettbetten fuer Sicherheitsanalysen von schweren Reaktorstoerfaellen

    Energy Technology Data Exchange (ETDEWEB)

    Berkhan, Ana; Starflinger, Joerg [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme (IKE)

    2013-07-01

    The computer code MEWA is used for the description of severe accident sequences in light-water reactors. During the reactor accident with core disruption the solidified core fragments are displaced into the lower plenum of the reactor pressure vessel (RPV) or in case of RPV failure into the water filled reactor sump. For the progress or cessation of the severe accident the cooling of the packed bed is of main importance. With the 3D version of the code it is possible to study spatially complex packed beds with respect to their coolability. Further extension of the MEWA code will include the optimization for the improvement of the calculation efficiency and reduction of computation time. The validation will be performed by re-calculation of experiments (for instance DEBRIS experiments at the IKE) and the comparison with results of the 2D version.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  10. Implementation of numerical simulation techniques in analysis of the accidents in complex technological systems

    Energy Technology Data Exchange (ETDEWEB)

    Klishin, G.S.; Seleznev, V.E.; Aleoshin, V.V. [RFNC-VNIIEF (Russian Federation)

    1997-12-31

    Gas industry enterprises such as main pipelines, compressor gas transfer stations, gas extracting complexes belong to the energy intensive industry. Accidents there can result into the catastrophes and great social, environmental and economic losses. Annually, according to the official data several dozens of large accidents take place at the pipes in the USA and Russia. That is why prevention of the accidents, analysis of the mechanisms of their development and prediction of their possible consequences are acute and important tasks nowadays. The accidents reasons are usually of a complicated character and can be presented as a complex combination of natural, technical and human factors. Mathematical and computer simulations are safe, rather effective and comparatively inexpensive methods of the accident analysis. It makes it possible to analyze different mechanisms of a failure occurrence and development, to assess its consequences and give recommendations to prevent it. Besides investigation of the failure cases, numerical simulation techniques play an important role in the treatment of the diagnostics results of the objects and in further construction of mathematical prognostic simulations of the object behavior in the period of time between two inspections. While solving diagnostics tasks and in the analysis of the failure cases, the techniques of theoretical mechanics, of qualitative theory of different equations, of mechanics of a continuous medium, of chemical macro-kinetics and optimizing techniques are implemented in the Conversion Design Bureau {number_sign}5 (DB{number_sign}5). Both universal and special numerical techniques and software (SW) are being developed in DB{number_sign}5 for solution of such tasks. Almost all of them are calibrated on the calculations of the simulated and full-scale experiments performed at the VNIIEF and MINATOM testing sites. It is worth noting that in the long years of work there has been established a fruitful and effective

  11. Possibility of Fire Accident Analysis in Road Tunnel

    Directory of Open Access Journals (Sweden)

    Peter Vidmar

    2004-09-01

    Full Text Available The basic aim of this treatise is to research the relevant featuresof the control and management of fire event in road tunnels.The simulation of 200 meters long cut out of the road tunnelis performed with the computer code FDS (Fire DynamicsSimulator which is based on the computational fluid dynamics.FDS is used to simulate fire dynamics behavior in three differentscenarios: natural ventilation, forced ventilation with axialventilators and vertical smoke extraction.It is believed that the present research, the methodology appliedand its findings should promote and improve safety andreliability of fire safety in road tunnels and to cope with thecomplex contemporary logistic demands of safety and reliabilityin the transportation of passengers and goods.

  12. Development of auditing technology for accident analysis of SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Kim, H. C.; Bae, K. H.; Lee, Y. J.; Chung, Y. J.; Jeong, J. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2003-06-15

    The objective of this project is to develop thermal hydraulic models of the regulatory auditing codes for the application of SMART-P integrated reactor. The current year fall under the first step of the 3 year project, and the main researches were focused on identifying the candidate thermal hydraulic models for improvement. Well known PIRT methodology has been applied to identify model improvement items. As a part of PIRT process, the identification of SMART-P system and compenent has been performed. The scenario of each key accident and phenonema have been identified. To identify SMART-P thermal-hydraulic characteristics, preliminary calculation has been performed and identify the applicability and inprovement items of current auditing code, RELAP5.

  13. Challenges on innovations of newly-developed safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

    2016-05-15

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  14. Formal Analysis of an Airplane Accident in N{Σ}-Labeled Calculus

    Science.gov (United States)

    Mizutani, Tetsuya; Igarashi, Shigeru; Ikeda, Yasuwo; Shio, Masayuki

    N{Σ}-labeled calculus is a formal system for representation, verification and analysis of time-concerned recognition, knowledge, belief and decision of humans or computer programs together with related external physical or logical phenomena.In this paper, a formal verification and analysis of the JAL near miss accident is presented as an example of cooperating systems controlling continuously changing objects including human factor with misunderstanding or incorrect recognition.

  15. Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

  16. Manometer Behavior Analysis using CATHENA, RELAP and GOTHIC Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Han, Kee Soo; Moon, Bok Ja; Jang, Misuk [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    In this presentation, simple thermal hydraulic behavior is analyzed using three codes to show the possibility of using alternative codes. We established three models of simple u-tube manometer using three different codes. CATHENA (Canadian Algorithm for Thermal hydraulic Network Analysis), RELAP (Reactor Excursion and Leak Analysis Program), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are used for this analysis. CATHENA and RELAP are widely used codes for the analysis of system behavior of CANDU and PWR. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. In this paper, the internal behavior of u-tube manometer was analyzed using 3 codes, CATHENA, RELAP and GOTHIC. The general transient behavior is similar among 3 codes. However, the behavior simulated using GOTHIC shows some different trend compared with the results from the other 2 codes at the end of the transient. It would be resulted from the use of different physical model in GOTHIC, which is specialized for the multi-phase thermal hydraulic behavior analysis of containment system unlike the other two codes.

  17. Analysis of Severe Accident for the SFP under the Condition of Drainage using MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jung-Min; Pack, Jae-Woo [Jeju National University, Jeju (Korea, Republic of)

    2015-10-15

    This study aims to analyze the effect of a LOCA of the spent fuel pool. We use the MECORE 1.8.6 code to compute the variation of the fuel cladding temperature after a completer loss of the cooling water in the spent fuel pool. A loss of coolant accident in a typical spent fuel pool has been simulated using the MELCOR 1.8.6 code to see the variation of key parameters such as the oxygen concentration in the fuel assembly region and the cladding temperature. In a commercial nuclear power plant, highly radioactive spent fuel assemblies unloaded from the nuclear reactor core are typically stored for a period of time in the spent fuel pool to reduce the radioactivity. The spent fuel assemblies are usually placed in long square racks. It is known that in the progress of the Fukushima nuclear power plant accident, the cooling water in the spent fuel storage was completely lost and the fuel was heated up and damaged. The simulation result shows that the cladding temperature exceeds the rupture temperature in most of the fuel rods and some part of the fuel rods suffers melting of the cladding.

  18. Analysis of an AP600 intermediate-size loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Lime, J.F. [Los Alamos National Lab., NM (United States)

    1995-09-01

    A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

  19. Dynamical analysis of innovative core designs facing unprotected transients with the MAT5 DYN code

    Energy Technology Data Exchange (ETDEWEB)

    Darmet, G.; Massara, S. [EDF R and D, 1 avenue du general de Gaulle, 92140 Clamart (France)

    2012-07-01

    Since 2007, advanced Sodium-cooled Fast Reactors (SFR) are investigated by CEA, AREVA and EDF in the framework of a joint French collaboration. A prototype called ASTRID, sets out to demonstrate progress made in SFR technology, is due to operate in the years 2020's. The modeling of unprotected transients by computer codes is one of the key safety issues in the design approach to such SFR systems. For that purpose, the activity on CATHARE, which is the reference code for the transient analysis of ASTRID, has been strengthened during last years by CEA. In the meantime, EDF has developed a simplified and multi-channel code, named MAT5 DYN, to analyze and validate innovative core designs facing protected and unprotected transients. First, the paper consists in a description of MAT5 DYN: a code based on the existing code MAT4 DYN including major improvements on geometry description and physical modeling. Second, two core designs based on the CFV core design developed at CEA are presented. Then, the dynamic response of those heterogeneous cores is analyzed during unprotected loss of flow (ULOF) transient and unprotected transient of power (UTOP). The results highlight the importance of the low void core effect specific to the CFV design. Such an effect, when combined with a sufficient primary pump halving time and an optimized cooling group scheme, allows to delay (or, possibly, avoid) the sodium boiling onset during ULOF accidents. (authors)

  20. Numerical analysis of grid plate melting after a severe accident in a Fast-Breeder Reactor (FBR)

    Indian Academy of Sciences (India)

    A Jasmin Sudha; K Velusamy

    2013-12-01

    Fast breeder reactors (FBRs) are provided with redundant and diverse plant protection systems with a very low failure probability (<10-6/reactor year), making core disruptive accident (CDA), a beyond design basis event (BDBE). Nevertheless, safety analysis is carried out even for such events with a view to mitigate their consequences by providing engineered safeguards like the in-vessel core catcher. During a CDA, a significant fraction of the hot molten fuel moves downwards and gets relocated to the lower plate of grid plate. The ability of this plate to resist or delay relocation of core melt further has been investigated by developing appropriate mathematical models and translating them into a computer code HEATRAN-1. The core melt is a time dependent volumetric heat source because of the radioactive decay of the fission products which it contains. The code solves the nonlinear heat conduction equation including phase change. The analysis reveals that if the bottom of grid plate is considered to be adiabatic, melt-through of grid plate (i.e., melting of the entire thickness of the plate) occurs between 800 s and 1000 s depending upon the initial conditions. Knowledge of this time estimate is essential for defining the initial thermal load on the core catcher plate. If heat transfer from the bottom of grid plate to the underlying sodium is taken into account, then melt-through does not take place, but the temperature of grid plate is high enough to cause creep failure.

  1. DATA ANALYSIS OF TRAFFIC ACCIDENTS AND THEIR CAUSES IN GOMEL FOR 2013 AND 2014

    Directory of Open Access Journals (Sweden)

    S. A. Azemsha

    2015-01-01

    Full Text Available Measures undertaken for improvement of road traffic safety presuppose clampdown on violation of road traffic regulations but no attention is paid to the fact that there is no possibility to ensure road traffic safety due to inconformity of roads to the modern safety requirements. Therefore reduction in accident rate is connected with some changes in approaches to designing, construction and maintenance of roads. Nowadays when the number of automobiles is extremely large and their number is increasing with every passing year driver’s professionalism becomes the most significant factor. In these circumstances the professionalism is demonstrated not so much while driving in bad road conditions as it was previously but the professionalism is more revealed in the case when it is necessary to drive in conditions of large workloads and high manoeuvring rate when a special important role is given to the capability to forecast a situation, in other words the capability to read the road. Moral climate on the road is no less important as well and it practically fully depends on a driver.The paper contains an analysis of the Gomel traffic police data on quantitative distribution of road traffic accidents and their victims according to the time of day and month, lighting conditions, weather conditions, age and sex factors, types of violations. Situations of traffic behavior, motivations of drivers and road users, drivers training have been shown in terms of impact on the road traffic accident risk. The paper considers a human factor which rather often causes an accident. An evaluation has been given to such factor as road speed of transport facilities which exerts a significant influence on an accident risk and its severity.

  2. Implementation of the Resonance Analysis Code SAMMY

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The multi-level multi-channel R-matrix SAMMY code is used for making the resonance parameters,which was developed by Oak Ridge National Laboratory (ORNL), and widely used around the USA(ORELA, KAPL, LANL, TUNL...) and around the world (Belgium, Japan, France, Bulgaria, etc.).Thecode SAMMY is an important program to CNDC.

  3. ANALYSIS OF MEDIA PUBLICATIONS ON THE FUKUSHIMA NUCLEAR POWER PLANT ACCIDENT

    Directory of Open Access Journals (Sweden)

    I. A. Zykova

    2011-01-01

    Full Text Available The analysis of informing the public about radiation and radiation risks is made on the basiс of publications on the Fukushima nuclear accident after the earthquake and tsunami of March 11, 2011, by five newspapers and the official internet sites of the Federal Service for Surveillance on Consumer Rights Protection and Human Well-being and the Federal Service for Hydrometeorology and Environmental Monitoring. The analysis of the data suggests that the population is satisfied with its request in timely, clear, and reliable information on the radiation situation, protective measures and forecast of the situation development in the future. The possibility to learn about different viewpoints positively affects the mood of readers and reduces their anxiety. However, it should be recognized that, if the information have been provided by a single media only, the population would not have adequate understanding of the situation related to the accident at the Fukushima nuclear power plant.

  4. Boolean Algebra Application in Analysis of Flight Accidents

    Directory of Open Access Journals (Sweden)

    Casandra Venera BALAN

    2015-12-01

    Full Text Available Fault tree analysis is a deductive approach for resolving an undesired event into its causes, identifying the causes of a failure and providing a framework for a qualitative and quantitative evaluation of the top event. An alternative approach to fault tree analysis methods calculus goes to logical expressions and it is based on a graphical representation of the data structure for a logic - based binary decision diagram representation. In this analysis, such sites will be reduced to a minimal size and arranged in the sense that the variables appear in the same order in each path. An event can be defined as a statement that can be true or false. Therefore, Boolean algebra rules allow restructuring of a Fault Tree into one equivalent to it, but simpler.

  5. An association between dietary habits and traffic accidents in patients with chronic liver disease: A data-mining analysis.

    Science.gov (United States)

    Kawaguchi, Takumi; Suetsugu, Takuro; Ogata, Shyou; Imanaga, Minami; Ishii, Kumiko; Esaki, Nao; Sugimoto, Masako; Otsuyama, Jyuri; Nagamatsu, Ayu; Taniguchi, Eitaro; Itou, Minoru; Oriishi, Tetsuharu; Iwasaki, Shoko; Miura, Hiroko; Torimura, Takuji

    2016-05-01

    The incidence of traffic accidents in patients with chronic liver disease (CLD) is high in the USA. However, the characteristics of patients, including dietary habits, differ between Japan and the USA. The present study investigated the incidence of traffic accidents in CLD patients and the clinical profiles associated with traffic accidents in Japan using a data-mining analysis. A cross-sectional study was performed and 256 subjects [148 CLD patients (CLD group) and 106 patients with other digestive diseases (disease control group)] were enrolled; 2 patients were excluded. The incidence of traffic accidents was compared between the two groups. Independent factors for traffic accidents were analyzed using logistic regression and decision-tree analyses. The incidence of traffic accidents did not differ between the CLD and disease control groups (8.8 vs. 11.3%). The results of the logistic regression analysis showed that yoghurt consumption was the only independent risk factor for traffic accidents (odds ratio, 0.37; 95% confidence interval, 0.16-0.85; P=0.0197). Similarly, the results of the decision-tree analysis showed that yoghurt consumption was the initial divergence variable. In patients who consumed yoghurt habitually, the incidence of traffic accidents was 6.6%, while that in patients who did not consume yoghurt was 16.0%. CLD was not identified as an independent factor in the logistic regression and decision-tree analyses. In conclusion, the difference in the incidence of traffic accidents in Japan between the CLD and disease control groups was insignificant. Furthermore, yoghurt consumption was an independent negative risk factor for traffic accidents in patients with digestive diseases, including CLD.

  6. Light water reactor fuel analysis code FEMAXI-V (Ver.1)

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-09-01

    A light water fuel analysis code FEMAXI-V is an advanced version which has been produced by integrating FEMAXI-IV(Ver.2), high burn-up fuel code EXBURN-I, and a number of functional improvements and extensions, to predict fuel rod behavior in normal and transient (not accident) conditions. The present report describes in detail the basic theories and structure, models and numerical solutions applied, improvements and extensions, and the material properties adopted in FEMAXI-V(Ver.1). FEMAXI-V deals with a single fuel rod. It predicts thermal and mechanical response of fuel rod to irradiation, including FP gas release. The thermal analysis predicts rod temperature distribution on the basis of pellet heat generation, changes in pellet thermal conductivity and gap thermal conductance, (transient) change in surface heat transfer to coolant, using radial one-dimensional geometry. The heat generation density profile of pellet can be determined by adopting the calculated results of burning analysis code. The mechanical analysis performs elastic/plastic, creep and PCMI calculations by FEM. The FP gas release model calculates diffusion of FP gas atoms and accumulation in bubbles, release and increase in internal pressure of rod. In every analysis, it is possible to allow some materials properties and empirical equations to depend on the local burnup or heat flux, which enables particularly analysis of high burnup fuel behavior and boiling transient of BWR rod. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output of the code are also described, and a sample output in an actual form is included. (author)

  7. The Factors Analysis on Food Safety Accidents Statistics

    Directory of Open Access Journals (Sweden)

    Boyi Xiang

    2015-05-01

    Full Text Available The study uses SPSS17.0 analysis of validity and reliability of the food enterprises questionnaire. Using AMOS17. 0 software for structural equation model test of goodness of fit and analysis of on the path. From the “melamine” to “Sudanred” and “steroid-tainted pork” events that have been exposed recently, series of typical food safety incidents resulted in the emergence of food safety issues become the focus of attention. A series of food processing can be contaminated by harmful substances, resulting in harmful food, thus constituting food safety issues and poses a serious threat to public and person’s health.

