WorldWideScience

Sample records for accident conditions key

  1. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  2. Key Characteristics of Combined Accident including TLOFW accident for PSA Modeling

    Kim, Bo Gyung; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Technology and Research, Abu Dhabi (United Arab Emirates)

    2015-05-15

    The conventional PSA techniques cannot adequately evaluate all events. The conventional PSA models usually focus on single internal events such as DBAs, the external hazards such as fire, seismic. However, the Fukushima accident of Japan in 2011 reveals that very rare event is necessary to be considered in the PSA model to prevent the radioactive release to environment caused by poor treatment based on lack of the information, and to improve the emergency operation procedure. Especially, the results from PSA can be used to decision making for regulators. Moreover, designers can consider the weakness of plant safety based on the quantified results and understand accident sequence based on human actions and system availability. This study is for PSA modeling of combined accidents including total loss of feedwater (TLOFW) accident. The TLOFW accident is a representative accident involving the failure of cooling through secondary side. If the amount of heat transfer is not enough due to the failure of secondary side, the heat will be accumulated to the primary side by continuous core decay heat. Transients with loss of feedwater include total loss of feedwater accident, loss of condenser vacuum accident, and closure of all MSIVs. When residual heat removal by the secondary side is terminated, the safety injection into the RCS with direct primary depressurization would provide alternative heat removal. This operation is called feed and bleed (F and B) operation. Combined accidents including TLOFW accident are very rare event and partially considered in conventional PSA model. Since the necessity of F and B operation is related to plant conditions, the PSA modeling for combined accidents including TLOFW accident is necessary to identify the design and operational vulnerabilities.The PSA is significant to assess the risk of NPPs, and to identify the design and operational vulnerabilities. Even though the combined accident is very rare event, the consequence of combined

  3. Containment leakage during severe accident conditions

    An alternate to the THRESHOLD model used in most severe accident risk assessments has been investigated. One reference plant for each of six containment types has been studied to determine the magnitude of containment leakage that would result from the pressures and temperatures associated with severe accident conditions. Containment penetrations having the greatest potential for early containment leakage are identified. The studies indicate that containment leakage through penetrations prior to reaching containment threshold pressures (currently reported containment shell failure pressures) should be considered in severe accident risk assessments. Failure of non-metallic seals for containment penetrations can be a significant source of containment leakage under severe accident pressure and temperature conditions. Although studies of containment types are useful in identifying sources of containment leakage, final conclusions may need to be plant specific. Recommendations concerning future studies to better develop the use of continuous leakage models are provided. 9 references, 4 figures, 2 tables

  4. Study on accident response robot for nuclear power plant and analysis of key technologies

    With the rapid development of nuclear power industry and improving demand for nuclear safety, the demand for developing accident response robot in nuclear power plant is increasingly urgent. Firstly, design analysis for accident response robot is taken with environmental conditions in nuclear power plant. Secondly, development for response robots after Chernobyl, JCO and Fukushima accidents are reviewed, and improvements for commercial mobile robot for use in radioactive environments are summarized. Finally, some key technologies including radiation-tolerance and system reliability are analyzed in details. (authors)

  5. Man-machine interaction in accident conditions

    The paper concerns the current activities in the area of enhancing the man-machine interface in accident conditions and stresses that the technique of artificial intelligence is the best way to attain significant progress in nuclear safety. The peculiarities of the WWER-440, model V-230, are discussed from the point of view of accident monitoring and management. Two expert systems - SAMES and RPES - are designated as operator aids in the event of an NPP accident with a radiation release. It is important to vary the content and the structure of the knowledge bases, depending on the user's requirements and responsibilities. Independent of the fact that both expert systems include some similar functions, for instance identification of the class of the accident, the diagnostic modules are different. This difference concerns the level of abstraction in pattern recognition and the different knowledge domains. RPES also includes different deterministic models for atmospheric transport, identification of the endangered area and estimation of the dose equivalents to the public. These allow the implementation of different protective measures to reduce the risk to the population. (author). 5 refs, 3 figs

  6. 10 CFR 71.73 - Hypothetical accident conditions.

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Hypothetical accident conditions. 71.73 Section 71.73... Package, Special Form, and LSA-III Tests 2 § 71.73 Hypothetical accident conditions. (a) Test procedures. Evaluation for hypothetical accident conditions is to be based on sequential application of the...

  7. ACCOUNT OF ROAD CONDITIONS WHILE INVESTIGATING TRAFFIC ACCIDENTS

    D. D. Selioukov; I. I. Leonovich

    2014-01-01

    The paper considers problems on better traffic safety at government, authority, engineering and driver activity levels, account of road conditions while investigating traffic accidents. The paper also provides road defects mentioned in forensic transport examinations of traffic accidents.