  8. Accelerator-driven transmutation reactor analysis code system (ATRAS)

    Energy Technology Data Exchange (ETDEWEB)

    Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

  9. Validation of GAMMA+ model for Evaluating Heat Transfer of VHTR core in Accident Conditions by CFD analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Dongho; Yoon, Sujong; Park, Gooncherl; Cho, Hyoungkyu [Seoul National Univ., Seoul (Korea, Republic of)

    2013-05-15

    KAERI has established a plan to demonstrate massive production of hydrogen using a VHTR by the early 2020s. In addition the GAMMA+ code is developed to analyze VHTR thermo-fluid transients at KAERI. One of the candidate reactor designs for VHTR is prismatic modular reactor (PMR), of which reference reactor is the 600MWth GT-MHR. This type of reactor has a passive safety system. During the High Pressure Conduction Cooling (HPCC) or Low Pressure Conduction Cooling (LPCC) accident, the core heats up by decay heat and then starts to cool down by conduction and radiation cooling to the Reactor Cavity Cooling System (RCCS) through the prismatic core. In this mechanism, the solid conduction occurs in graphite and fuel blocks, and the gas conduction and radiation occurs in coolant holes and bypass gaps. It is important to predict conduction and radiation heat transfer in the core for safety analysis. Effective thermal conductivity is derived by Maxwell's far-field methodology Radiation effect is expressed as corresponding conductivity and added to gas conductivity. In this study, ETC model used in GAMMA+ code is validated with the commercial CFD code, CFX-13. In this study, the effective thermal conductivity model of the GAMMA+ was evaluated by comparison of CFD analysis. The CFD analysis was conducted for various numbers and volume fractions of coolant holes and temperatures. Although slight disagreement was shown for the cases run with small number of holes, the result of GAMMA+ model is accurate for the large numbers of holes sufficiently. Since there are 102 coolant holes and 210 fuel holes in a fuel block, it is concluded that GAMMA+ model is proper formula for predicting effective thermal conductivity of the VHTR fuel block. However, in high temperature region above 500 .deg. C, the GAMMA+ model underestimates the effective thermal conductivity since radiation heat transfer is not reflected precisely. Further researches on it seem to be necessary.

  10. Human and organisational factors in maritime accidents: analysis of collisions at sea using the HFACS.

    Science.gov (United States)

    Chauvin, Christine; Lardjane, Salim; Morel, Gaël; Clostermann, Jean-Pierre; Langard, Benoît

    2013-10-01

    Over the last decade, the shipping industry has implemented a number of measures aimed at improving its safety level (such as new regulations or new forms of team training). Despite this evolution, shipping accidents, and particularly collisions, remain a major concern. This paper presents a modified version of the Human Factors Analysis and Classification System, which has been adapted to the maritime context and used to analyse human and organisational factors in collisions reported by the Marine Accident and Investigation Branch (UK) and the Transportation Safety Board (Canada). The analysis shows that most collisions are due to decision errors. At the precondition level, it highlights the importance of the following factors: poor visibility and misuse of instruments (environmental factors), loss of situation awareness or deficit of attention (conditions of operators), deficits in inter-ship communications or Bridge Resource Management (personnel factors). At the leadership level, the analysis reveals the frequent planning of inappropriate operations and non-compliance with the Safety Management System (SMS). The Multiple Accident Analysis provides an important finding concerning three classes of accidents. Inter-ship communications problems and Bridge Resource Management deficiencies are closely linked to collisions occurring in restricted waters and involving pilot-carrying vessels. Another class of collisions is associated with situations of poor visibility, in open sea, and shows deficiencies at every level of the socio-technical system (technical environment, condition of operators, leadership level, and organisational level). The third class is characterised by non-compliance with the SMS. This study shows the importance of Bridge Resource Management for situations of navigation with a pilot on board in restricted waters. It also points out the necessity to investigate, for situations of navigation in open sea, the masters' decisions in critical conditions

  11. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Brown, J. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)] [and others

    1997-06-01

    This volume is the second of a two-volume document that summarizes a joint project by the US Nuclear Regulatory and the Commission of European Communities to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This two-volume report, which examines mechanisms and uncertainties of transfer through the food chain, is the first in a series of five such reports. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain transfer that affect calculations of offsite radiological consequences. Seven of the experts reported on transfer into the food chain through soil and plants, nine reported on transfer via food products from animals, and two reported on both. The expert judgment elicitation procedure and its outcomes are described in these volumes. This volume contains seven appendices. Appendix A presents a brief discussion of the MAACS and COSYMA model codes. Appendix B is the structure document and elicitation questionnaire for the expert panel on soils and plants. Appendix C presents the rationales and responses of each of the members of the soils and plants expert panel. Appendix D is the structure document and elicitation questionnaire for the expert panel on animal transfer. The rationales and responses of each of the experts on animal transfer are given in Appendix E. Brief biographies of the food chain expert panel members are provided in Appendix F. Aggregated results of expert responses are presented in graph format in Appendix G.

  12. Mathematical model and computer code for the analysis of advanced fast reactor dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Schukin, N.V. (Moscow Engineering Physics Inst. (Russian Federation)); Korsun, A.S. (Moscow Engineering Physics Inst. (Russian Federation)); Vitruk, S.G. (Moscow Engineering Physics Inst. (Russian Federation)); Zimin, V.G. (Moscow Engineering Physics Inst. (Russian Federation)); Romanin, S.D. (Moscow Engineering Physics Inst. (Russian Federation))

    1993-04-01

    Efficient algorithms for mathematical modeling of 3-D neutron kinetics and thermal hydraulics are described. The model and appropriate computer code make it possible to analyze a variety of transient events ranging from normal operational states to catastrophic accident excursions. To verify the code, a number of calculations of different kind of transients was carried out. The results of the calculations show that the model and the computer code could be used for conceptual design of advanced liquid metal reactors. The detailed description of calculations of TOP WS accident is presented. (orig./DG)

  13. Analysis of cost regression and post-accident absence

    Science.gov (United States)

    Wojciech, Drozd

    2017-07-01

    The article presents issues related with costs of work safety. It proves the thesis that economic aspects cannot be overlooked in effective management of occupational health and safety and that adequate expenditures on safety can bring tangible benefits to the company. Reliable analysis of this problem is essential for the description the problem of safety the work. In the article attempts to carry it out using the procedures of mathematical statistics [1, 2, 3].

  14. Impact of traffic congestion on road accidents: a spatial analysis of the M25 motorway in England.

    Science.gov (United States)

    Wang, Chao; Quddus, Mohammed A; Ison, Stephen G

    2009-07-01

    Traffic congestion and road accidents are two external costs of transport and the reduction of their impacts is often one of the primary objectives for transport policy makers. The relationship between traffic congestion and road accidents however is not apparent and less studied. It is speculated that there may be an inverse relationship between traffic congestion and road accidents, and as such this poses a potential dilemma for transport policy makers. This study aims to explore the impact of traffic congestion on the frequency of road accidents using a spatial analysis approach, while controlling for other relevant factors that may affect road accidents. The M25 London orbital motorway, divided into 70 segments, was chosen to conduct this study and relevant data on road accidents, traffic and road characteristics were collected. A robust technique has been developed to map M25 accidents onto its segments. Since existing studies have often used a proxy to measure the level of congestion, this study has employed a precise congestion measurement. A series of Poisson based non-spatial (such as Poisson-lognormal and Poisson-gamma) and spatial (Poisson-lognormal with conditional autoregressive priors) models have been used to account for the effects of both heterogeneity and spatial correlation. The results suggest that traffic congestion has little or no impact on the frequency of road accidents on the M25 motorway. All other relevant factors have provided results consistent with existing studies.

  15. Sensitivity Analysis of Evacuation Speed in Hypothetical NPP Accident by Earthquake

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung-yeop; Lim, Ho-Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Effective emergency response in emergency situation of nuclear power plant (NPP) can make consequences be different therefore it is regarded important when establishing an emergency response plan and assessing the risk of hypothetical NPP accident. Situation of emergency response can be totally changed when NPP accident caused by earthquake or tsunami is considered due to the failure of roads and buildings by the disaster. In this study evacuation speed has been focused among above various factors and reasonable evacuation speed in earthquake scenario has been investigated. Finally, sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Evacuation scenario can be entirely different in the situation of seismic hazard and the sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Various references were investigated and earthquake evacuation model has been developed considering that evacuees may convert their evacuation method from using a vehicle to walking when they face the difficulty of using a vehicle due to intense traffic jam, failure of buildings and roads, and etc. The population dose within 5 km / 30 km have been found to be increased in earthquake situation due to decreased evacuation speed and become 1.5 - 2 times in the severest earthquake evacuation scenario set up in this study. It is not agreed that using same emergency response model which is used for normal evacuation situations when performing level 3 probabilistic safety assessment for earthquake and tsunami event. Investigation of data and sensitivity analysis for constructing differentiated emergency response model in the event of seismic hazard has been carried out in this study.

  16. Video analysis of the biomechanics of a bicycle accident resulting in significant facial fractures.

    Science.gov (United States)

    Syed, Shameer H; Willing, Ryan; Jenkyn, Thomas R; Yazdani, Arjang

    2013-11-01

    This study aimed to use video analysis techniques to determine the velocity, impact force, angle of impact, and impulse to fracture involved in a video-recorded bicycle accident resulting in facial fractures. Computed tomographic images of the resulting facial injury are presented for correlation with data and calculations. To our knowledge, such an analysis of an actual recorded trauma has not been reported in the literature. A video recording of the accident was split into frames and analyzed using an image editing program. Measurements of velocity and angle of impact were obtained from this analysis, and the force of impact and impulse were calculated using the inverse dynamic method with connected rigid body segments. These results were then correlated with the actual fracture pattern found on computed tomographic imaging of the subject's face. There was an impact velocity of 6.25 m/s, impact angles of 14 and 6.3 degrees of neck extension and axial rotation, respectively, an impact force of 1910.4 N, and an impulse to fracture of 47.8 Ns. These physical parameters resulted in clinically significant bilateral mid-facial Le Fort II and III pattern fractures. These data confer further understanding of the biomechanics of bicycle-related accidents by correlating an actual clinical outcome with the kinematic and dynamic parameters involved in the accident itself and yielding a concrete evidence of the velocity, force, and impulse necessary to cause clinically significant facial trauma. These findings can aid in the design of protective equipment for bicycle riders to help avoid this type of injury.

  17. Application of Dynamic Probabilistic Safety Assessment Approach for Accident Sequence Precursor Analysis: Case Study for Steam Generator Tube Rupture

    National Research Council Canada - National Science Library

    Lee, Hansul; Kim, Taewan; Heo, Gyunyoung

    2017-01-01

    ...) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios...

  18. Safety culture and accident analysis--a socio-management approach based on organizational safety social capital.

    Science.gov (United States)

    Rao, Suman

    2007-04-11

    One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, the key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization--seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly.

  19. Core-seis: a code for LMFBR core seismic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chellapandi, P.; Ravi, R.; Chetal, S.C.; Bhoje, S.B. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India). Reactor Group

    1995-12-31

    This paper deals with a computer code CORE-SEIS specially developed for seismic analysis of LMFBR core configurations. For demonstrating the prediction capability of the code, results are presented for one of the MONJU reactor core mock ups which deals with a cluster of 37 subassemblies kept in water. (author). 3 refs., 7 figs., 2 tabs.

  20. Dependency graph for code analysis on emerging architectures

    Energy Technology Data Exchange (ETDEWEB)

    Shashkov, Mikhail Jurievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lipnikov, Konstantin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-08

    Direct acyclic dependency (DAG) graph is becoming the standard for modern multi-physics codes.The ideal DAG is the true block-scheme of a multi-physics code. Therefore, it is the convenient object for insitu analysis of the cost of computations and algorithmic bottlenecks related to statistical frequent data motion and dymanical machine state.

  1. Source Code Analysis Laboratory (SCALe) for Energy Delivery Systems

    Science.gov (United States)

    2010-12-01

    and standar - dizing them through the ISO/IEC process should eliminate many of the problems encountered at the NIST SATE and also increase the...view the source code, using both structured and unstruct ons of secure coding rules discovered. However, manua than automated analysis, and the

  2. Performance analysis of adaptive turbo coded modulation with time delay

    Institute of Scientific and Technical Information of China (English)

    伍守豪; 宋文涛; 罗汉文

    2004-01-01

    The method of data fitting is applied to obtain the BER expression for turbo coded modulation, and a fitting mathematical model is proposed, which resolves the problem that there is no exact BER expression for turbo coded modulation in performance analysis. With the time delay consideration, the performance of BER of adaptive turbo coded modulation is analyzed and simulated. The results show that adaptive turbo coded modulation is very sensitive to time delay. In order to meet the target BER requirement, the total time delay should be less than 0. 001/fD.

  3. Improvement of QR Code Recognition Based on Pillbox Filter Analysis

    OpenAIRE

    Jia-Shing Sheu; Kai-Chung Teng

    2013-01-01

    The objective of this paper is to perform the innovation design for improving the recognition of a captured QR code image with blur through the Pillbox filter analysis. QR code images can be captured by digital video cameras. Many factors contribute to QR code decoding failure, such as the low quality of the image. Focus is an important factor that affects the quality of the image. This study discusses the out-of-focus QR code image and aims to improve the recognition of the conte...

  4. 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

    Energy Technology Data Exchange (ETDEWEB)

    Chino, Masamichi; Yamazawa, Hiromi; Nagai, Haruyasu; Moriuchi, Shigeru [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ishikawa, Hirohiko

    1995-09-01

    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).

  5. Radiological accidents: analysis of the information disseminated by media and public acceptance of nuclear technology

    Energy Technology Data Exchange (ETDEWEB)

    Delgado, Jose Ubiratan; Tauhata, Luiz [Instituto de Radioprotecao e Dosimetria (IRD), Rio de Janeiro, RJ (Brazil); Garcia, Marcia Maria [Fundacao Inst. Oswaldo Cruz (FIOCRUZ), Rio de Janeiro, RJ (Brazil). Dept. de Virologia

    1995-12-31

    A methodology to treat quantitatively information by Media concerning a nuclear or a radiological accident is presented. It allows us to classify information according to the amount, importance and way of showing, into one indicator, named Information Equivalent. This establishes a procedure for analysis of released information and includes: number of head-lines, illustrations, printed lines, editorials, authorities quoted and so on. Interpretation becomes easier when the evolution and statistical trend of this indicator is observed. The application to evaluate the dissemination of the accident which took place in 1987 in Goiania, Brazil, was satisfactory and allowed us to purpose a model. This will aid the planning, the decision making process and it will improve relationships between technical staff and media during the emergency. (author). 5 refs., 4 figs., 3 tabs.

  6. Thermal-hydraulic analysis of SWAMUP facility using ATHLET-SC code

    Directory of Open Access Journals (Sweden)

    Zidi eWang

    2015-03-01

    Full Text Available 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 the supercritical pressure to the subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. In order to simulate the trans-critical transient, a number of system codes for SCWR have been developed up to date. However, the validation work for the trans-critical models in these codes is still missing. The test facility Supercritical WAter MUltiPurpose loop (SWAMUP with 2×2 rod bundle in Shanghai Jiao Tong University (SJTU will be applied to provide test data for code validation. Some pre-test calculations are important and necessary to show the feasibility of the experiment. In this study, trans-critical transient analysis is performed for the SWAMUP facility with the system code ATHLET-SC, which is modified in SJTU, for supercritical water system. This paper presents the system behavior e.g. system pressure, coolant mass flow, cladding temperature during the depressurization. The effects of some important parameters such as heating power, depressurization rate on the system characteristics are also investigated in this paper. Additionally, some sensitivities study of the code models, e.g. heat transfer coefficient, CHF correlation, are analyzed and discussed. The results indicate that the revised system code ATHLET-SC is capable of simulating thermal hydraulic behavior during the trans-critical transient. According to the results, the cladding temperature during the transient is kept at a low value. However, the pressure difference of the heat exchanger after depressurization could reach 6 MPa, which should be considered in the experiment.

  7. Analysis of pedestrian accident costs in Sudan using the willingness-to-pay method.

    Science.gov (United States)

    Mofadal, Adam I A; Kanitpong, Kunnawee; Jiwattanakulpaisarn, Piyapong

    2015-05-01

    The willingness-to-pay (WTP) with contingent valuation (CV) method has been proven to be a valid tool for the valuation of non-market goods or socio-economic costs of road traffic accidents among communities in developed and developing countries. Research on accident costing tends to estimate the value of statistical life (VOSL) for all road users by providing a principle for the evaluation of road safety interventions in cost-benefit analysis. As in many other developing countries, the economic loss of traffic accidents in Sudan is noticeable; however, analytical research to estimate the magnitude and impact of that loss is lacking. Reports have shown that pedestrians account for more than 40% of the total number of fatalities. In this study, the WTP-CV approach was used to determine the amount of money that pedestrians in Sudan are willing to pay to reduce the risk of their own death. The impact of the socioeconomic factors, risk levels, and walking behaviors of pedestrians on their WTP for fatality risk reduction was also evaluated. Data were collected from two cities-Khartoum and Nyala-using a survey questionnaire that included 1400 respondents. The WTP-CV Payment Card Questionnaire was designed to ensure that Sudan pedestrians can easily determine the amount of money that would be required to reduce the fatality risk from a pedestrian-related accident. The analysis results show that the estimated VOSL for Sudanese pedestrians ranges from US$0.019 to US$0.101 million. In addition, the willingness-to-pay by Sudanese pedestrians to reduce their fatality risk tends to increase with age, household income, educational level, safety perception, and average time spent on social activities with family and community.

  8. Safety analysis methodology for Chinshan nuclear power plant spent fuel pool under Fukushima-like accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Hao-Tzu [Institute of Nuclear Energy Research, Taoyuan, Taiwan (China). Research Atomic Energy Council; Li, Wan-Yun; Wang, Jong-Rong; Tseng, Yung-Shin; Chen, Hsiung-Chih; Shih, Chunkuan; Chen, Shao-Wen [National Tsing Hua Univ., HsinChu, Taiwan (China). Inst. of Nuclear Engineering and Science

    2017-03-15

    Chinshan nuclear power plant (NPP), a BWR/4 plant, is the first NPP in Taiwan. After Fukushima NPP disaster occurred, there is more concern for the safety of NPPs in Taiwan. Therefore, in order to estimate the safety of Chinshan NPP spent fuel pool (SFP), by using TRACE, MELCOR, CFD, and FRAPTRAN codes, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of Chinshan NPP SFP. There were two main steps in this research. The first step was the establishment of Chinshan NPP SFP models. And the transient analysis under the SFP cooling system failure condition (Fukushima-like accident) was performed. In addition, the sensitive study of the time point for water spray was also performed. The next step was the fuel rod performance analysis by using FRAPTRAN and TRACE's results. Finally, the animation model of Chinshan NPP SFP was presented by using the animation function of SNAP with MELCOR analysis results.

  9. Questions concerning safety and risk after the nuclear accidents in Japan. Deepened accident analysis for the Fukushima Daiichi power plant; Sicherheits- und Risikofragen im Nachgang zu den nuklearen Stoer- und Unfaellen in Japan. Vertiefte Ereignisanalyse zur Anlage Fukushima-Daini

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Englert, Matthias [Oeko-Institut e.V. - Institut fuer Angewandte Oekologie, Darmstadt (Germany)

    2015-02-25

    The study questions concerning safety and risk in Japanese power plants following the disastrous nuclear accident covers the following issues: the nuclear facility Fukushima Daiichi, site characterization, important technical equipment, important electro-technical equipment, personal; description of the accident progression in the Fukushima nuclear power plant: impact of the earthquake, impact of the tsunami, short-term measures of the operating personnel, pressure and temperature situation in the containments, restoration of the after-heat cooling system in the units 1/2 and 4, fuel element storage pool, summarized parameters during the accident progress; comparative analysis of the accident progression at the Fukushima Daiichi site.

  10. Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a Level 2 probabilistic safety assessment or severe accident management strategy developments. The code has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Castillo G, F.

    2015-07-01

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

  13. Analysis 320 coal mine accidents using structural equation modeling with unsafe conditions of the rules and regulations as exogenous variables.

    Science.gov (United States)

    Zhang, Yingyu; Shao, Wei; Zhang, Mengjia; Li, Hejun; Yin, Shijiu; Xu, Yingjun

    2016-07-01

    Mining has been historically considered as a naturally high-risk industry worldwide. Deaths caused by coal mine accidents are more than the sum of all other accidents in China. Statistics of 320 coal mine accidents in Shandong province show that all accidents contain indicators of "unsafe conditions of the rules and regulations" with a frequency of 1590, accounting for 74.3% of the total frequency of 2140. "Unsafe behaviors of the operator" is another important contributory factor, which mainly includes "operator error" and "venturing into dangerous places." A systems analysis approach was applied by using structural equation modeling (SEM) to examine the interactions between the contributory factors of coal mine accidents. The analysis of results leads to three conclusions. (i) "Unsafe conditions of the rules and regulations," affect the "unsafe behaviors of the operator," "unsafe conditions of the equipment," and "unsafe conditions of the environment." (ii) The three influencing factors of coal mine accidents (with the frequency of effect relation in descending order) are "lack of safety education and training," "rules and regulations of safety production responsibility," and "rules and regulations of supervision and inspection." (iii) The three influenced factors (with the frequency in descending order) of coal mine accidents are "venturing into dangerous places," "poor workplace environment," and "operator error."

  14. Performance Analysis of 3-Dimensional Turbo Codes

    CERN Document Server

    Rosnes, Eirik

    2011-01-01

    In this work, we consider the minimum distance properties and convergence thresholds of 3-dimensional turbo codes (3D-TCs), recently introduced by Berrou et al.. Here, we consider binary 3D-TCs while the original work of Berrou et al. considered double-binary codes. In the first part of the paper, the minimum distance properties are analyzed from an ensemble perspective, both in the finite-length regime and in the asymptotic case of large block lengths. In particular, we analyze the asymptotic weight distribution of 3D-TCs and show numerically that their typical minimum distance dmin may, depending on the specific parameters, asymptotically grow linearly with the block length, i.e., the 3D-TC ensemble is asymptotically good for some parameters. In the second part of the paper, we derive some useful upper bounds on the dmin when using quadratic permutation polynomial (QPP) interleavers with a quadratic inverse. Furthermore, we give examples of interleaver lengths where an upper bound appears to be tight. The b...

  15. Analysis of Accidents in Nine Iranian Gas Refineries: 2007–2011

    OpenAIRE

    AR Shakibmanesh; Bolouri, A.; R Mehrdad

    2013-01-01

    Background: Occupational accidents are one of the major health hazards in industries and associated with high mortality, morbidity, spiritual damage and economic losses in the world.Objective: To determine the incidence of occupational accidents in 9 Iranian gas refineries between March 2007 and February 2011.Methods: Data on all occupational accidents occurred between March 2007 and February 2011, as well as other possible associated variables including time of accident, whether the accident...

  16. Comparative Analysis of CTF and Trace Thermal-Hydraulic Codes Using OECD/NRC PSBT Benchmark Void Distribution Database

    Directory of Open Access Journals (Sweden)

    M. Avramova

    2013-01-01

    Full Text Available The international OECD/NRC PSBT benchmark has been established to provide a test bed for assessing the capabilities of thermal-hydraulic codes and to encourage advancement in the analysis of fluid flow in rod bundles. The benchmark was based on one of the most valuable databases identified for the thermal-hydraulics modeling developed by NUPEC, Japan. The database includes void fraction and departure from nucleate boiling measurements in a representative PWR fuel assembly. On behalf of the benchmark team, PSU in collaboration with US NRC has performed supporting calculations using the PSU in-house advanced thermal-hydraulic subchannel code CTF and the US NRC system code TRACE. CTF is a version of COBRA-TF whose models have been continuously improved and validated by the RDFMG group at PSU. TRACE is a reactor systems code developed by US NRC to analyze transient and steady-state thermal-hydraulic behavior in LWRs and it has been designed to perform best-estimate analyses of LOCA, operational transients, and other accident scenarios in PWRs and BWRs. The paper presents CTF and TRACE models for the PSBT void distribution exercises. Code-to-code and code-to-data comparisons are provided along with a discussion of the void generation and void distribution models available in the two codes.

  17. Analysis of National Major Work Safety Accidents in China, 2003–2012

    Science.gov (United States)

    YE, Yunfeng; ZHANG, Siheng; RAO, Jiaming; WANG, Haiqing; LI, Yang; WANG, Shengyong; DONG, Xiaomei

    2016-01-01

    Background: This study provides a national profile of major work safety accidents in China, which cause more than 10 fatalities per accident, intended to provide scientific basis for prevention measures and strategies to reduce major work safety accidents and deaths. Methods: Data from 2003–2012 Census of major work safety accidents were collected from State Administration of Work Safety System (SAWS). Published literature and statistical yearbook were also included to implement information. We analyzed the frequency of accidents and deaths, trend, geographic distribution and injury types. Additionally, we discussed the severity and urgency of emergency rescue by types of accidents. Results: A total of 877 major work safety accidents were reported, resulting in 16,795 deaths and 9,183 injuries. The numbers of accidents and deaths, mortality rate and incidence of major accidents have declined in recent years. The mortality rate and incidence was 0.71 and 1.20 per 106 populations in 2012, respectively. Transportation and mining contributed to the highest number of major accidents and deaths. Major aviation and railway accidents caused more casualties per incident, while collapse, machinery, electrical shock accidents and tailing dam accidents were the most severe situation that resulted in bigger proportion of death. Conclusion: Ten years’ major work safety accident data indicate that the frequency of accidents and number of eaths was declined and several safety concerns persist in some segments. PMID:27057515

  18. Acceptable Risk Analysis for Abrupt Environmental Pollution Accidents in Zhangjiakou City, China.

    Science.gov (United States)

    Du, Xi; Zhang, Zhijiao; Dong, Lei; Liu, Jing; Borthwick, Alistair G L; Liu, Renzhi

    2017-04-20

    Abrupt environmental pollution accidents cause considerable damage worldwide to the ecological environment, human health, and property. The concept of acceptable risk aims to answer whether or not a given environmental pollution risk exceeds a societally determined criterion. This paper presents a case study on acceptable environmental pollution risk conducted through a questionnaire survey carried out between August and October 2014 in five representative districts and two counties of Zhangjiakou City, Hebei Province, China. Here, environmental risk primarily arises from accidental water pollution, accidental air pollution, and tailings dam failure. Based on 870 valid questionnaires, demographic and regional differences in public attitudes towards abrupt environmental pollution risks were analyzed, and risk acceptance impact factors determined. The results showed females, people between 21-40 years of age, people with higher levels of education, public servants, and people with higher income had lower risk tolerance. People with lower perceived risk, low-level risk knowledge, high-level familiarity and satisfaction with environmental management, and without experience of environmental accidents had higher risk tolerance. Multiple logistic regression analysis indicated that public satisfaction with environmental management was the most significant factor in risk acceptance, followed by perceived risk of abrupt air pollution, occupation, perceived risk of tailings dam failure, and sex. These findings should be helpful to local decision-makers concerned with environmental risk management (e.g., selecting target groups for effective risk communication) in the context of abrupt environmental accidents.

  19. Analysis of 78 cases of prehospital death due to traffic accident injury

    Institute of Scientific and Technical Information of China (English)

    胡孝菽; 洪勇; 等

    1999-01-01

    Objective The cause and time of prehospital death for the injured patients caused by traffic accidents were studied in order to improve traffic management and clinical treatment,and reduce mortality.Methods The characteristics of the injury,the rescue procedure,the status of the injury leading to death were analyzed based on the retrospective data of 78 cases died before admission.Results The main causes of prehospital death in the traffic accidents included:1.head injury,2.bleeding,3.chest and heart wound,4.spinal cord injury at upper cervix.Death happened immediately after injury was in 17 cases.Death happened from the accident site to our hospital was in 47 cases.Death happened within half an hour after reaching emergency room was in 14 cases.In all of the cases,the death on the transfer took up 62.5%.Conclusions Findings from analysis of the data will be presented on a wide range of traffic safety issues.These include enhancing education of traffic safety and administration of drivers and motor vehicles,establishing a perfect emergency medical service system and a well-trained team of first aid,and popularizing first aid knowledge to all people.

  20. Improvement of QR Code Recognition Based on Pillbox Filter Analysis

    Directory of Open Access Journals (Sweden)

    Jia-Shing Sheu

    2013-04-01

    Full Text Available The objective of this paper is to perform the innovation design for improving the recognition of a captured QR code image with blur through the Pillbox filter analysis. QR code images can be captured by digital video cameras. Many factors contribute to QR code decoding failure, such as the low quality of the image. Focus is an important factor that affects the quality of the image. This study discusses the out-of-focus QR code image and aims to improve the recognition of the contents in the QR code image. Many studies have used the pillbox filter (circular averaging filter method to simulate an out-of-focus image. This method is also used in this investigation to improve the recognition of a captured QR code image. A blurred QR code image is separated into nine levels. In the experiment, four different quantitative approaches are used to reconstruct and decode an out-of-focus QR code image. These nine reconstructed QR code images using methods are then compared. The final experimental results indicate improvements in identification.

  1. Behavior analysis of container ship in maritime accident in order to redefine the operating criteria

    Science.gov (United States)

    Ancuţa, C.; Stanca, C.; Andrei, C.; Acomi, N.

    2017-08-01

    In order to enhance the efficiency of maritime transport, container ships operators proceeded to increase the sizes of ships. The latest generation of ships in operation has 19,000 TEU capacity and the perspective is 21,000 TEU within the next years. The increasing of the sizes of container ships involves risks of maritime accidents occurrences. Nowadays, the general rules on operational security, tend to be adjusted as a result of the evaluation for each vessel. To create the premises for making an informed decision, the captain have to be aware of ships behavior in such situations. Not less important is to assure permanent review of the procedures for operation of ship, including the specific procedures in special areas, confined waters or separation schemes. This paper aims at analysing the behavior of the vessel and the respond of the structure of a container ship in maritime accident, in order to redefine the operating criteria. The method selected by authors for carrying out the research is computer simulations. Computer program provides the responses of the container ship model in various situations. Therefore, the simulations allow acquisition of a large category of data, in the scope of improving the prevention of accidents or mitigation of effects as much as possible. Simulations and assessments of certain situations that the ship might experience will be carried out to redefine the operating criteria. The envisaged scenarios are: introducing of maneuver speed for specific areas with high risk of collision or grounding, introducing of flooding scenarios of some compartments in loading programs, conducting of complex simulations in various situations for each vessel type. The main results of this work are documented proposals for operating criteria, intended to improve the safety in case of marine accidents, collisions and groundings. Introducing of such measures requires complex cost benefit analysis, that should not neglect the extreme economic impact

  2. Road accidents at night in the Netherlands : a national analysis according to official road accident data. Contribution to OECD Research Group TS 3 on Improving Road Safety at Night.

    NARCIS (Netherlands)

    Harris, S.

    1979-01-01

    The questionnaire about night-time accident data of the OECD Research Group TS 3 on Improving Road Safety at Night was filled in for the Netherlands. Thereafter a national analysis was written, using the already completed accident data questionnaire. Guidelines for the contents and presentation fo

  3. Radiation protection: an analysis of thyroid blocking. [Effectiveness of KI in reducing radioactive uptake following potential reactor accident

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D C; Blond, R M

    1980-01-01

    An analysis was performed to provide guidance to policymakers concerning the effectiveness of potassium iodide (KI) as a thyroid blocking agent in potential reactor accident situations, the distance to which (or area within which) it should be distributed, and its relative effectiveness compared to other available protective measures. The analysis was performed using the Reactor Safety Study (WASH-1400) consequence model. Four categories of accidents were addressed: gap activity release accident (GAP), GAP without containment isolation, core melt with a melt-through release, and core melt with an atmospheric release. Cost-benefit ratios (US $/thyroid nodule prevented) are given assuming that no other protective measures are taken. Uncertainties due to health effects parameters, accident probabilities, and costs are assessed. The effects of other potential protective measures, such as evacuation and sheltering, and the impact on children (critical population) are evaluated. Finally, risk-benefit considerations are briefly discussed.