  8. Primary pump vibration under accident conditions

    This report presents the results of an international survey on the subject of vibration in nuclear primary coolant pumps due to two-phase flow, accident conditions. The literature search also revealed few Canadian references other than those of Ontario Hydro. Ontario Hydro's work has been extensive. Confidence in the mechanical integrity of the pumpsets is good, given the extent of the testing. However, conclusions with respect to piping integrity and thermal-hydraulic performance are difficult to determine due to the inexact geometry of the piping and the difficulties in estimating fluid conditions at the pump. The tests help to understand the phenomena and provide background information for analysis, but should be applied with caution to plant analyses. Much of the discussion in the report relates to pump head instability. This is perceived to be the most important flow regime causing vibration, as attested by the emphasis of the reviewed literature. A method for quantitative assessment of the forcing functions acting on the pump-piping system due to void generation and collapse is recommended. A relatively fundamental analytical approach is proposed, supplemented by reduced scale testing in the latter stages. 151 refs

  9. Network conditioning under conflicting goals: Accident causation

    Networks based on the Barto-Sutton architecture (BSA) of neural-like elements have an information-processing structure that is analogous to the cognitive structure of a human. Given a set of explicitly stated rules of conduct, such networks develop a set of skills that is capable of satisfying the rules. In this sense, the network acts as a translator of rules into skill-based behavior. The BSA acquires its skills through casual, correlation-based scheduling. Stated briefly, it first constructs an internal representation, or model, of the rules of conduct, and then uses the model to correct deficiencies in its skill. It learns in a manner that closely resembles classical conditioning, shifting the onset of signals associated with unconditioned stimuli forward in time to coincide with the onset of conditioning stimuli. The low-level positive reinforcement the network receives from enhancing its operational efficiency is immediate and direct. In the absence of countervailing influences, this continuous pressure is sufficient to discount the recollection of past failures and leads to accidents with a predictable regularity

  10. Digital rate meters in radiological instruments for accident conditions

    Advantages of digital rate meter, when applied in radiological protection instruments for accident conditions, re discussed. Some requirements imposed by accident conditions on such instruments are indicated. The specific properties of digital rate meters are shown and some solutions which enable realisation of defined tasks are pointed out. The behaviour of one solution is illustrated by an example simulated on the computer. (author)

  11. Using modular neural networks to monitor accident conditions in nuclear power plants

    Nuclear power plants are very complex systems. The diagnoses of transients or accident conditions is very difficult because a large amount of information, which is often noisy, or intermittent, or even incomplete, need to be processed in real time. To demonstrate their potential application to nuclear power plants, neural networks axe used to monitor the accident scenarios simulated by the training simulator of TVA's Watts Bar Nuclear Power Plant. A self-organization network is used to compress original data to reduce the total number of training patterns. Different accident scenarios are closely related to different key parameters which distinguish one accident scenario from another. Therefore, the accident scenarios can be monitored by a set of small size neural networks, called modular networks, each one of which monitors only one assigned accident scenario, to obtain fast training and recall. Sensitivity analysis is applied to select proper input variables for modular networks

  12. A radioactive waste transportation package monitoring system for normal transport and accident emergency response conditions

    This paper addresses spent fuel and high level waste transportation history and prospects, discusses accident histories of radioactive material transport, discusses emergency responder needs and provides a general description of the Transportation Intelligent Monitoring System (TRANSIMS) design. The key objectives of the monitoring system are twofold: (1) to facilitate effective emergency response to accidents involving a radioactive waste transportation package, while minimizing risk to the public and emergency first-response personnel, and (2) to allow remote monitoring of transportation vehicle and payload conditions to enable research into radioactive material transportation for normal and accident conditions. (J.P.N.)

  13. Key regulatory and safety issues emerging NEA activities. Lessons Learned from Fukushima Dai-ichi NPS Accident - Key Regulatory and Safety Issues

    A presentation was provided on the key safety and regulatory issues and an update of activities undertaken by the NEA and its members in response to the accident at the Fukushima Daiichi nuclear power stations (NPS) on 11 March 2011. An overview of the accident sequence and the consequences was provided that identified the safety functions that were lost (electrical power, core cooling, and primary containment) that lead to units 1, 2, and 3 being in severe accident conditions with large off-site releases. Key areas identified for which activities of the NEA and member countries are in progress include accident management; defence-in-depth; crisis communication; initiating events; operating experience; deterministic and probabilistic assessments; regulatory infrastructure; radiological protection and public health; and decontamination and recovery. For each of these areas, a brief description of the on-going and planned NEA activities was provided within the three standing technical committees of the NEA with safety and regulatory mandates (the Committee on Nuclear Regulatory Activities - CNRA, the Committee on the Safety of Nuclear Installations - CSNI, and the Committee on Radiation Protection and Public Health - CRPPH). On-going activities of CNRA include a review of enhancement being made to the regulatory aspects for the oversight of on-site accident management strategies and processes in light of the lessons learned from the accident; providing guidance to regulators on crisis communication; and supporting the peer review of the safety assessments of risk-significant research reactor facilities in light of the accident. Within the scope of the CSNI mandate, activities are being undertaken to better understand accident progression; characteristics of new fuel designs; and a benchmarking study of fast-running software for estimating source term under severe accident conditions to support protective measure recommendations. CSNI also has ongoing work in human