  4. Improvement of reflood model in RELAP5 code based on sensitivity analysis

    Energy Technology Data Exchange (ETDEWEB)

    Li, Dong; Liu, Xiaojing; Yang, Yanhua, E-mail: yanhuay@sjtu.edu.cn

    2016-07-15

    Highlights: • Sensitivity analysis is performed on the reflood model of RELAP5. • The selected influential models are discussed and modified. • The modifications are assessed by FEBA experiment and better predictions are obtained. - Abstract: Reflooding is an important and complex process to the safety of nuclear reactor during loss of coolant accident (LOCA). Accurate prediction of the reflooding behavior is one of the challenge tasks for the current system code development. RELAP5 as a widely used system code has the capability to simulate this process but with limited accuracy, especially for low inlet flow rate reflooding conditions. Through the preliminary assessment with six FEBA (Flooding Experiments with Blocked Arrays) tests, it is observed that the peak cladding temperature (PCT) is generally underestimated and bundle quench is predicted too early compared to the experiment data. In this paper, the improvement of constitutive models related to reflooding is carried out based on single parametric sensitivity analysis. Film boiling heat transfer model and interfacial friction model of dispersed flow are selected as the most influential models to the results of interests. Then studies and discussions are specifically focused on these sensitive models and proper modifications are recommended. These proposed improvements are implemented in RELAP5 code and assessed against FEBA experiment. Better agreement between calculations and measured data for both cladding temperature and quench time is obtained.

  5. Cyclical Fluctuations in Workplace Accidents

    OpenAIRE

    Boone, J.; J. C. VAN OURS

    2002-01-01

    This Paper presents a theory and an empirical investigation on cyclical fluctuations in workplace accidents. The theory is based on the idea that reporting an accident dents the reputation of a worker and raises the probability that he is fired. Therefore a country with a high or an increasing unemployment rate has a low (reported) workplace accident rate. The empirical investigation concerns workplace accidents in OECD countries. The analysis confirms that workplace accident rates are invers...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. CRITICA: coding region identification tool invoking comparative analysis

    Science.gov (United States)

    Badger, J. H.; Olsen, G. J.; Woese, C. R. (Principal Investigator)

    1999-01-01

    Gene recognition is essential to understanding existing and future DNA sequence data. CRITICA (Coding Region Identification Tool Invoking Comparative Analysis) is a suite of programs for identifying likely protein-coding sequences in DNA by combining comparative analysis of DNA sequences with more common noncomparative methods. In the comparative component of the analysis, regions of DNA are aligned with related sequences from the DNA databases; if the translation of the aligned sequences has greater amino acid identity than expected for the observed percentage nucleotide identity, this is interpreted as evidence for coding. CRITICA also incorporates noncomparative information derived from the relative frequencies of hexanucleotides in coding frames versus other contexts (i.e., dicodon bias). The dicodon usage information is derived by iterative analysis of the data, such that CRITICA is not dependent on the existence or accuracy of coding sequence annotations in the databases. This independence makes the method particularly well suited for the analysis of novel genomes. CRITICA was tested by analyzing the available Salmonella typhimurium DNA sequences. Its predictions were compared with the DNA sequence annotations and with the predictions of GenMark. CRITICA proved to be more accurate than GenMark, and moreover, many of its predictions that would seem to be errors instead reflect problems in the sequence databases. The source code of CRITICA is freely available by anonymous FTP (rdp.life.uiuc.edu in/pub/critica) and on the World Wide Web (http:/(/)rdpwww.life.uiuc.edu).

  8. CRITICA: coding region identification tool invoking comparative analysis

    Science.gov (United States)

    Badger, J. H.; Olsen, G. J.; Woese, C. R. (Principal Investigator)

    1999-01-01

    Gene recognition is essential to understanding existing and future DNA sequence data. CRITICA (Coding Region Identification Tool Invoking Comparative Analysis) is a suite of programs for identifying likely protein-coding sequences in DNA by combining comparative analysis of DNA sequences with more common noncomparative methods. In the comparative component of the analysis, regions of DNA are aligned with related sequences from the DNA databases; if the translation of the aligned sequences has greater amino acid identity than expected for the observed percentage nucleotide identity, this is interpreted as evidence for coding. CRITICA also incorporates noncomparative information derived from the relative frequencies of hexanucleotides in coding frames versus other contexts (i.e., dicodon bias). The dicodon usage information is derived by iterative analysis of the data, such that CRITICA is not dependent on the existence or accuracy of coding sequence annotations in the databases. This independence makes the method particularly well suited for the analysis of novel genomes. CRITICA was tested by analyzing the available Salmonella typhimurium DNA sequences. Its predictions were compared with the DNA sequence annotations and with the predictions of GenMark. CRITICA proved to be more accurate than GenMark, and moreover, many of its predictions that would seem to be errors instead reflect problems in the sequence databases. The source code of CRITICA is freely available by anonymous FTP (rdp.life.uiuc.edu in/pub/critica) and on the World Wide Web (http:/(/)rdpwww.life.uiuc.edu).

  9. Software and codes for analysis of concentrating solar power technologies.

    Energy Technology Data Exchange (ETDEWEB)

    Ho, Clifford Kuofei

    2008-12-01

    This report presents a review and evaluation of software and codes that have been used to support Sandia National Laboratories concentrating solar power (CSP) program. Additional software packages developed by other institutions and companies that can potentially improve Sandia's analysis capabilities in the CSP program are also evaluated. The software and codes are grouped according to specific CSP technologies: power tower systems, linear concentrator systems, and dish/engine systems. A description of each code is presented with regard to each specific CSP technology, along with details regarding availability, maintenance, and references. A summary of all the codes is then presented with recommendations regarding the use and retention of the codes. A description of probabilistic methods for uncertainty and sensitivity analyses of concentrating solar power technologies is also provided.

  10. Benchmarking Of Improved DPAC Transient Deflagration Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, James E.; Hensel, Steve J.

    2013-03-21

    The transient deflagration code DPAC (Deflagration Pressure Analysis Code) has been upgraded for use in modeling hydrogen deflagration transients. The upgraded code is benchmarked using data from vented hydrogen deflagration tests conducted at the HYDRO-SC Test Facility at the University of Pisa. DPAC originally was written to calculate peak deflagration pressures for deflagrations in radioactive waste storage tanks and process facilities at the Savannah River Site. Upgrades include the addition of a laminar flame speed correlation for hydrogen deflagrations and a mechanistic model for turbulent flame propagation, incorporation of inertial effects during venting, and inclusion of the effect of water vapor condensation on vessel walls. In addition, DPAC has been coupled with CEA, a NASA combustion chemistry code. The deflagration tests are modeled as end-to-end deflagrations. The improved DPAC code successfully predicts both the peak pressures during the deflagration tests and the times at which the pressure peaks.

  11. Analysis of Iterated Hard Decision Decoding of Product Codes with Reed-Solomon Component Codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom

    2007-01-01

    Products of Reed-Solomon codes are important in applications because they offer a combination of large blocks, low decoding complexity, and good performance. A recent result on random graphs can be used to show that with high probability a large number of errors can be corrected by iterating...... minimum distance decoding. We present an analysis related to density evolution which gives the exact asymptotic value of the decoding threshold and also provides a closed form approximation to the distribution of errors in each step of the decoding of finite length codes....

  12. Analysis of a hypothetical loss of coolant accident in a Konvoi type NPP by GASFLOW and COCOSYS

    Energy Technology Data Exchange (ETDEWEB)

    Benz, Stefan; Royl, Peter [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Band, Sebastian [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

    2013-07-01

    The 3D computational fluid dynamics code GASFLOW and the German containment code system COCOSYS, which is based on a lumped-parameter approach, are used to simulate the hydrogen-air-steam distribution and hydrogen mitigation in a Konvoi type nuclear power plant in a postulated hypothetical core melt accident. A break in a coolant loop and the subsequent loss of the coolant causes a strong heat-up of the core. As a consequence hydrogen is produced by oxidation of cladding tubes. The residual steam and the produced hydrogen are released into the containment through the break in the coolant loop. Without suitable counter measures, sensitive mixtures can build up with a combustion potential which could threaten the integrity of the containment. A model of a Konvoi type nuclear power plant which is equipped with passive autocatalytic recombiners is used to simulate such accident scenario. COCOSYS allows comprehensive simulation of all relevant processes of severe accidents, whereas GASFLOW is primarily designed to simulate the distribution of steam and hydrogen within the containment. This paper presents the comparison of GASFLOW and COCOSYS simulation results for the in-vessel phase of the selected accident. (orig.)

  13. Guide to Using Onionskin Analysis Code (U)

    Energy Technology Data Exchange (ETDEWEB)

    Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Statistical Sciences Group; Morzinski, Jerome Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Statistical Sciences Group

    2016-09-15

    This document is a guide to using R-code written for the purpose of analyzing onionskin experiments. We expect the user to be very familiar with statistical methods and the R programming language. For more details about onionskin experiments and the statistical methods mentioned in this document see Storlie, Fugate, et al. (2013). Engineers at LANL experiment with detonators and high explosives to assess performance. The experimental unit, called an onionskin, is a hemisphere consisting of a detonator and a booster pellet surrounded by explosive material. When the detonator explodes, a streak camera mounted above the pole of the hemisphere records when the shock wave arrives at the surface. The output from the camera is a two-dimensional image that is transformed into a curve that shows the arrival time as a function of polar angle. The statistical challenge is to characterize a baseline population of arrival time curves and to compare the baseline curves to curves from a new, so-called, test series. The hope is that the new test series of curves is statistically similar to the baseline population.

  14. Characteristic Analysis of Fire Modeling Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Hwan; Yang, Joon Eon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Jong Hoon [Kyeongmin College, Ujeongbu (Korea, Republic of)

    2004-04-15

    This report documents and compares key features of four zone models: CFAST, COMPBRN IIIE, MAGIC and the Fire Induced Vulnerability Evaluation (FIVE) methodology. CFAST and MAGIC handle multi-compartment, multi-fire problems, using many equations; COMPBRN and FIVE handle single compartment, single fire source problems, using simpler equation. The increased rigor of the formulation of CFAST and MAGIC does not mean that these codes are more accurate in every domain; for instance, the FIVE methodology uses a single zone approximation with a plume/ceiling jet sublayer, while the other models use a two-zone treatment without a plume/ceiling jet sublayer. Comparisons with enclosure fire data indicate that inclusion of plume/ceiling jet sublayer temperatures is more conservative, and generally more accurate than neglecting them. Adding a plume/ceiling jet sublayer to the two-zone models should be relatively straightforward, but it has not been done yet for any of the two-zone models. Such an improvement is in progress for MAGIC.

  15. Fuel handling accident analysis for the University of Missouri Research Reactor's High Enriched Uranium to Low Enriched Uranium fuel conversion initiative

    Science.gov (United States)

    Rickman, Benjamin

    In accordance with the 1986 amendment concerning licenses for research and test reactors, the MU Research Reactor (MURR) is planning to convert from using High-Enriched Uranium (HEU) fuel to the use of Low-Enriched Uranium (LEU) fuel. Since the approval of a new LEU fuel that could meet the MURR's performance demands, the next phase of action for the fuel conversion process is to create a new Safety Analysis Report (SAR) with respect to the LEU fuel. A component of the SAR includes the Maximum Hypothetical Accident (MHA) and accidents that qualify under the class of Fuel Handling Accidents (FHA). In this work, the dose to occupational staff at the MURR is calculated for the FHAs. The radionuclide inventory for the proposed LEU fuel was calculated using the ORIGEN2 point-depletion code linked to the MURR neutron spectrum. The MURR spectrum was generated from a Monte Carlo Neutron transPort (MCNP) simulation. The coupling of these codes create MONTEBURNS, a time-dependent burnup code. The release fraction from each FHA within this analysis was established by the methodology of the 2006 HEU SAR, which was accepted by the NRC. The actual dose methodology was not recorded in the HEU SAR, so a conservative path was chosen. In compliance to NUREG 1537, when new methodology is used in a HEU to LEU analysis, it is necessary to re-evaluate the HEU accident. The Total Effective Dose Equivalent (TEDE) values were calculated in addition to the whole body dose and thyroid dose to operation personnel. The LEU FHA occupational TEDE dose was 349 mrem which is under the NRC regulatory occupational dose limit of 5 rem TEDE, and under the LEU MHA limit of 403 mrem. The re-evaluated HEU FHA occupational TEDE dose was 235 mrem, which is above the HEU MHA TEDE dose of 132 mrem. Since the new methodology produces a dose that is larger than the HEU MHA, we can safely assume that it is more conservative than the previous, unspecified dose.

  16. Analysis of Maximum Reasonably Foreseeable Accidents for the Yucca Mountain Draft Environmental Impact Statement (DEIS)

    Energy Technology Data Exchange (ETDEWEB)

    S.B. Ross; R.E. Best; S.J. Maheras; T.I. McSweeney

    2001-08-17

    Accidents could occur during the transportation of spent nuclear fuel and high-level radioactive waste. This paper describes the risks and consequences to the public from accidents that are highly unlikely but that could have severe consequences. The impact of these accidents would include those to a collective population and to hypothetical maximally exposed individuals (MEIs). This document discusses accidents with conditions that have a chance of occurring more often than 1 in 10 million times in a year, called ''maximum reasonably foreseeable accidents''. Accidents and conditions less likely than this are not considered to be reasonably foreseeable.

  17. IMPROVEMENT OF EXPERT ANALYSIS FOR ROAD TRAFFIC ACCIDENTS USING COMPUTER SIMULATION PROGRAMS

    Directory of Open Access Journals (Sweden)

    S. A. Azemsha

    2015-01-01

    Full Text Available The existing methods for auto-technical expertise presuppose selection of some parameters on the basis of the expert’s intuition and experience. Type of a vehicle and its loading rate, road conditions are not taken into account also in the case when deceleration is to be determined. While carrying out the analysis it has been established that an application of special software makes it possible to improve significantly efficiency of the executed works directed on solution of the assigned tasks, to speed up calculation processes, to decrease qualitatively probability of arithmetic errors and provides the possibility to visualize results of the conducted investigations. Possibility of using various models for dynamic motion simulation and collision of vehicles (in the form of 3D-models has been established in the paper. In such a case specific features of vehicle technical conditions, its loading rate and condition of roadway surface have been taken account in the paper. The given paper also permits to obtain a dynamic display of reconstructed accident mechanism in axonometric projection, to film video-clips when a camera is positioned at any spatial point: road, roadside, raised position, moving vehicle, driver's seat in the vehicle.The paper contains an analysis of possibilities of road traffic accident simulation programs, a statistical analysis that shows significance in differences between simulation results when various programs have been used. The paper presents initial data and results of vehicle speed calculation on the basis of braking track length which have been obtained with the help of road traffic accident express analysis (a classical approach and PC-Crash when additional influencing factors are taken into account. A number of shortcomings have been revealed while analyzing the simulation results of the applied software. The shortcomings must be removed in the analyzed software products.On the basis of the executed analysis in

  18. Development of the severe accident risk information database management system SARD

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

    The main purpose of this report is to introduce essential features and functions of a severe accident risk information management system, SARD (Severe Accident Risk Database Management System) version 1.0, which has been developed in Korea Atomic Energy Research Institute, and database management and data retrieval procedures through the system. The present database management system has powerful capabilities that can store automatically and manage systematically the plant-specific severe accident analysis results for core damage sequences leading to severe accidents, and search intelligently the related severe accident risk information. For that purpose, the present database system mainly takes into account the plant-specific severe accident sequences obtained from the Level 2 Probabilistic Safety Assessments (PSAs), base case analysis results for various severe accident sequences (such as code responses and summary for key-event timings), and related sensitivity analysis results for key input parameters/models employed in the severe accident codes. Accordingly, the present database system can be effectively applied in supporting the Level 2 PSA of similar plants, for fast prediction and intelligent retrieval of the required severe accident risk information for the specific plant whose information was previously stored in the database system, and development of plant-specific severe accident management strategies.

  19. Deterministic Analysis of a Beyond Design Basis Accident in a Low Power, Pin-Type Fuel Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nagah Abdou, Hesham Mohammed [INVAP S. E., Bariloche (Argentina)

    2013-07-01

    A Beyond Design Basis Accident has been analyzed for a pool type research reactor with pin-type, Zry4 clad fuel. This is a low power research reactor (maximum power: 100kW) with neutron beam facilities. Two scenarios are considered: a neutron beam rapture that results in a fraction of the core submerged in water and a catastrophic failure that results in a fully uncovered core. The paper discusses the different cooling mechanisms for these two BDBAs and compares results for both scenarios, with predictions of no core damage in any situation. Core damage is defined as CHFR↔1.5 and/or Tclad→T start of breakaway oxidation temperature. In addition, the paper compares calculations with a thermalhydraulic code and an analytical model. This paper allows to analyze the applicability of regular thermalhydraulic codes to BDBA accident scenarios in low power research reactors.

  20. Syntactic Accidents in Program Analysis: On the Impact of the CPS Transformation

    DEFF Research Database (Denmark)

    Daniel, Damian; Danvy, Olivier

    2000-01-01

    control-flow analysis yields equivalent results on a direct-style program and on its CPS counterpart, a monovariant binding-time analysis yields less precise results on a direct-style program than on its CPS counterpart, and an enhanced monovariant binding-time analysis yields equivalent results...... on a direct-style program and on its CPS counterpart. Our proof technique amounts to constructing the CPS counterpart of flow information and of binding times. Our results formalize and confirm a folklore theorem about traditional binding-time analysis, namely that CPS has a positive effect on binding times....... What may be more surprising is that the benefit does not arise from a standard refinement of program analysis, as, for instance, duplicating continuations. The present study is symptomatic of an unsettling property of program analyses: their quality is unpredictably vulnerable to syntactic accidents...

  1. Visualization of Traffic Accidents

    Science.gov (United States)

    Wang, Jie; Shen, Yuzhong; Khattak, Asad

    2010-01-01

    Traffic accidents have tremendous impact on society. Annually approximately 6.4 million vehicle accidents are reported by police in the US and nearly half of them result in catastrophic injuries. Visualizations of traffic accidents using geographic information systems (GIS) greatly facilitate handling and analysis of traffic accidents in many aspects. Environmental Systems Research Institute (ESRI), Inc. is the world leader in GIS research and development. ArcGIS, a software package developed by ESRI, has the capabilities to display events associated with a road network, such as accident locations, and pavement quality. But when event locations related to a road network are processed, the existing algorithm used by ArcGIS does not utilize all the information related to the routes of the road network and produces erroneous visualization results of event locations. This software bug causes serious problems for applications in which accurate location information is critical for emergency responses, such as traffic accidents. This paper aims to address this problem and proposes an improved method that utilizes all relevant information of traffic accidents, namely, route number, direction, and mile post, and extracts correct event locations for accurate traffic accident visualization and analysis. The proposed method generates a new shape file for traffic accidents and displays them on top of the existing road network in ArcGIS. Visualization of traffic accidents along Hampton Roads Bridge Tunnel is included to demonstrate the effectiveness of the proposed method.

  2. HTGR accident initiation and progression analysis status report. Volume V. AIPA fission product source terms

    Energy Technology Data Exchange (ETDEWEB)

    Alberstein, D.; Apperson, C.E. Jr.; Hanson, D.L.; Myers, B.F.; Pfeiffer, W.W.

    1976-02-01

    The primary objective of the Accident Initiation and Progression Analysis (AIPA) Program is to provide guidance for high-temperature gas-cooled reactor (HTGR) safety research and development. Among the parameters considered in estimating the uncertainties in site boundary doses are uncertainties in fission product source terms generated under normal operating conditions, i.e., fuel body inventories, circulating coolant activity, total plateout activity in the primary circuit, and plateout distributions. The volume presented documents the analyses of these source term uncertainties. The results are used for the detailed consequence evaluations, and they provide the basis for evaluation of fission products important for HTGR maintenance and shielding.

  3. Code conversion for system design and safety analysis of NSSS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hae Cho; Kim, Young Tae; Choi, Young Gil; Kim, Hee Kyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    This report describes overall project works related to conversion, installation and validation of computer codes which are used in NSSS design and safety analysis of nuclear power plants. Domain/os computer codes for system safety analysis are installed and validated on Apollo DN10000, and then Apollo version are converted and installed again on HP9000/700 series with appropriate validation. Also, COOLII and COAST which are cyber version computer codes are converted into versions of Apollo DN10000 and HP9000/700, and installed with validation. This report details whole processes of work involved in the computer code conversion and installation, as well as software verification and validation results which are attached to this report. 12 refs., 8 figs. (author)

  4. Development of computer code models for analysis of subassembly voiding in the LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Hinkle, W [ed.

    1979-12-01

    The research program discussed in this report was started in FY1979 under the combined sponsorship of the US Department of Energy (DOE), General Electric (GE) and Hanford Engineering Development Laboratory (HEDL). The objective of the program is to develop multi-dimensional computer codes which can be used for the analysis of subassembly voiding incoherence under postulated accident conditions in the LMFBR. Two codes are being developed in parallel. The first will use a two fluid (6 equation) model which is more difficult to develop but has the potential for providing a code with the utmost in flexibility and physical consistency for use in the long term. The other will use a mixture (< 6 equation) model which is less general but may be more amenable to interpretation and use of experimental data and therefore, easier to develop for use in the near term. To assure that the models developed are not design dependent, geometries and transient conditions typical of both foreign and US designs are being considered.

  5. Analysis of credible accidents for Argonaut reactors. Report for October 1980-April 1981

    Energy Technology Data Exchange (ETDEWEB)

    Hawley, S.C.; Kathren, R.L.; Robkin, M.A.

    1981-04-01

    Five areas of potential accidents have been evaluated for the Argonaut-UTR reactors. They are: insertion of excess reactivity, catastrophic rearrangement of the core, explosive chemical reaction, graphite fire, and a fuel-handling accident.

  6. Determinants of within-country variation in traffic accident mortality in Italy: A geographical analysis

    NARCIS (Netherlands)

    G. La Torre (Giuseppe); E.F. van Beeck (Ed); G. Quaranta (Gianluigi); A. Mannocci (Alice); W. Ricciardi

    2007-01-01

    textabstractObjective: To identify determinants of regional differences in traffic accident mortality in Italy. Data and methods: Multiple linear regression models were conducted assessing the associations between regional differences in traffic mortality, case fatality and accident rates (dependent

  7. AP1000给水丧失事故定性分析%Qualitative Accident Analysis on Loss of Normal Feedwater for AP1000

    Institute of Scientific and Technical Information of China (English)

    李延凯; 林萌; 侯东; 李美琳; 杨燕华

    2012-01-01

    针对AP1000核电站,基于两流体最佳估算系统程序RELAP5建立热工水力模型,基于Matlab/Simulink软件及工业组态软件建立相关控制系统数学模型,用于对正常给水丧失事故的计算分析.建模数据主要参考AP1000 Design Control Document(AP1000 DCD),由于建模数据不够充分、详尽,模型不够精确,文中事故分析以定性分析为主.计算结果表明:RELAP5具备计算自然循环的能力,计算结果与DCD中正常给水丧失事故结果总体趋势基本一致,非能动余热排出系统(PRHRS)、堆芯补水箱(CMT)系统能够及时、有效地排出堆芯余热和堆芯衰变热,确保堆芯安全.PRHRS余热排出能力对事故发展有明显影响,模型中PRHRS余热排出能力较强,使冷却剂温度更快地降低到较低水平,导致CMT更早投入以及随后反应堆各参数响应的不同.%For the analysis of loss of normal feedwater accident for AP1000, a thermal-hydraulic model was built based on a two-fluid best estimate code, RELAP5. A control system model was built based on Matlab/Simulink code. As the main data for modeling was from AP1000 Design Control Document (AP1000 DCD), which was inadequate and not accurate enough, the accident analysis was qualitative. The calculation results, overall consistent with the same accident in DCD, show that the RELAP5 code has the ability to calculate the natural circulation. PRHRS and CMT are able to remove the core residual heat, decay heat timely and efficiently during the accident to guarantee the safety of the core. In the calculation model, the capability of PRHRS is stronger, which makes the coolant temperature quickly reduced to a lower level, leading to an earlier actuation of CMT, followed by different responses of the plant.

  8. A systemic approach to accident analysis: a case study of the Stockwell shooting.

    Science.gov (United States)

    Jenkins, Daniel P; Salmon, Paul M; Stanton, Neville A; Walker, Guy H

    2010-01-01

    This paper uses a systemic approach to accident investigation, based upon AcciMaps, to model the events leading up to the shooting of Jean Charles de Menezes at Stockwell Underground station in July 2005. The model captures many of the findings of the Independent Police Complaints Commission's report in a single representation, modelling their interdependencies and the causal flow. Furthermore, by taking a systemic approach, the analysis identifies further considerations related to the suitability of the Metropolitan Police Service's organisational structure to support rapid-paced operations, where reliable identification of a suspect is not possible. Based upon the analysis, the paper questions the division of functions between teams and the suitability of an organisational structure that relies upon the complex flow of information between separate teams for surveillance and for controlling the suspect. A dynamic organisational structure is proposed that changes in response to operation type and unfolding events. STATEMENT OF RELEVANCE: This paper provides much needed and called for validation for a systemic approach to accident analysis. A widely reported case study is used to illustrate the process. The paper shows how such an approach can consolidate the key findings of much larger reports as well as draw out additional recommendations.

  9. Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

    Directory of Open Access Journals (Sweden)

    Eltayeb Yousif

    2017-01-01

    Full Text Available Many reactor safety simulation codes for nuclear power plants (NPPs have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR. RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

  10. Risk-based Analysis of Construction Accidents in Iran During 2007-2011-Meta Analyze Study

    OpenAIRE

    Mehran Amiri; Abdollah Ardeshir; Mohammad Hossein Fazel Zarandi

    2014-01-01

    Abstract Background The present study aimed to investigate the characteristics of occupational accidents and frequency and severity of work related accidents in the construction industry among Iranian insured workers during the years 20072011. Methods The Iranian Social Security Organization (ISSO) accident database containing 21,864 cases between the years 2007-2011 was applied in this study. In the next step, Total Accident Rate (TRA), Total Severity Index (TSI), and Risk Factor (RF) were d...

  11. Effect of In-Vessel Retention Strategies under Postulated SGTR Accidents of OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Wonjun; Lee, Yongjae; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of); Kim, Hwan-Yeol; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, MELCOR code was used to simulate the severe accident of the OPR1000. MELCOR code is computer code which enables to simulate the progression of the severe accident for light water reactors. It has been developed by Sandia National Laboratories for plant risk assessment and source term analysis since 1982. According to the probabilistic safety analysis (PSA) Level 1 of OPR1000, typical severe accident scenarios of high probability of a transition to severe accident for OPR1000 were identified as Small Break Loss of Coolant Accident (SBLOCA), Station Black out (SBO), Total Loss of Feed Water (TLOFW), and Steam Generator Tube Rupture. While the first three accidents are expected to result in the generation and transportation of the radioactive nuclides within the containment building as consequence of the core damage and subsequent reactor pressure vessel (RPV) failure, the latter accident scenario may be progressed with possible direct release of the radioactive nuclides to the environment by bypassing the containment building. Thus it is of significance to investigate the SGTR accident with a sophisticated severe accident code. This code can simulate the whole phenomena of a severe accident such as thermal-hydraulic response, core heat-up, oxidation and relocation, and fission product release and transport. Thus many researchers have used MELCOR in severe accident studies. In this study, in-vessel retention strategies were applied for postulated SGTR accidents. Mitigation effect and adverse effect of in-vessel strategies was studied in aspect of RPV failure, fission product release and containment thermal-hydraulic and hydrogen behavior. Base case of SGTR accident and three mitigation cases were simulated using MELCOR code 1.8.6. For each mitigation cases, mitigation effect and adverse effect were investigated. Conclusions can be summarized as follows: (1) RPV failure of SGTR base case occurred at 5.62 hours and fission product of RCS released to

  12. A Content Analysis of News Media Coverage of the Accident at Three Mile Island.

    Science.gov (United States)

    Stephens, Mitchell; Edison, Nadyne G.

    A study was conducted for the President's Commission on the Accident at Three Mile Island to analyze coverage of the accident by ten news organizations: two wire services, three commercial television networks, and five daily newspapers. Copies of all stories and transcripts of news programs during the first week of the accident were examined from…

  13. Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP

    Science.gov (United States)

    Maruyama, Soh; Fujimoto, Nozomu; Kiso, Yoshihiro; Murakami, Tomoyuki; Sudo, Yukio

    1988-09-01

    This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T(sub 1-M)) with simulated fuel rods and fuel blocks.

  14. Narrative text analysis of accident reports with tractors, self-propelled harvesting machinery and materials handling machinery in Austrian agriculture from 2008 to 2010 - a comparison.

    Science.gov (United States)

    Mayrhofer, Hannes; Quendler, Elisabeth; Boxberger, Josef

    2014-01-01

    The aim of this study was the identification of accident scenarios and causes by analysing existing accident reports of recognized agricultural occupational accidents with tractors, self-propelled harvesting machinery and materials handling machinery from 2008 to 2010. As a result of a literature-based evaluation of past accident analyses, the narrative text analysis was chosen as an appropriate method. A narrative analysis of the text fields of accident reports that farmers used to report accidents to insurers was conducted to obtain detailed information about the scenarios and causes of accidents. This narrative analysis of reports was made the first time and yielded first insights for identifying antecedents of accidents and potential opportunities for technical based intervention. A literature and internet search was done to discuss and confirm the findings. The narrative text analysis showed that in more than one third of the accidents with tractors and materials handling machinery the vehicle rolled or tipped over. The most relevant accident scenarios with harvesting machinery were being trapped and falling down. The direct comparison of the analysed machinery categories showed that more than 10% of the accidents in each category were caused by technical faults, slippery or muddy terrain and incorrect or inappropriate operation of the vehicle. Accidents with tractors, harvesting machinery and materials handling machinery showed similarities in terms of causes, circumstances and consequences. Certain technical and communicative measures for accident prevention could be used for all three machinery categories. Nevertheless, some individual solutions for accident prevention, which suit each specific machine type, would be necessary.

  15. Analysis of Public Perception on Radiation: with One Year after Fukushima Nuclear Accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Bang Ju [Korean Science Reporters Association, Seoul (Korea, Republic of)

    2012-03-15

    A year has passed since the nuclear power plant accident in Fukushima on March 11, 2011, and a survey for public perception on radiation by Korean people has been made. The methodological design was based on a quantitative survey and a frequency analysis was done. The analysis objects were survey papers (n=2,754pcs) answered by random ordinary citizens chosen from all over the country. The questionnaires, and study tool, were directly distributed and collected. A total of 40 questionnaires using a coefficient of Cronbach's {alpha} per each area was 'self perception of radiation' (0.620), 'radiation risk' (0.830), 'benefit from radiation' (0.781), 'radiation controlled' (0.685), 'informative source of radiation' (0.831), 'influence degree from Fukushima accident' (0.763), showing rather high score from all areas. As the result of the questionnaires, the knowledge of radiation concept was 69.50 out of 100 points, which shows a rather significant difference from the result of 'know well about radiation' (53.7%) and 'just know about radiation' (37.40%). According to the survey, one of the main reasons why radiation seems risky was that once exposed to radiation, it may not have negative impacts presently but, the next generation could see negative impacts (66.1%). About 41% of our respondents showed a negative position against the government's report on radiation while 39.5% of respondents said that we should stop running nuclear power in light of Fukushima nuclear power plant accident. This study was done for the first time by Korean people's public perception on radiation after the Fukushima nuclear power plant accident. We expect this might have significant contributions to the establishment of the government's policy on radiation.

  16. Injuries to New Zealanders participating in adventure tourism and adventure sports: an analysis of Accident Compensation Corporation (ACC) claims.

    Science.gov (United States)

    Bentley, Tim; Macky, Keith; Edwards, Jo

    2006-12-15

    The aim of this study was to examine the involvement of adventure tourism and adventure sports activity in injury claims made to the Accident Compensation Corporation (ACC). Epidemiological analysis of ACC claims for the period, July 2004 to June 2005, where adventure activities were involved in the injury. 18,697 adventure tourism and adventure sports injury claims were identified from the data, representing 28 activity sectors. Injuries were most common during the summer months, and were most frequently located in the major population centres. The majority of injuries were incurred by claimants in the 20-50 years age groups, although claimants over 50 years of age had highest claims costs. Males incurred 60% of all claims. Four activities (horse riding, mountain biking, tramping/hiking, and surfing) were responsible for approximately 60% of all adventure tourism and adventure sports-related injuries. Slips, trips, and falls were the most common injury initiating events, and injuries were most often to the back/spine, shoulder, and knee. These findings suggest the need to investigate whether regulatory intervention in the form of codes of practice for high injury count activities such as horse riding and mountain biking may be necessary. Health promotion messages and education programs should focus on these and other high-injury risk areas. Improved risk management practices are required for commercial adventure tourism and adventure sports operators in New Zealand if safety is to be improved across this sector.

  17. Recent applications of the transonic wing analysis computer code, TWING

    Science.gov (United States)

    Subramanian, N. R.; Holst, T. L.; Thomas, S. D.