  14. Development of solution behavior observation system under criticality accident conditions

    A solution behavior observation system was developed for observing the behavior of fissile solution and radiolytic voids under criticality accident conditions in TRACY. The system consisted of a radiation-resistive optical fiberscope and a CCD color video camera. The system functioned properly in the mixed high radiation fields of gamma rays and neutrons under criticality accident conditions, and it succeeded in taking the images of their behavior. They provide an important information to understand phenomena of fuel solution at criticality accidents and to construct computational kinetic models. The images can also be used as teaching materials for plant workers and students in universities. (author)

  15. Graphite Oxidation Simulation in HTR Accident Conditions

    El-Genk, Mohamed

    2012-10-19

    Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.

  16. Full-length fuel rod behavior under severe accident conditions

    This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors

  17. Fuel pins and core response under LMFBR top accident conditions

    Out-of-reactor experiments are currently being performed at Argonne National Laboratory to examine fuel sweepout and related post-failure phenomena under hypothetical TOP accident conditions. These tests are supplementing the TREAT MARK-II loop data base by keying on effects of important parameter variations such as system hydraulics and intrabundle coherency. In these tests, molten UO2, generated by a thermite reaction at 34700K, is injected over approximately 40 msec into flowing sodium in a bundle of simulated LMFBR-type fuel pins. Hydraulic conditions in the bundle are selected to match conditions in either the MARK-II loop (HUMP-series) or the current design LMFBR subassembly (CAMEL-series). To date, four tests have been performed in both single-pin and seven-pin configurations representing coherent and incoherent subassembly power-to-flow cases, respectively. Details of the fuel motion were observed using a flash x-ray cine system. A compilation of significant findings from the four sweepout tests is presented

  18. Preliminary Assessment of Accident Tolerant Fuel Performance at Normal and Accident Conditions

    The interest for improving the safety of light water reactors (LWRs) fuel designs, which has significantly grown after the Fukushima Daiichi Accident, has driven the U.S. Department of Energy (DOE) to fund three industry-led programs to facilitate the development of accident tolerant fuels (ATF) for LWRs. Westinghouse is leading one of them and engaged in developing a combined accident resistant cladding and high density fuel pellet. It is important to develop and apply fuel performance codes and other computational methods to model the novel fuel forms to better understand the in-core performance and to guide new fuel designs. In this paper, a preliminary assessment on the performance of various ATF concepts during normal and accident conditions is presented. These concepts include various combinations of accident tolerant fuel and cladding materials: UN/SiC, U3Si2/SiC, UN/Coated Zircaloy, and U3Si2/Coated Zircaloy. The properties of the new materials were collected from literature and their irradiation data will be selected from various test reactor experiments. The impact of ATF properties on design basis accidents and beyond design basis accident is also discussed. (author)

  19. Evaluation of current regulations and real accident conditions

    In order to improve estimates of the effectiveness of current regulatory standards, a program was initiated by the US Nuclear Regulatory Commission (NRC) to have the Lawrence Livermore Laboratory (LLNL) evaluate regulatory standards against real world accident conditions. This paper presents the results of the evaluation performed for the hypothetical 30-foot drop onto an unyielding surface and real world impact conditions which might be experienced by a spent fuel cask being transported by a truck. The results of the evaluations performed for other pertinent accident conditions for truck and train transport will be documented at the conclusion of the program. 10 refs., 8 figs., 3 tabs

  20. Accident management advisor system (AMAS): A Decision Aid for Interpreting Instrument Information and Managing Accident Conditions in Nuclear Power Plants

    for the development of models specifically tailored to real-time accident management. While it is almost impossible to develop and utilize exact models of the evolution of all possible accident sequences for each given type of nuclear power plant and containment design (e.g., PWR and BWR designs at various power ratings, large dry containment or ice condenser types, etc.), it appears possible to develop a sound approach to monitor the progression of an accident with respect to the integrity and effectiveness of a set of principal safety functions. The key to doing this is the development of a knowledge base 'housing structure', where uncertain knowledge regarding the predicted plant behavior and real-time, but also uncertain, information compiled from plant instrumentation readings can be compared and matched to produce the best possible identification of plant states and possible accident control actions. In summary: This paper illustrates the concept and the architecture of the Accident Management Advisor System, a decision aid which enables the use of combined instrument information to reduce uncertainty in decision making associated with nuclear plant accident conditions. The principal benefits offered by this concept are the definition of an approach to utilize instrument information under uncertain accident conditions in such a way as to allow the best possible assessment of plant status and the implementation of a formalized accident management decision-making strategy by means of a computer-based operator assistance tool. When fully developed, we expect AMAS to find application in both the commercial and government sections of the U.S. nuclear industry. We currently plan to have a working prototype of the system, ready to demonstrate its functionality for a representative commercial PWR plant, by the end of the next phase of our research, in which both model development and software development activities will have to be carried out. Finally, the AMAS