    1982-01-01

    An evaluation of the transonic-wing-analysis computer code TWING is given. TWING utilizes a fully implicit approximate factorization iteration scheme to solve the full potential equation in conservative form. A numerical elliptic-solver grid-generation scheme is used to generate the required finite-difference mesh. Several wing configurations were analyzed, and the limits of applicability of this code was evaluated. Comparisons of computed results were made with available experimental data. Results indicate that the code is robust, accurate (when significant viscous effects are not present), and efficient. TWING generally produces solutions an order of magnitude faster than other conservative full potential codes using successive-line overrelaxation. The present method is applicable to a wide range of isolated wing configurations including high-aspect-ratio transport wings and low-aspect-ratio, high-sweep, fighter configurations.

  18. QR Codes in the Library: Are They Worth the Effort? Analysis of a QR Code Pilot Project

    OpenAIRE

    Wilson, Andrew M

    2012-01-01

    The literature is filled with potential uses for Quick Response (QR) codes in the library. Setting, but few library QR code projects have publicized usage statistics. A pilot project carried out in the Eda Kuhn Loeb Music Library of the Harvard College Library sought to determine whether library patrons actually understand and use QR codes. Results and analysis of the pilot project are provided, attempting to answer the question as to whether QR codes are worth the effort for libraries.

  19. Severe accident risks from external events

    Institute of Scientific and Technical Information of China (English)

    Randall O Gauntt

    2013-01-01

    This paper reviews the early development of design requirements for seismic events in USA early developing nuclear electric generating fleet.Notable safety studies,including WASH-1400,Sandia Siting Study and the NUREG-1150 probabilistic risk study,are briefly reviewed in terms of their relevance to extreme accidents arising from seismic and other severe accident initiators.Specific characteristic about the nature of severe accidents in nuclear power plant (NPP) are reviewed along with present day state-of-art analysis methodologies (methods for estimation of leakages and consequences of releases (MELCOR) and MELCOR accident consequence code system (MACCS)) that are used to evaluate severe accidents and to optimize mitigative and protective actions against such accidents.It is the aim of this paper to make nuclear operating nations aware of the risks that accompany a much needed energy resource and to identify some of the tools,techniques and landmark safety studies that serve to make the technology safer and to maintain vigilance and adequate safety culture for the responsible management of this valuable but unforgiving technology.

  20. Analysis of LAPAN-IPB image lossless compression using differential pulse code modulation and huffman coding

    Science.gov (United States)

    Hakim, P. R.; Permala, R.

    2017-01-01

    LAPAN-A3/IPB satellite is the latest Indonesian experimental microsatellite with remote sensing and earth surveillance missions. The satellite has three optical payloads, which are multispectral push-broom imager, digital matrix camera and video camera. To increase data transmission efficiency, the multispectral imager data can be compressed using either lossy or lossless compression method. This paper aims to analyze Differential Pulse Code Modulation (DPCM) method and Huffman coding that are used in LAPAN-IPB satellite image lossless compression. Based on several simulation and analysis that have been done, current LAPAN-IPB lossless compression algorithm has moderate performance. There are several aspects that can be improved from current configuration, which are the type of DPCM code used, the type of Huffman entropy-coding scheme, and the use of sub-image compression method. The key result of this research shows that at least two neighboring pixels should be used for DPCM calculation to increase compression performance. Meanwhile, varying Huffman tables with sub-image approach could also increase the performance if on-board computer can support for more complicated algorithm. These results can be used as references in designing Payload Data Handling System (PDHS) for an upcoming LAPAN-A4 satellite.

  1. EVAPRED - A CODE FOR FATIGUE ANALYSIS OPTIMIZATION

    Directory of Open Access Journals (Sweden)

    Dorin LOZICI-BRÎNZEI

    2010-03-01

    Full Text Available The fatigue can be, in fact, defined as: “failure under a repeated or otherwise varying load, which never reaches a level sufficient to cause failure in a single application”.Physical testing is clearly unrealistic for every design component. In most applications, fatigue-safe life design requires the prediction of the component fatigue life that accounts for predicted service loads and materials. The primary tool for both understanding and being able to predict and avoid fatigue has proven to be the finite element analysis (FEA. Computer-aided engineering (CAE programs use three major methods to determine the total fatigue life: Stress life (SN, Strain life (EN and Fracture Mechanics (FM. FEA can predict stress concentration areas and can help design engineers to predict how long their designs are likely to last before experiencing the onset of fatigue.

  2. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments.

  3. COMPARATIVE ANALYSIS OF SOME HEALTH INDICATORS OF VARIOUS RADIATION ACCIDENTS LIQUIDATORS

    Directory of Open Access Journals (Sweden)

    V. M. Shubik

    2010-01-01

    Full Text Available The article presents the results of comparative investigation of morbidity and immunity of liquidators of radiation accidents occurred in South Urals (Kyshtym accident, at Chernobyl NPP and nuclear submarines (NS consequences. The most evident immunity and health changes were revealed for liquidators of Chernobyl NPP accident (ChNPP. Investigations of Kyshtym accident liquidators revealed long-term immunological losses. Comparison of health indicators of Chernobyl and nuclear submarine accident liquidators reveals the possibility of combined influence of radiation and stress on the immunity and health.

  4. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  5. Safety culture and accident analysis-A socio-management approach based on organizational safety social capital

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Suman [Risk Analyst (India)]. E-mail: sumanashokrao@yahoo.co.in

    2007-04-11

    One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, Key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization-seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly.

  6. StarFinder: A code for stellar field analysis

    Science.gov (United States)

    Diolaiti, Emiliano; Bendinelli, Orazio; Bonaccini, Domenico; Close, Laird M.; Currie, Doug G.; Parmeggiani, Gianluigi

    2000-11-01

    StarFinder is an IDL code for the deep analysis of stellar fields, designed for Adaptive Optics well-sampled images with high and low Strehl ratio. The Point Spread Function is extracted directly from the frame, to take into account the actual structure of the instrumental response and the atmospheric effects. The code is written in IDL language and organized in the form of a self-contained widget-based application, provided with a series of tools for data visualization and analysis. A description of the method and some applications to Adaptive Optics data are presented.

  7. Analysis of isoplanatic high resolution stellar fields by Starfinder code

    CERN Document Server

    Diolaiti, E; Bonaccini, D; Close, L M; Currie, D; Parmeggiani, G

    2000-01-01

    We describe a new code for the deep analysis of stellar fields, designed for Adaptive Optics Nyquist-sampled images with high and low Strehl ratio. The Point Spread Function is extracted directly from the image frame, to take into account the actual structure of the instrumental response and the atmospheric effects. The code is written in IDL language and organized in the form of a self-contained widget-based application, provided with a series of tools for data visualization and analysis. A description of the method and some applications to AO data are presented.

  8. Of Disasters and Dragon Kings: A Statistical Analysis of Nuclear Power Incidents & Accidents

    CERN Document Server

    Wheatley, Spencer; Sornette, Didier

    2015-01-01

    We provide, and perform a risk theoretic statistical analysis of, a dataset that is 75 percent larger than the previous best dataset on nuclear incidents and accidents, comparing three measures of severity: INES (International Nuclear Event Scale), radiation released, and damage dollar losses. The annual rate of nuclear accidents, with size above 20 Million US$, per plant, decreased from the 1950s until dropping significantly after Chernobyl (April, 1986). The rate is now roughly stable at 0.002 to 0.003, i.e., around 1 event per year across the current fleet. The distribution of damage values changed after Three Mile Island (TMI; March, 1979), where moderate damages were suppressed but the tail became very heavy, being described by a Pareto distribution with tail index 0.55. Further, there is a runaway disaster regime, associated with the "dragon-king" phenomenon, amplifying the risk of extreme damage. In fact, the damage of the largest event (Fukushima; March, 2011) is equal to 60 percent of the total damag...

  9. Accident Analysis and Countermeasures of the Enterprises Involved in Ammonia%涉氨企业事故分析与对策

    Institute of Scientific and Technical Information of China (English)

    卢均臣; 王延平

    2015-01-01

    从事故发生环节、事故类型、事故设备、事故原因等几个方面分析了2005年-2014年全国涉氨企业发生的事故。分析表明:储存和使用环节事故最多;事故主要类型事故泄漏和中毒;主要发生在食品厂、肉类加工厂、冷饮厂、水产公司、果蔬公司、制药厂的制冷车间;主要发生在管道、储罐、阀门、法兰等部位;材料失效事故占比最高。最后,提出了预防此类事故的建议措施。%Accidents occurred in 2005-2014 in our country were analyzed from the aspects of the accident link, accident type, accident equipment, accident cause. Analysis showed that accidents in links of the storage and use were the most, the main types of accidents were leakage and poisoning, accidents mainly occured in food factory, meat processing factory, beverage factory, etc, accidents mainly occured in pipeline, storage tanks, valves, flanges, etc, the amount of accidents caused by material failure was the largest. Suggestions for preventing such accidents were put forward.

  10. Diversity Analysis of Bit-Interleaved Coded Multiple Beamforming

    CERN Document Server

    Park, Hong Ju

    2008-01-01

    In this paper, diversity analysis of bit-interleaved coded multiple beamforming (BICMB) is extended to the case of general spatial interleavers, removing a condition on their previously known design criteria and quantifying the resulting diversity order. The diversity order is determined by a parameter Qmax which is inherited from the convolutional code and the spatial de-multiplexer used in BICMB. We introduce a method to find this parameter by employing a transfer function approach as in finding the weight spectrum of a convolutional code. By using this method, several Qmax's are shown and verified to be identical with the results from a computer searching program tracing paths on the trellis. The diversity analysis and the method to find the parameter are supported by simulation results.

  11. Development of a subchannel analysis code MATRA (Ver. {alpha})

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Y. J.; Hwang, D. H

    1998-04-01

    A subchannel analysis code MATRA-{alpha}, an interim version of MATRA, has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-IV-I. This MATRA code is a thermal-hydraulic analysis code based on the subchannel approach for calculating the enthalpy and flow distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. MATRA-{alpha} has been provided with an improved structure, various functions, and models to give the more convenient user environment and to increase the code accuracy, various functions, and models to give the more convenient user environment and to increase the code accuracy. Among them, the pressure drop model has been improved to be applied to non-square-lattice rod arrays, and the lateral transport models between adjacent subchannels have been improved to increase the accuracy in predicting two-phase flow phenomena. Also included in this report are the detailed instructions for input data preparation and for auxiliary pre-processors to serve as a guide to those who want to use MATRA-{alpha}. In addition, we compared the predictions of MATRA-{alpha} with the experimental data on the flow and enthalpy distribution in three sample rod-bundle cases to evaluate the performance of MATRA-{alpha}. All the results revealed that the prediction of MATRA-{alpha} were better than those of COBRA-IV-I. (author). 16 refs., 1 tab., 13 figs.

  12. Analysis of SMA Hybrid Composite Structures using Commercial Codes

    Science.gov (United States)

    Turner, Travis L.; Patel, Hemant D.

    2004-01-01

    A thermomechanical model for shape memory alloy (SMA) actuators and SMA hybrid composite (SMAHC) structures has been recently implemented in the commercial finite element codes MSC.Nastran and ABAQUS. The model may be easily implemented in any code that has the capability for analysis of laminated composite structures with temperature dependent material properties. The model is also relatively easy to use and requires input of only fundamental engineering properties. A brief description of the model is presented, followed by discussion of implementation and usage in the commercial codes. Results are presented from static and dynamic analysis of SMAHC beams of two types; a beam clamped at each end and a cantilevered beam. Nonlinear static (post-buckling) and random response analyses are demonstrated for the first specimen. Static deflection (shape) control is demonstrated for the cantilevered beam. Approaches for modeling SMAHC material systems with embedded SMA in ribbon and small round wire product forms are demonstrated and compared. The results from the commercial codes are compared to those from a research code as validation of the commercial implementations; excellent correlation is achieved in all cases.

  13. The Analysis of the Contribution of Human Factors to the In-Flight Loss of Control Accidents

    Science.gov (United States)

    Ancel, Ersin; Shih, Ann T.

    2012-01-01

    In-flight loss of control (LOC) is currently the leading cause of fatal accidents based on various commercial aircraft accident statistics. As the Next Generation Air Transportation System (NextGen) emerges, new contributing factors leading to LOC are anticipated. The NASA Aviation Safety Program (AvSP), along with other aviation agencies and communities are actively developing safety products to mitigate the LOC risk. This paper discusses the approach used to construct a generic integrated LOC accident framework (LOCAF) model based on a detailed review of LOC accidents over the past two decades. The LOCAF model is comprised of causal factors from the domain of human factors, aircraft system component failures, and atmospheric environment. The multiple interdependent causal factors are expressed in an Object-Oriented Bayesian belief network. In addition to predicting the likelihood of LOC accident occurrence, the system-level integrated LOCAF model is able to evaluate the impact of new safety technology products developed in AvSP. This provides valuable information to decision makers in strategizing NASA's aviation safety technology portfolio. The focus of this paper is on the analysis of human causal factors in the model, including the contributions from flight crew and maintenance workers. The Human Factors Analysis and Classification System (HFACS) taxonomy was used to develop human related causal factors. The preliminary results from the baseline LOCAF model are also presented.

  14. Accident Analysis of Chinese CPR1000 in Response to Station Blackout

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Juyoul [FNC Technology Co., Yongin (Korea, Republic of); Cilliers, Anthonie [North-West University, Potchefstroom (South Africa)

    2016-10-15

    Stress tests required evaluation of the consequences of loss of safety functions from any initiating event (e.g., earthquake or flooding) causing loss of electrical power, including station blackout (SBO). The SBO scenario involves a loss of offsite power, failure of the redundant emergency diesel generators, failure of alternate current (AC) power restoration and the eventual degradation of the reactor coolant pump (RCP) seals resulting in a long term loss of coolant. Using PCTRAN/CPR1000, this study analyses the station blackout on a Chinese CPR1000 which is the most representative type reactor in terms of number of reactors, operating period, power capacity and geological distance from Korean Peninsula. Both the physical effects of the accidents as well as the releases of radioisotopes are calculated and discussed. Station blackout simulation was conducted in this study. The resulting effects seen are consistent with other stress test station blackout tests used utilizing licensed simulation codes. An exact comparison is however not possible as the plants on which the simulations was done vary greatly and the limitations of availability to Chinese FSAR. PCTRAN/CPR1000 is an extremely useful simulation package that provides engineers and scientists very accurate feedback to how a nuclear power plant would react as a whole under various plant conditions. It is able to do this extremely fast as well. As a training tool PCTRAN/CPR1000 provides hands-on experience with many of the primary plant operations and develops an intuitive understanding of the plant.

  15. An analysis of Japan radiation protection measurements after the Fukushima nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Arlene A. dos; Souza-Santos, Denison; Melo, Dunstana R. de; Hunt, John G.; Juliao, Ligia M.Q C.; Conti, Luiz F.C.; Pires do Rio, Monica A.; Reis, Rocio G., E-mail: arlene@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    On March 11th 2011, Japan was struck by a devastating earthquake followed by a tsunami wave that took the lives of thousands and started a major nuclear accident in the Fukushima Dai-ichi power plant complex. Right from the beginning, the information published by the Japanese government and by the International Atomic Energy Agency (IAEA) was followed by a team of experts at the Institute for Radiation Protection and Dosimetry (IRD) of Brazil. Radiation monitoring data, such as radionuclide activity concentration in water and food, ambient dose rate and fallout concentration in specific cities have been compiled and analyzed, with emphasis on dose limits established by Brazilian regulatory authority. A computer code for dose assessment, developed at the IRD and based upon the IAEA documents TECDOC-1162 and TECDOC-955, was used to assess the doses due to intakes of radionuclides and external exposure for individuals of different age groups. The IAEA model predictions for the ambient dose rates, when the fallout is known, are compared with the measured values in different cities. The Japanese recommendations for evacuation, sheltering and restriction of food and water consumption are evaluated with regards to the Brazilian limits defined in the CNEN NN 3.01 standard. (author)

  16. 76 FR 56413 - Building Energy Codes Cost Analysis

    Science.gov (United States)

    2011-09-13

    ... analysis using an appropriate building energy estimation tool. DOE intends to use the EnergyPlus \\3...-revision codes, two prototype buildings would be analyzed--one that exactly complies with the pre-revision... in the following sections. \\4\\ ``Exactly complies'' means that the prototype complies with...

  17. Course and Research Analysis Using a Coded Classification System.

    Science.gov (United States)

    Lochstet, Gwenn S.