  1. Response of HEPA filters to simulated-accident conditions

    High-efficiency particulate air (HEPA) filters have been subjected to simulated accident conditions to determine their response to abnormal operating events. Both domestic and European standard and high-capacity filters have been evaluated to determine their response to simulated fire, explosion, and tornado conditions. The HEPA filter structural limitations for tornado and explosive loadings are discussed. In addition, filtration efficiencies during these accident conditions are reported for the first time. Our data indicate efficiencies between 80% and 90% for shock loadings below the structural limit level. We describe two types of testing for ineffective filtration - clean filters exposed to pulse-entrained aerosol and dirty filters exposed to tornado and shock pulses. Efficiency and material loss data are described. Also, the resonse of standard HEPA filters to simulated fire conditions is presented. We describe a unique method of measuring accumulated combustion products on the filter. Additionally, data relating to pressure drop vs accumulated mass during plugging are reported for simulated combustion aerosols. The effects of concentration and moisture levels on filter plugging were evaluated. We are obtaining all of the above data so that mathematical models can be developed for fire, explosion, and tornado accident analysis computer codes. These computer codes can be used to assess the response of nuclear air cleaning systems to accident conditions

  2. Predictions of structural integrity of steam generator tubes under normal operating, accident, and severe accident conditions

    Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating, accident, and severe accident conditions are reviewed. Tests conducted in the past, though limited, tended to show that the earlier flow-stress model for part-through-wall axial cracks overestimated the damaging influence of deep cracks. This observation is confirmed by further tests at high temperatures as well as by finite element analysis. A modified correlation for deep cracks can correct this shortcoming of the model. Recent tests have shown that lateral restraint can significantly increase the failure pressure of tubes with unsymmetrical circumferential cracks. This observation is confirmed by finite element analysis. The rate-independent flow stress models that are successful at low temperatures cannot predict the rate sensitive failure behavior of steam generator tubes at high temperatures. Therefore, a creep rupture model for predicting failure is developed and validated by tests under varying temperature and pressure loading expected during severe accidents

  3. Countermeasures for traffic accidents due to road conditions in China

    PEI Yu-long; MA Ji

    2005-01-01

    Regarding the postulate of traffic infrastmcture and vehicles, much attention should be given to the effect of road conditions on accidents. With large numbers of traffic accidents on Shenda Freeway, Liaoning Province, Harbin City and others in P. R. China, parameters and the effect of accidents caused by horizontal alignment, vertical alignment, cross section and intersection are studied systematically The disciplinary analysis of these effects are presented in this paper. The viewpoint is acknowledged that high sub grade and steep slopes are against traffic safety, which is common and ignored in high-usage highways in China. Design parameters of the current design criteria and the corresponding countermeasures are suggested for safety on our highways.

  4. Hydrogen-control systems for severe LWR accident conditions - a state-of-technology report

    This report reviews the current state of technology regarding hydrogen safety issues in light water reactor plants. Topics considered in this report relate to control systems and include combustion prevention, controlled combustion, minimization of combustion effects, combination of control concepts, and post-accident disposal. A companion report addresses hydrogen generation, distribution, and combustion. The objectives of the study were to identify the key safety issues related to hydrogen produced under severe accident conditions, to describe the state of technology for each issue, and to point out ongoing programs aimed at resolving the open issues

  5. Investigation of VVER 1000 Fuel Behavior in Severe Accident Condition

    This paper presents the results obtained during a simulation of fuel behavior with the MELCOR computer code in case of severe accident for the VVER reactor core. The work is focused on investigating the influence of some important parameters, such as porosity, on fuel behavior starting from oxidation of the fuel cladding, fusion product release in the primary circuit after rupture of the fuel cladding, melting of the fuel and reactor core internals and its further relocation to the bottom of the reactor vessel. In the analyses are modeled options for blockage of melt and debris during its relocation. In the work is investigated the uncertainty margin of reactor vessel failure based on modeling of the reactor core and an investigation of its behavior. This is achieved by performing sensitivity analyses for VVER 1000 reactor core with gadolinium fuel type. The paper presents part of the work performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research. The performed work continues the effort in the modeling of fuel behavior during severe accidents such as Station Blackout sequence for VVER 1000 reactors based on parametric study. The work is oriented towards the investigation of fuel behavior during severe accident conditions starting from the initial phase of fuel damaging through melting and relocation of fuel elements and reactor internals until the late in-vessel phase, when melt and debris are relocated almost entirely on the bottom head of the reactor vessel. The received results can be used in support of PSA2 as well as in support of analytical validation of Sever Accident Management Guidance for VVER 1000 reactors. The main objectives of this work area better understanding of fuel behavior during severe accident conditions as well as plant response in such situations. (author)

  6. Investigations on pressure suppression system loads at accident conditions

    For simulation of the integral behavior of pressure suppresion systems at accident conditions a mathematical model was developed which simulates a wide range of the loads occurring during a loss-of coolant accident. The multi-zone point model DRASYS serves for mathematical simulation of quasistatic (pressure and temperature build-up in the dry well and the suppression chamber) as well as dynamic loads (free-blowing process, water throw-up and condensing oscillations) in the course of a loss-of-coolant accident. For determination of the state variations with time in the individual pressure sections thermodynamic equilibrium is assumed between steam and water phases. Thermal non-equilibrium states are taken into account if phase separation interfaces between water and steam/air mixture exist. The flows between the individual pressure sections are treated as homogeneous, nonsteady, incompressible flows. For verification of the mathematical model recalculations were made of experiments performed at various test stands. Teh recalculations showed that the mathematical model has got a wide range of application and is suited for design and assessment of pressure suppression systems at accident conditions. (orig.)