    1997-01-01

    A system of course analysis was developed and used to code and compare faculty research, courses, and library materials in the Mathematics, Physics, and Statistics departments of the University of South Carolina. The purpose is to provide a guide in determining how well the library's collection supports the academic needs of these departments. (10…

  18. Who by accident? The social morphology of car accidents.

    Science.gov (United States)

    Factor, Roni; Yair, Gad; Mahalel, David

    2010-09-01

    Prior studies in the sociology of accidents have shown that different social groups have different rates of accident involvement. This study extends those studies by implementing Bourdieu's relational perspective of social space to systematically explore the homology between drivers' social characteristics and their involvement in specific types of motor vehicle accident. Using a large database that merges official Israeli road-accident records with socioeconomic data from two censuses, this research maps the social order of road accidents through multiple correspondence analysis. Extending prior studies, the results show that different social groups indeed tend to be involved in motor vehicle accidents of different types and severity. For example, we find that drivers from low socioeconomic backgrounds are overinvolved in severe accidents with fatal outcomes. The new findings reported here shed light on the social regularity of road accidents and expose new facets in the social organization of death. © 2010 Society for Risk Analysis.

  19. FARO and KROTOS code simulation and analysis at JRC Ispra

    Energy Technology Data Exchange (ETDEWEB)

    Annunziato, A.; Yerkess, A.; Addabbo, C. [European Commission-Joint Research Centre, Inst. for Systems, Informatics and Safety, 21020 Ispra (Italy)

    1998-01-01

    The paper summarizes relevant results from the pre and post test calculations of fuel coolant interaction and quenching tests performed in the FARO and KROTOS test facilities. The main analytical tools adopted at JRC Ispra are the COMETA and the TEXAS codes. COMETA pre and post test calculations of FARO Test L-20 as well as an application of the code to KROTOS test facility are presented. The analysis provides the need to account for H{sub 2} generation models into the pre-mixing calculations. In addition salient results from the application of TEXAS to FARO and KROTOS tests are shown. (author)

  20. Analysis of traffic accident size for Korean highway using structural equation models.

    Science.gov (United States)

    Lee, Ju-Yeon; Chung, Jin-Hyuk; Son, Bongsoo

    2008-11-01

    Accident size can be expressed as the number of involved vehicles, the number of damaged vehicles, the number of deaths and/or the number of injured. Accident size is the one of the important indices to measure the level of safety of transportation facilities. Factors such as road geometric condition, driver characteristic and vehicle type may be related to traffic accident size. However, all these factors interact in complicate ways so that the interrelationships among the variables are not easily identified. A structural equation model is adopted to capture the complex relationships among variables because the model can handle complex relationships among endogenous and exogenous variables simultaneously and furthermore it can include latent variables in the model. In this study, we use 2649 accident data occurred on highways in Korea and estimate relationship among exogenous factors and traffic accident size. The model suggests that road factors, driver factors and environment factors are strongly related to the accident size.

  1. Starfinder a code for crowded stellar fields analysis

    CERN Document Server

    Diolaiti, E; Bonaccini, D; Close, L M; Currie, D; Parmeggiani, G

    1999-01-01

    Starfinder is an IDL code for the deep analysis of stellar fields, designed for well-sampled images with high and low Strehl factor. An important feature is represented by the possibility to measure the anisoplanatic effect in wide-field Adaptive Optics observations and exploit this knowledge to improve the analysis of the observed field. A description of the method and applications to real AO data are presented.

  2. Statistical Analysis and Safety Management on China's Coal Mine Gas Accident from 2006 to 2015

    Directory of Open Access Journals (Sweden)

    ZHOU Xuan

    2016-03-01

    Full Text Available . From 2006 to 2015 the causes and characteristics of coal mine gas accidents are analyzed from the aspect of the death toll and number of accidents in China based on the theory of statistics. And the results show that the level of gas accident casualties is obviously regional distribution and points and time distribution is relatively concentrated as well. Therefore, the corresponding management and measures are put forward based on Orbit Intersecting theory.

  3. A Human Reliability Analysis of Post- Accident Human Errors in the Low Power and Shutdown PSA of KSNP

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Daeil; Kim, J. H.; Jang, S. C

    2007-03-15

    Korea Atomic Energy Research Institute, using the ANS low power and shutdown (LPSD) probabilistic risk assessment (PRA) Standard, evaluated the LPSD PSA model of the KSNP, Yonggwang Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the post-accident human errors in the LPSD PSA model for the KSNP showed that 10 items among 19 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for post-accident human errors in the LPSD PSA model for the KSNP. Following tasks are the improvements in the HRA of post-accident human errors of the LPSD PSA model for the KSNP compared with the previous one: Interviews with operators in the interpretation of the procedure, modeling of operator actions, and the quantification results of human errors, site visit. Applications of limiting value to the combined post-accident human errors. Documentation of information of all the input and bases for the detailed quantifications and the dependency analysis using the quantification sheets The assessment results for the new HRA results of post-accident human errors using the ANS LPSD PRA Standard show that above 80% items of its supporting requirements for post-accident human errors were graded as its Category II. The number of the re-estimated human errors using the LPSD Korea Standard HRA method is 385. Among them, the number of individual post-accident human errors is 253. The number of dependent post-accident human errors is 135. The quantification results of the LPSD PSA model for the KSNP with new HEPs show that core damage frequency (CDF) is increased by 5.1% compared with the previous baseline CDF It is expected that this study results will be greatly helpful to improve the PSA quality for the domestic nuclear power plants because they have sufficient PSA quality to meet the Category II of Supporting Requirements for the post-accident

  4. Development and assessment of best estimate integrated safety analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu (and others)

    2007-03-15

    Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published.

  5. Review of core disruptive accident analysis for liquid-metal cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. C.; Na, B. C.; Hahn, D. H

    1997-04-01

    Analysis methodologies of core disruptive accidents (CDAs) are reviewed. The role of CDAS in the overall safety evaluation of fast reactors has not always been well defined nor universally agreed upon. However, they have become a traditional issue in LMR safety, design, and licensing. The study is for the understanding of fast reactor behavior under CDA conditions to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features for the KALIMER developments. The methods used to analyze CDAs from initiating event to complete core disruption are described. Two examples of CDA analyses for CRBRP and ALMR are given and R and D needed for better understanding of CDA phenomena are proposed. (author). 10 refs., 2 tabs., 3 figs

  6. Health effects models for off-site radiological consequence analysis on nuclear reactor accidents (II)

    Energy Technology Data Exchange (ETDEWEB)

    Homma, Toshimitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takahashi, Tomoyuki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Yonehara, Hidenori [National Inst. of Radiological Sciences, Chiba (Japan)] [eds.

    2000-12-01

    This report is a revision of JAERI-M 91-005, 'Health Effects Models for Off-Site Radiological Consequence Analysis of Nuclear Reactor Accidents'. This revision provides a review of two revisions of NUREG/CR-4214 reports by the U.S. Nuclear Regulatory Commission which is the basis of the JAERI health effects models and other several recent reports that may impact the health effects models by international organizations. The major changes to the first version of the JAERI health effects models and the recommended parameters in this report are for late somatic effects. These changes reflect recent changes in cancer risk factors that have come from longer followup and revised dosimetry in major studies on the Japanese A-bomb survivors. This report also provides suggestions about future revisions of computational aspects on health effects models. (author)

  7. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-15

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

  8. Large break loss-of-coolant accident analysis for China Qinshan-2 nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Ban, Chang Hwan; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Wang, Rongzhong; Yu, Hongxing [Nuclear Power Institute of China, Chengdu, SC (China)

    1994-12-01

    Large break LOCA analysis for China Qinshan-2 nuclear power plant has been performed using realistic evaluation model which has been being developed by KAERI. RELAP5/MOD3/KAERI code, which is a modified version of RELAP5/MOD3, is coupled with CONTEMPT4/MOD5 and is used as a best estimate code to predict the thermal hydraulic behavior of the system. PCT uncertainty which stems from code uncertainty, plant application uncertainty, scaling uncertainty and PCT bias are discussed. Among them, plant application uncertainty is described in detail. The licensing PCT is calculated by adding all the uncertainties to the best-estimate PCT. The result indicates the Qinshan-2 nuclear power plant has at least 37 deg C safety margin for large break LOCA. (Author) 10 refs., 47 figs., 14 tabs.

  9. Analysis Codes of Severe Accident%严重事故分析程序

    Institute of Scientific and Technical Information of China (English)

    郎明刚; 高祖瑛

    2002-01-01

    严重事故过程可分为堆芯解体、压力容器熔穿、安全壳失效三阶段,裂变产物随之释放严重事故分析程序有两类,系统性程序与机理性程序.系统性程序能计算完整的事故序列,机理性程序偏重事故进程的局部细节.主要介绍系统性程序STCP、MELCOR、ASTEC与机理性程序RELAP5/SCDAP、VICTORIA、CONTAIN。

  10. Safety analysis of a loss-of-coolant accident in a breeding blanket for experimental fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rocco, P.; Casini, G.; Djerassi, H.; Papa, L.; Pautasso, G.; Renda, V.; Rouyer, J.L.

    1985-07-01

    A LOCA in a blanket design proposed for NET (Next European Torus) is investigated. The structural analysis of a damaged breeder unit shows that this first containment barrier has a high probability of survival to this accident. The radioactive sources involved are evaluated and an assessment is made of all containment barriers and associated protection systems.

  11. SisRadiologia: a new software tool for analysis of radiological accidents and incidents in industrial radiography

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Camila M. Araujo; Silva, Francisco C.A. da, E-mail: araujocamila@yahoo.com.br, E-mail: dasilva@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Araujo, Rilton A., E-mail: consultoria@maximindustrial.com.br [Maxim Industrial Assessoria TI, Rio de Janeiro, RJ (Brazil)

    2013-07-01

    According to the International Atomic Energy Agency (IAEA), many efforts have been made by Member states, aiming a better control of radioactive sources. Accidents mostly happened in practices named as high radiological risk and classified by IAEA in categories 1 and 2, being highlighted those related to radiotherapy, large irradiators and industrial radiography. Worldwide, more than 40 radiological accidents have been recorded in the industrial radiography. Worldwide, more than 40 radiological accidents have been recorded in the industrial radiography area, involving 37 workers, 110 members of the public and 12 fatalities. Records display 5 severe radiological accidents in industrial radiography activities in Brazil, in which 7 workers and 19 members of the public were involved. Such events led to hands and fingers radiodermatitis, but to no death occurrence. The purpose of this study is to present a computational program that allows the data acquisition and recording in the company, in such a way to ease a further detailed analysis of radiological event, besides providing the learning cornerstones aiming the avoidance of future occurrences. After one year of the 'Industrial SisRadiologia' computational program application - and mostly based upon the workshop about Analysis and Dose Calculation of Radiological Accidents in Industrial Radiography (Workshop sobre Analise e Calculo de dose de acidentes Radiologicos em Radiografia Industrial - IRD 2012), in which several Radiation Protection officers took part - it can be concluded that the computational program is a powerful tool to data acquisition, as well as, to accidents and incidents events recording and surveying in Industrial Radiography. The program proved to be efficient in the report elaboration to the Brazilian Regulatory Authority, and very useful in workers training to fix the lessons learned from radiological events.

  12. Serial Concatenation of RS Codes with Kite Codes: Performance Analysis, Iterative Decoding and Design

    CERN Document Server

    Ma, Xiao; Bai, Baoming; Zhang, Xiaoyi

    2011-01-01

    In this paper, we propose a new ensemble of rateless forward error correction (FEC) codes. The proposed codes are serially concatenated codes with Reed-Solomon (RS) codes as outer codes and Kite codes as inner codes. The inner Kite codes are a special class of prefix rateless low-density parity-check (PRLDPC) codes, which can generate potentially infinite (or as many as required) random-like parity-check bits. The employment of RS codes as outer codes not only lowers down error-floors but also ensures (with high probability) the correctness of successfully decoded codewords. In addition to the conventional two-stage decoding, iterative decoding between the inner code and the outer code are also implemented to improve the performance further. The performance of the Kite codes under maximum likelihood (ML) decoding is analyzed by applying a refined Divsalar bound to the ensemble weight enumerating functions (WEF). We propose a simulation-based optimization method as well as density evolution (DE) using Gaussian...

  13. A Study for Appropriateness of National Nuclear Policy by using Economic Analysis Methodology after Fukushima accident

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Jong Myoung; Roh, Myung Sub [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2013-10-15

    The aim of this paper is to clarify the appropriateness of national nuclear policy in BPE of Korea from an economic perspective. To do this, this paper only focus on the economic analysis methodology without any considering other conditions such as political, cultural, or historical things. In a number of countries, especially Korea, nuclear energy policy is keeping the status quo after Fukushima accident. However the nation's nuclear policy may vary depending on the choice of people. Thus, to make the right decisions, it is important to deliver accurate information and knowledge about nuclear energy to the people. As proven in this paper, the levelized cost of nuclear power is the most inexpensive among the base load units. As the reliance on nuclear power is getting stronger through the economic logic, the nuclear safety and environmental elements will be strengthened. Based on this, national nuclear policy should be promoted. In the aftermath of the Fukushima accident recognized as the world's worst nuclear disaster since the Chernobyl, there are some changes in the nuclear energy policy of various countries. Germany, for example, called a halt to operate Nuclear Power Plant (NPP) which accounts for about 7.5% of the national power generation capacity of 6.3GW. In developing countries such as China and India they conducted the safety check of the nuclear power plants again before preceding their nuclear business. Korea government announced 'The 6th Basic Plan for Long-term Electricity Supply and Demand (BPE)', considering the safety and general public acceptance of the nuclear power plants. According to BPE, they postponed a plan for additional NPP construction, except for constructions that had been already reflected in the 5th BPE. All told, the responses for nuclear energy policy of countries are different depending on their own circumstances.

  14. Analysis of the rotation accident of assemblies in boiling water reactors; Analisis del accidente de rotacion de ensambles en reactores de agua en ebullicion

    Energy Technology Data Exchange (ETDEWEB)

    Becerril-Gonzalez M, J. J. [Universidad Autonoma de Yucatan, Av. Industrias no contaminantes por Anillo Periferico Norte s/n, Apdo. Postal 150 Cordemex, Merida, Yucatan (Mexico); Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia de Cueto, R., E-mail: juanjosebecerril_1@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    For this work was analyzed the impact that would cause the load of a rotated fuel assembly in the behaviour of the core in the Cycle 14 of the Unit 1 of the nuclear power plant of Laguna Verde. To carry out this analysis the code Simulate-3 was used, with which was possible to analyze the behavior of the effective multiplication factor and the thermal limits (MAPRAT, MFLPD and MFLCPR). The rotation of fuel assemblies to 90, 180 and 270 grades was analyzed with regard to the design position, with 0, 1, 2 and 3 burnt cycles for these assemblies. The results show that the thermal limits remain inside the allowed values, therefore if this accident type happened the reactor could continue operating in a sure way. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

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

  16. WORK ACCIDENTS ANALYSIS, EMPHASIZING THE FOREST SECTORS, IN A FEDERAL HIGH EDUCATION INSTITUTION

    Directory of Open Access Journals (Sweden)

    Héder Alencar Vianna

    2008-09-01

    Full Text Available This research analyzed the probable causes of accidents at work, of the forest sectors, of a Federal Institution of Higher education in Minas Gerais State looking for minimizing risks damages to worker’s health and safety. The analyses were accomplished by means of the Accidents Communications in Service data bank and the results showed that in the studied period, for the forest sector, 50.71% of the accidents were attributed to the insecure act, 16.98% to insecure condition and 17.94% to personal factor of insecurity. The functional evaluation indicated that in 7.14% of the accidents occurred with workers that were not in the function established by the employment respective contracts. The rates of gravity and frequency of accidents calculated for the forest sectors allowed to conclude that the indices were higher than the others sectors and much higher than the values published by the Brazilian Ministry of Social Service and Assistance.

  17. Performance Analysis for Dispensing Mechanism of Active Code

    Institute of Scientific and Technical Information of China (English)

    2003-01-01

    The dispensing mechanism of active code is a key technology in an active network. Conventional capsule and programmable switch approaches have their own shortcomings. The DCCAN(distributed code caching for active network) mechanism presented in this paper overcomes these shortcomings. In this paper, capsule and programmable switch approaches are introduced, and their shortcomings are analyzed. The principle of the DCCAN mechanism is described. The theory analysis in transmit width based on the DCCAN mechanism and capsule approach are described. The theory analysis shows that the DCCAN mechanism has many good characteristics and can improve the efficiency of an active network. Key factors which affect the transmit width based on the DCCAN mechanism are discussed. The using condition of the DCCAN mechanism is also discussed.

  18. An epidemiologic survey of road traffic accidents in Iran: analysis of driver-related factors.

    Science.gov (United States)

    Moafian, Ghasem; Aghabeigi, Mohammad-Reza; Heydari, Seyed Taghi; Hoseinzadeh, Amin; Lankarani, Kamran Bagheri; Sarikhani, Yaser

    2013-01-01

    Road traffic accident (RTA) and its related injuries contribute to a significant portion of the burden of diseases in Iran. This paper explores the association between driver-related factors and RTA in the country. This cross-sectional study was conducted in Iran and all data regarding RTAs from March 20, 2010 to June 10, 2010 were obtained from the Traffic Police Department. We included 538 588 RTA records, which were classified to control for the main confounders: accident type, final cause of accident, time of accident and driver-related factors. Driver-related factors included sex, educational level, license type, type of injury, duration between accident and getting the driving license and driver's error type. A total of 538 588 drivers (91.83% male, sex ratio of almost 13:1) were involved in the RTAs. Among them 423 932 (78.71%) were uninjured; 224 818 (41.74%) had a diploma degree. Grade 2 driving license represented the highest proportion of all driving licenses (290 811, 54.00%). The greatest number of accidents took place at 12:00-13:59 (75 024, 13.93%). The proportion of drivers involved in RTAs decreased from 15.90% in the first year of getting a driving license to 3.13% after 10 years'of driving experience. Neglect of regulations was the commonest cause of traffic crashes (345 589, 64.17%). Non-observance of priority and inattention to the front were the most frequent final causes of death (138 175, 25.66% and 129 352, 24.02%, respectively). We found significant association between type of accident and sex, education, license type, time of accident, final cause of accident, driver's error as well as duration between accident and getting the driving license (all P less than 0.001). Our results will improve the traffic law enforcement measures, which will change inappropriate behavior of drivers and protect the least experienced road users.

  19. Prediction of the Containment Pressure under Severe Accidents Using CFNN

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Young Do; Choi, Geon Pil; Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2016-10-15

    It is important to keep containment integrity by measuring main risk factors such as temperature and hydrogen concentration that occur pressure rise in the containment and by operating safety features at the right time. In this study, the circumstance that instrumentation equipment in NPPs is uncertain under severe accidents after DBA is assumed. This is to keep containment integrity by manually generating the safety injection actuation signal (SIAS) and to assess integrity of accident equipment through early prediction of the containment pressure under extreme circumstances when main factors such as temperature and hydrogen concentration that rise pressure in containment may not be adequately measured. In this study, the cascaded fuzzy neural network (CFNN) model is used to predict containment pressure using LOCA break sizes as input data. Because the real severe accident data cannot be obtained from actual NPP accidents, they were gained by numerically simulating severe accident scenarios of the optimized power reactor (OPR1000) using modular accident analysis program (MAAP) code. Temperature and hydrogen concentration in the containment are risk factors that increase containment pressure under severe accidents. Therefore, LOCA sizes as the input data for the CFNN model are used to predict the containment pressure. This input data is the simulation data obtained by using MAAP4 code for the OPR1000 reactor. As a result of using the CFNN model, the RMS errors are within 0.4% to 1.4. Accordingly, The CFNN could be a model that reliably predict the containment pressure and the data through CFNN model could figure out the containment integrity and assess the survivability of severe accident equipment under accidents.

  20. Bicycle accidents.

    Science.gov (United States)

    Lind, M G; Wollin, S

    1986-01-01

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

  1. LACEwING: A New Moving Group Analysis Code

    Science.gov (United States)

    Riedel, Adric R.; Blunt, Sarah C.; Lambrides, Erini L.; Rice, Emily L.; Cruz, Kelle L.; Faherty, Jacqueline K.

    2017-03-01

    We present a new nearby young moving group (NYMG) kinematic membership analysis code, LocAting Constituent mEmbers In Nearby Groups (LACEwING), a new Catalog of Suspected Nearby Young Stars, a new list of bona fide members of moving groups, and a kinematic traceback code. LACEwING is a convergence-style algorithm with carefully vetted membership statistics based on a large numerical simulation of the Solar Neighborhood. Given spatial and kinematic information on stars, LACEwING calculates membership probabilities in 13 NYMGs and three open clusters within 100 pc. In addition to describing the inputs, methods, and products of the code, we provide comparisons of LACEwING to other popular kinematic moving group membership identification codes. As a proof of concept, we use LACEwING to reconsider the membership of 930 stellar systems in the Solar Neighborhood (within 100 pc) that have reported measurable lithium equivalent widths. We quantify the evidence in support of a population of young stars not attached to any NYMGs, which is a possible sign of new as-yet-undiscovered groups or of a field population of young stars.

  2. Development of Parameter Network for Accident Management Applications

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  3. On the use of ZIP codes and ZIP code tabulation areas (ZCTAs for the spatial analysis of epidemiological data

    Directory of Open Access Journals (Sweden)

    Matisziw Timothy C

    2006-12-01

    Full Text Available Abstract Background While the use of spatially referenced data for the analysis of epidemiological data is growing, issues associated with selecting the appropriate geographic unit of analysis are also emerging. A particularly problematic unit is the ZIP code. Lacking standardization and highly dynamic in structure, the use of ZIP codes and ZIP code tabulation areas (ZCTA for the spatial analysis of disease present a unique challenge to researchers. Problems associated with these units for detecting spatial patterns of disease are explored. Results A brief review of ZIP codes and their spatial representation is conducted. Though frequently represented as polygons to facilitate analysis, ZIP codes are actually defined at a narrower spatial resolution reflecting the street addresses they serve. This research shows that their generalization as continuous regions is an imposed structure that can have serious implications in the interpretation of research results. ZIP codes areas and Census defined ZCTAs, two commonly used polygonal representations of ZIP code address ranges, are examined in an effort to identify the spatial statistical sensitivities that emerge given differences in how these representations are defined. Here, comparative analysis focuses on the detection of patterns of prostate cancer in New York State. Of particular interest for studies utilizing local, spatial statistical tests, is that differences in the topological structures of ZIP code areas and ZCTAs give rise to different spatial patterns of disease. These differences are related to the different methodologies used in the generalization of ZIP code information. Given the difficulty associated with generating ZIP code boundaries, both ZIP code areas and ZCTAs contain numerous representational errors which can have a significant impact on spatial analysis. While the use of ZIP code polygons for spatial analysis is relatively straightforward, ZCTA representations contain

  4. Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Han Sul; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Kim, Tae Wan [Incheon National University, Incheon (Korea, Republic of)

    2017-03-15

    The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective.

  5. Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

  6. [Spatial analysis of road traffic accidents with fatalities in Spain, 2008-2011].

    Science.gov (United States)

    Gómez-Barroso, Diana; López-Cuadrado, Teresa; Llácer, Alicia; Palmera Suárez, Rocío; Fernández-Cuenca, Rafael

    2015-09-01

    To estimate the areas of greatest density of road traffic accidents with fatalities at 24 hours per km(2)/year in Spain from 2008 to 2011, using a geographic information system. Accidents were geocodified using the road and kilometer points where they occurred. The average nearest neighbor was calculated to detect possible clusters and to obtain the bandwidth for kernel density estimation. A total of 4775 accidents were analyzed, of which 73.3% occurred on conventional roads. The estimated average distance between accidents was 1,242 meters, and the average expected distance was 10,738 meters. The nearest neighbor index was 0.11, indicating that there were aggregations of accidents in space. A map showing the kernel density was obtained with a resolution of 1 km(2), which identified the areas of highest density. This methodology allowed a better approximation to locating accident risks by taking into account kilometer points. The map shows areas where there was a greater density of accidents. This could be an advantage in decision-making by the relevant authorities. Copyright © 2014 SESPAS. Published by Elsevier Espana. All rights reserved.

  7. Analysis of Surface Water Pollution Accidents in China: Characteristics and Lessons for Risk Management

    Science.gov (United States)

    Yao, Hong; Zhang, Tongzhu; Liu, Bo; Lu, Feng; Fang, Shurong; You, Zhen

    2016-04-01

    Understanding historical accidents is important for accident prevention and risk mitigation; however, there are no public databases of pollution accidents in China, and no detailed information regarding such incidents is readily available. Thus, 653 representative cases of surface water pollution accidents in China were identified and described as a function of time, location, materials involved, origin, and causes. The severity and other features of the accidents, frequency and quantities of chemicals involved, frequency and number of people poisoned, frequency and number of people affected, frequency and time for which pollution lasted, and frequency and length of pollution zone were effectively used to value and estimate the accumulated probabilities. The probabilities of occurrences of various types based on origin and causes were also summarized based on these observations. The following conclusions can be drawn from these analyses: (1) There was a high proportion of accidents involving multi-district boundary regions and drinking water crises, indicating that more attention should be paid to environmental risk prevention and the mitigation of such incidents. (2) A high proportion of accidents originated from small-sized chemical plants, indicating that these types of enterprises should be considered during policy making. (3) The most common cause (49.8 % of the total) was intentional acts (illegal discharge); accordingly, efforts to increase environmental consciousness in China should be enhanced.

  8. Analysis on distribution of freeway accidents under various conditions in China

    Directory of Open Access Journals (Sweden)

    Xiaofei Wang

    2016-08-01

    Full Text Available This study aims to provide a current survey on the situation of freeway accidents in China. The results show that the accident rate, death toll, injury toll, and direct loss of property are 3.2, 8.4, 7.2, and 24.3 times that of the average for an ordinary highway in China. Freeway accidents occur mainly in Southern (20.77% and Central (20.2% China. With detailed data from Guangdong Province, the number of accidents in freeways with 80 km/h design speed (29.58/km/103 pcu was more than three times that in freeways with design speed of 100 km/h (9.54/km/103 pcu and 120 km/h (9.42/km/103 pcu. The total accident rate increased monotonously with the decrease in horizontal radius. The results indicate that 54.54% (/km/103 pcu of accidents occurred on a steep slope (4%–5%, representing about 10 times that of 3%–4% slope and 20 times that of the less than 3% slope. Based on the data, the safety situation of China’s highway transportation is obviously grim, and improving freeway heavy traffic management in economically developed regions, strengthening the safeguarding of mountain freeways, applying small radius and large vertical grade with caution, and developing a monitoring system of tunnels and interchanges could be used as effective measures to prevent freeway accidents.

  9. Nonresponse analysis and adjustment in a mail survey on car accidents.

    Science.gov (United States)

    Tivesten, Emma; Jonsson, Sofia; Jakobsson, Lotta; Norin, Hans

    2012-09-01

    Statistical accident data plays an important role for traffic safety development involving the road system, vehicle design, and driver education. Vehicle manufacturers use data from accident mail surveys as an integral part of the product development process. Low response rates has, however, lead to concerns on whether estimates from a mail survey can be trusted as a source for making strategic decisions. The main objective of this paper was to investigate nonresponse bias in a mail survey addressing driver behaviour in accident situations. Insurance data, available for both respondents and nonrespondents were used to analyze, as well as adjust for nonresponse. Response propensity was investigated by using descriptive statistics and logistic regression analyses. The survey data was then weighted by using inverse propensity weights. Two specific examples of survey estimates are addressed, namely driver vigilance and driver's distraction just before the accident. The results from this paper reveal that driver age and accident type were the most influential variables for nonresponse weighting. Driver gender and size of town where the driver resides also had some influence, but not for all survey variables investigated. The main conclusion of this paper is that nonresponse weighting can increase confidence in accident data collected by a mail survey, especially when response rates are low. Weighting has a moderate influence on this survey, but a larger influence may be expected if applied on a more diverse driver population. The development of auxiliary data collection can further improve accident mail survey methodology in future. Copyright © 2012 Elsevier Ltd. All rights reserved.

  10. A Human Reliability Analysis of Pre-Accident Human Errors in the Low Power and Shutdown PSA of the KSNP

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Daeil; Jang, Seungchul

    2007-03-15

    Korea Atomic Energy Research Institute, using the ANS Low Power /Shutdown (LPSD)PRA Standard, evaluated the LPSD PSA model of the KSNP, Younggwang (YGN) Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the pre-accident human errors in the LPSD PSA model of the KSNP showed that 13 items among 15 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for pre-accident human errors in the LPSD PSA model for the KSNP to improve its quality. We considered potential pre-accident human errors for all manual valves and control/instrumentation equipment of the systems modeled in the KSNP LPSD PSA model except reactor protection system/ engineering safety features actuation system. We reviewed 160 manual valves and 56 control/instrumentation equipment. The number of newly identified pre-accident human errors is 101. Among them, the number of those related to testing/maintenance tasks is 56. The number of those related to calibration tasks is 45. The number of those related to only shutdown operation is 10. It was shown that the pre-accident human errors related to only shutdown operation contributions to the core damage frequency of LPSD PSA model for the KSNP was negligible.The self-assessment results for the new HRA results of pre-accident human errors using the ANS LPSD PRA Standard show that above 80% items of its supporting requirements for post-accident human errors were graded as its Category II or III. It is expected that the HRA results for the pre-accident human errors presented in this study will be greatly helpful to improve the PSA quality for the domestic nuclear power plants because they have sufficient PSA quality to meet the Category II of supporting requirements for the postaccident human errors in the ANS LPSD PRA Standard.

  11. A Human Reliability Analysis of Pre-Accident Human Errors in the Low Power and Shutdown PSA of the KSNP

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Daeil; Jang, Seungchul

    2007-03-15

    Korea Atomic Energy Research Institute, using the ANS Low Power /Shutdown (LPSD)PRA Standard, evaluated the LPSD PSA model of the KSNP, Younggwang (YGN) Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the pre-accident human errors in the LPSD PSA model of the KSNP showed that 13 items among 15 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for pre-accident human errors in the LPSD PSA model for the KSNP to improve its quality. We considered potential pre-accident human errors for all manual valves and control/instrumentation equipment of the systems modeled in the KSNP LPSD PSA model except reactor protection system/ engineering safety features actuation system. We reviewed 160 manual valves and 56 control/instrumentation equipment. The number of newly identified pre-accident human errors is 101. Among them, the number of those related to testing/maintenance tasks is 56. The number of those related to calibration tasks is 45. The number of those related to only shutdown operation is 10. It was shown that the pre-accident human errors related to only shutdown operation contributions to the core damage frequency of LPSD PSA model for the KSNP was negligible.The self-assessment results for the new HRA results of pre-accident human errors using the ANS LPSD PRA Standard show that above 80% items of its supporting requirements for post-accident human errors were graded as its Category II or III. It is expected that the HRA results for the pre-accident human errors presented in this study will be greatly helpful to improve the PSA quality for the domestic nuclear power plants because they have sufficient PSA quality to meet the Category II of supporting requirements for the postaccident human errors in the ANS LPSD PRA Standard.

  12. Thermal-hydraulic analysis of a heavy-water reactor moderator tank using the CUPID Code

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Su Ryong; Jeong, Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Kim, Hyoung Tae; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, a preliminary analysis is performed for the CANDU moderator tank. The calculation results using the basic case input showed a unrealistic, thermal stratification in the upper region, which was caused by the lack of the momentum of the cooling water from the inlet nozzle. To increase the flow momentum from the inlet nozzle, the cross-section area of each inlet nozzle was reduced by half and, then, the calculation showed very realistic results. It is clear that the modeling of the inlet nozzle affects the calculation result significantly. Further studies are needed for a realistic and efficient simulation of the flow in the Calandria tank. When the core cooling system fails to remove the decay heat from the fuel channels during a loss of coolant accident (LOCA), the pressure tube (PT) could strain to contact its surrounding Calandria tube (CT), which leads to sustained CTs dry out, finally resulting in damages to nuclear fuel. This situation can occur when the degree of the subcooling of the moderator inside the Calandria vessel is insufficient. In this regard, to estimate the local subcooling of the moderator inside the Calandria vessel is very important. However, the local temperature is measured at the inlet and outlet of the vessel only. Therefore, we need to accurately predict the local temperature inside the Calandria vessel.In this study, the thermal-hydraulic analysis of the real-scale heavy-water reactor moderator is carried out using the CUPID code. The applicability of the CUPID code to the analysis of the flow in the Calandria vessel has been assessed in the previous studies.

  13. The effect of gamma-ray transport on afterheat calculations for accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, S.; Latkowski, J.F.; Sanz, J.

    2000-05-01

    Radioactive afterheat is an important source term for the release of radionuclides in fusion systems under accident conditions. Heat transfer calculations are used to determine time-temperature histories in regions of interest, but the true source term needs to be the effective afterheat, which considers the transport of penetrating gamma rays. Without consideration of photon transport, accident temperatures may be overestimated in others. The importance of this effect is demonstrated for a simple, one-dimensional problem. The significance of this effect depends strongly on the accident scenario being analyzed.

  14. Development of NASA's Accident Precursor Analysis