  7. SARNET. Severe Accident Research Network - key issues in the area of source term

    About fifty European organisations integrate in SARNET (Network of Excellence of the EU 6th Framework Programme) their research capacities in resolve better the most important remaining uncertainties and safety issues concerning existing and future Nuclear Power Plants (NPPs) under hypothetical Severe Accident (SA) conditions. Wishing to maintain a long-lasting cooperation, they conduct three types of activities: integrating activities, spreading of excellence and jointly executed research. This paper summarises the main results obtained by the network after the first year, giving more prominence to those from jointly executed research in the Source Term area. Integrating activities have been performed through different means: the ASTEC integral computer code for severe accident transient modelling, through development of PSA2 methodologies, through the setting of a structure for definition of evolving R and D priorities and through the development of a web-network of data bases that hosts experimental data. Such activities have been facilitated by the development of an Advanced Communication Tool. Concerning spreading of excellence, educational courses covering Severe Accident Analysis Methodology and Level 2 PSA have been set up, to be given in early 2006. A detailed text book on Severe Accident Phenomenology has been designed and agreed amongst SARNET members. A mobility programme for students and young researchers is being developed, some detachments are already completed or in progress, and examples are quoted. Jointly executed research activities concern key issues grouped in the Corium, Containment and Source Term areas. In Source Term, behaviour of the highly radio-toxic ruthenium under oxidising conditions (like air ingress) for HBU and MOX fuel has been investigated. First modelling proposals for ASTEC have been made for oxidation of fuel and of ruthenium. Experiments on transport of highly volatile oxide ruthenium species have been performed. Reactor

  8. Behaviour of fission-product iodine under severe accident conditions

    On account of the radiological properties of I-131 the behaviour of fission-product iodine is of great importance under severe reactor accident conditions. The chemical properties of iodine: Its easy conversion into several oxidation compounds, its capability of forming not only volatile (organo-iodide, elemental iodine), hardly volatile, readily soluble (cesium iodide/iodate) but also insoluble (silver iodide) compounds, and its susceptibility to ionizing radiation, are further aspects of significance. Intensive investigations on iodine behaviour under reactor accident conditions carried out worldwide over the last ten years have shown - even though a number of details have yet to be elucidated - that physicochemical processes form a natural, i.e. passive, barrier against the possible release of iodine. (orig.)

  9. The measurement of power reactor stack releases under accident conditions

    The performance of a typical Swedish monitor for ventilation stack radioactivity releases is examined critically with respect to accident generated radioactive particles. The conditions in the stack, particle character, and the monitor design are considered. A large LOCA outside the containment leads to high relative humidity, and high temperature, or mist in the stack. A small external LOCA results in a moderate increase in temperature and humidity, and condensing conditions only with reduced ventilation. Particle size and stickiness are estimated for different types of accident. A particle is sticky if it adheres after contact with a solid, smooth, dry, and clean surface. The monitor performance is concluded to be poor for large, sticky particles, like mist droplets. Dense aerosols, like fire smoke, will plug the sampling filter. Non-sticky particles are generally sampled with acceptable accuracy. (au)

  10. Large-Scale Containment Cooler Performance Experiments under Accident Conditions

    Kapulla, Ralf; Mignot, Guillaume; Paladino, Domenico

    2012-01-01

    Computational Fluid Dynamics codes are increasingly used to simulate containment conditions after various transient accident scenarios. This paper presents validation experiments, conducted in the frame of the OECD/SETH-2 project. These experiments address the combined effects of mass sources and heat sinks related to gas mixing and hydrogen transport within containment compartments. A wall jet interacts with an operating containment cooler located in the middle (M-configuration) and the top ...

  11. Dynamic response of MARS reactor under design basis accident conditions

    The 600 MWth MARS (Multipurpose Advanced Reactor Inherently Safe) one single loop reactor for electric power and/or industrial heat generation relies on a totally inherent and passive safety concept. The key issue of residual heat evacuation in case of accident is solved through a completely passive Emergency Core Cooling System (ECCS) which consists of two independent circuits based on natural circulation triggered by passive check valves activated by the primary pump trip. In principle such a scheme for the decay heat removal system provides for an infinite cooling capability and no man intervention is required. In case of accident the ECCS is activated by the primary pump trip and after a first transient phase, the natural forcing head assures natural convection in the ECCS. The accident analysis related to those design basis events such as Station Blackout, Steam Line Break and Steam Generator Tube Rupture, demonstrates that thanks to its inherent and passive safety features, the reactor is always correctly cooled within the required safety limits. The results evidentiate that the ECCS intervenes in a relatively short time and provides adequate coolant flow rates so that no damage to the fuel and core structures is to be expected. Even in the residual case of lack of both air condensers in the ECCS, the about 100 hours grace period' provided by the water reserve stored in the pool, reasonably allows for undertaking the most appropriate countermeasures. (author)

  12. On the removal of airborne particulate radioactivity under accident conditions

    In the case of an accident, the filter elements in the ventilation systems of a nuclear facility may become a part of the remaining fission product barrier. Within the framework of the Project Nuclear Safety of the Karlsruhe Nuclear Research Center, contributions are made to an increase in reliability of the air cleaning systems under accident conditions. These include the development and verification of computer programs for the estimation of those conditions prevailing inside the air cleaning systems in the case of an accident. Experimental investigations into the response of HEPA filters to differential pressures involving both dry and moist air have demonstrated the occurence of structural failures with subsequent loss of efficiency at relatively low values of differential pressures. With regard to further investigations, a new test facility was put into operation for the realization of superimposed challenges. A new method for testing particulate removal efficiency under high temperature or high humidity was developed. Finally, first results of code development work and of the corresponding verification experiments are reported on. (orig.)

  13. Predictions of structural integrity of steam generator tubes under normal operating, accident, an severe accident conditions

    Majumdar, S. [Argonne National Lab., IL (United States)

    1997-02-01

    Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating, accident, and severe accident conditions are reviewed. Tests conducted in the past, though limited, tended to show that the earlier flow-stress model for part-through-wall axial cracks overestimated the damaging influence of deep cracks. This observation was confirmed by further tests at high temperatures, as well as by finite-element analysis. A modified correlation for deep cracks can correct this shortcoming of the model. Recent tests have shown that lateral restraint can significantly increase the failure pressure of tubes with unsymmetrical circumferential cracks. This observation was confirmed by finite-element analysis. The rate-independent flow stress models that are successful at low temperatures cannot predict the rate-sensitive failure behavior of steam generator tubes at high temperatures. Therefore, a creep rupture model for predicting failure was developed and validated by tests under various temperature and pressure loadings that can occur during postulated severe accidents.

  14. Noble gas control room accident filtration system for severe accident conditions (N-CRAFT)

    Severe accidents might cause the release of airborne radioactive substances to the environment of the NPP either due to containment leakages or due to intentional filtered containment venting. In the latter case aerosols and iodine are retained, however noble gases are not retainable by the FCVS or by conventional air filtration systems like HEPA filters and iodine absorbers. Radioactive noble gases nevertheless dominate the activity release depending on the venting procedure and the weather conditions. To prevent unacceptable contamination of the control room atmosphere by noble gases, AREVA GmbH has developed a noble gas control room accident filtration system (CRAFT) which can supply purified fresh air to the control room without time limitation. The retention process is based on dynamic adsorption of noble gases on activated carbon. The system consists of delay lines (carbon columns) which are operated by a continuous and simultaneous adsorption and desorption process. CRAFT allows minimization of the dose rate inside the control room and ensures low radiation exposure to the staff by maintaining the control room environment suitable for prolonged occupancy throughout the duration of the accident. CRAFT consists of a proven modular design either transportable or permanently installed. (author)

  15. Noble gas control room accident filtration system for severe accident conditions N-CRAFT. System design

    Severe accidents might cause the release of airborne radioactive substances to the environment of the NPP. This can either be due to leakages of the containment or due to a filtered containment venting in order to ensure the overall integrity of the containment. During the containment venting process aerosols and iodine can be retained by the FCVS which prevents long term ground contamination. Noble gases are not retainable by the FCVS. From this it follows that a large amount of radioactive noble gases (e.g. xenon, krypton) might be present in the nearby environment of the plant dominating the activity release, depending on the venting procedure and the weather conditions. Accident management measures are necessary in case of severe accidents and the prolonged stay of staff inside the main control room (MCR) or emergency response center (ERC) is essential. Therefore, the in leakage and contamination of the MRC and ERC with airborne activity has to be prevented. The radiation exposure of the crises team needs to be minimized. The entrance of noble gases cannot be sufficiently prevented by the conventional air filtration systems such as HEPA filters and iodine absorbers. With the objective to prevent an unacceptable contamination of the MCR/ERC atmosphere by noble gases AREVA GmbH has developed a noble gas retention system. The noble gas control room accident filtration system CRAFT is designed for this case and provides supply of fresh air to the MCR/ERC without time limitation. The retention process of the system is based on the dynamic adsorption of noble gases on activated carbon. The system consists of delay lines (carbon columns) which are operated by a continuous and simultaneous adsorption and desorption process. These cycles ensure a periodic load and flushing of the delay lines retaining the noble gases from entering the MCR. CRAFT allows a minimization of the dose rate inside MCR/ERC and ensures a low radiation exposure to the staff on shift maintaining

  16. Flaw tolerance of steam generator tubes under accident conditions

    Tests were carried out with a bank of tubes in a water tunnel to determine the tolerance of flawed nuclear reactor steam generator tubes to accident conditions which could result in high cross-flow velocities. Fifteen specimen tubes in all were tested, each having one of five types of circumferential slots machined into the outside wall near one end. The tubes were tested at flow velocities sufficient to induce high fluidelastic-type vibrations. All tubes were tested to failure, either until a leak occurred or to complete severance. Failure surfaces were characterized after testing. (author). 5 refs., 2 tabs., 5 figs

  17. Fission product release from irradiated LWR fuel under accident conditions

    Fission product release from irradiated LWR fuel is being studied by heating fuel rod segments in flowing steam and an inert carrier gas to simulate accident conditions. Fuels with a range of irradiation histories are being subjected to several steam flow rates over a wide range of temperatures. Fission product release during each test is measured by gamma spectroscopy and by detailed examination of the collection apparatus after the test has been completed. These release results are complemented by a detailed posttest examination of samples of the fuel rod segment. Results of release measurements and fuel rod characterizations for tests at 1400 through 20000C are presented in this paper

  18. Investigation of air cleaning system response to accident conditions

    Air cleaning system response to the stress of accident conditions are being investigated. A program overview and hghlight recent results of our investigation are presented. The program includes both analytical and experimental investigations. Computer codes for predicting effects of tornados, explosions, fires, and material transport are described. The test facilities used to obtain supportive experimental data to define structural integrity and confinement effectiveness of ventilation system components are described. Examples of experimental results for code verification, blower response to tornado transients, and filter response to tornado and explosion transients are reported

  19. Investigation of air cleaning system response to accident conditions

    We are investigating air cleaning system response to the stress of accident conditions. In this paper we present a program overview and highlight recent results of our investigations. The program includes both analytical and experimental investigations. Computer codes for predicting effects of tornados, explosions, fires, and material transport are described. We also describe the test facilities we use to obtain supportive experimental data to define structural integrity and confinement effectiveness of ventilation system components. Examples of experimental results for code verification, blower response to tornado transients, and filter response to tornado and explosion transients are reported

  20. Supervision of operating conditions and retrospection of accident conditions for HTR-10

    The author summarizes the design methods that use the digital control system (DCS) as supervision of operating conditions and retrospection of accident conditions for 10 MW high temperature gas-cooled reactor (HTR-10). It involves the configuration design of historical databases, accidental retrospection databases, human-computer interfaces

  1. Numerical Study of Severe Accidents on Containment Venting Conditions

    Lee, Na Rae; Bang, Young Suk; Park, Tong Kyu; Lee, Doo Yong [FNC Technology Co., Yongin (Korea, Republic of); Choi, Yu Jung; Lee, Sang Won; Kim, Hyeong Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    Under severe accident, the containment integrity can be challenged due to over-pressurization by steam and non-condensable gas generation. According to Seismic Probabilistic Safety Assessment (PSA) result, the late containment failure by over-pressurization has been identified as the most probable containment failure mode. In addition, the analyses of Fukushima nuclear power plant accident reveal the necessity of the proper containment depressurization to prevent the large release of the radionuclide to environment. Containment venting has been considered as an effective approach to maintain the containment integrity from over-pressurization. Basic idea of containment venting is to relieve the pressure inside of the containment by establishing a flow path to the external environment. To ensure the containment integrity under over-pressure conditions, it is crucial to conduct the containment vent in a timely manner with a sufficient discharge flow rate. It is also important to optimize the vent line size to prevent additional risk of leakage and to install at the site with limited space availability. The purpose of this study is to identify the effective venting conditions for preventing the containment over-pressurization and investigate the vent flow characteristics to minimize the consequence of the containment ventilation.. In order that, thermodynamic behavior of the containment and the discharged flow depending on different vent strategies are analyzed and compared. The representative accident scenarios are identified by reviewing the Level 2 PSA result and the sensitivity analyses with varying conditions (i.e. vent line size and vent initiation pressure) are conducted. MAAP5 model for the OPR1000 Korea nuclear power plant has been used for severe accident simulations. Containment venting can be an effective strategy to prevent the significant failure of the containment due to over-pressurization. However, it should be carefully conducted because the vented

  2. Numerical Study of Severe Accidents on Containment Venting Conditions

    Under severe accident, the containment integrity can be challenged due to over-pressurization by steam and non-condensable gas generation. According to Seismic Probabilistic Safety Assessment (PSA) result, the late containment failure by over-pressurization has been identified as the most probable containment failure mode. In addition, the analyses of Fukushima nuclear power plant accident reveal the necessity of the proper containment depressurization to prevent the large release of the radionuclide to environment. Containment venting has been considered as an effective approach to maintain the containment integrity from over-pressurization. Basic idea of containment venting is to relieve the pressure inside of the containment by establishing a flow path to the external environment. To ensure the containment integrity under over-pressure conditions, it is crucial to conduct the containment vent in a timely manner with a sufficient discharge flow rate. It is also important to optimize the vent line size to prevent additional risk of leakage and to install at the site with limited space availability. The purpose of this study is to identify the effective venting conditions for preventing the containment over-pressurization and investigate the vent flow characteristics to minimize the consequence of the containment ventilation.. In order that, thermodynamic behavior of the containment and the discharged flow depending on different vent strategies are analyzed and compared. The representative accident scenarios are identified by reviewing the Level 2 PSA result and the sensitivity analyses with varying conditions (i.e. vent line size and vent initiation pressure) are conducted. MAAP5 model for the OPR1000 Korea nuclear power plant has been used for severe accident simulations. Containment venting can be an effective strategy to prevent the significant failure of the containment due to over-pressurization. However, it should be carefully conducted because the vented

  3. Behaviour of gas cooled reactor fuel under accident conditions

    The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

  4. Identification of traffic accident risk-prone areas under low-light conditions

    Ivan, K.; I. HAIDU; J. BENEDEK; S. M. Ciobanu

    2015-01-01

    Besides other non-behavioural factors, low-light conditions significantly influence the frequency of traffic accidents in an urban environment. This paper intends to identify the impact of low-light conditions on traffic accidents in the city of Cluj-Napoca, Romania. The dependence degree between light and the number of traffic accidents was analysed using the Pearson correlation, and the relation between the spatial distribution of traffic accidents and the light conditio...

  5. Hypothetical accident conditions thermal analysis of the 5320 package

    An axisymmetric model of the 5320 package was created to perform hypothetical accident conditions (HAC) thermal calculations. The analyses assume the 5320 package contains 359 grams of plutonium-238 (203 Watts) in the form of an oxide powder at a minimum density of 2.4 g/cc or at a maximum density of 11.2 g/cc. The solution from a non-solar 100 F ambient steady-state analysis was used as the initial conditions for the fire transient. A 30 minute 1,475 F fire transient followed by cooling via natural convection and thermal radiation to a 100 F non-solar environment was analyzed to determine peak component temperatures and vessel pressures. The 5320 package was considered to be horizontally suspended within the fire during the entire transient

  6. Behaviour of organic iodides under pwr accident conditions

    Laboratory experiments were performed to study the behaviour of radioactive methyl iodide under PWR loss-of-coolant conditions. The pressure relief equipment consisted of an autoclave for simulating the primary circuit and of an expansion vessel for simulating the conditions after a rupture in the reactor coolant system. After pressure relief, the composition of the CH3sup(127/131)I-containing steam-air mixture within the expansion vessel was analysed at 80 0C over a period of 42 days. On the basis of the values measured and of data taken from the literature, both qualitative and quantitative assessments have been made as to the behaviour of radioactive methyl iodide in the event of loss-of-coolant accidents. (author)

  7. The behaviour of radioactive waste packages under fire accident conditions

    An experimental study has been made of the behaviour of packaged Intermediate Level Wastes (ILW) subjected to heat. The conditions used represented fire accidents in the transport of the ILW to the repository in shielded transport containers and in the handling of the packages at the repository. The behaviour of four waste materials immobilised in cement and organic resin were studied. Each waste used had features which allowed the results to be applied to a wide range of other waste streams. Samples of these materials have been heated under controlled and well instrumented conditions in furnaces and pool fires. Inactive simulant wastes were used in small and full scale experiments. Fully active waste materials were used in small scale experiments only. Data are presented on the temperature profiles through the packaged ILW and on the release of volatile and particulate materials as a function of time and temperature. (orig.)

  8. Methyl Iodide Formation Under Postulated Nuclear Reactor Accident Conditions

    The formation of methyl iodide under conditions of postulated nuclear reactor accidents is discussed. Although thermodynamic calculations indicate the equilibrium methyl iodide concentrations would be quite low, calculations based on a simple kinetic scheme involving reaction between small hydrocarbon species and iodine indicate that concentrations higher than equilibrium can occur during the course of the reaction. Such calculations were performed over a wide range of initial species concentrations and a range of temperatures representative of some reactor accident situations. These calculations suggest that little methyl iodide would be expected within the core volume where temperatures are maximum. As the gas leaves the core volume and expands into the plenum region, it cools and the concentration of methyl iodide increases. At the intermediate temperatures which might characterize this region, the formation of methyl iodide from thermally induced reactions could reach its maximum rate. The gas continues to cool, however, and it is probable that by the time it leaves the plenum region it has cooled to the point where thermally induced reactions may be of little importance. Although the thermally induced reactions will become slower as the gas expands and cools, the radiation-induced reactions will not be slowed to the same extent. The gases leaving the core carry fission products and hence a radiation source is available to initiate reaction by a temperature-independent process. An investigation of the radiation chemical formation and decomposition of methyl iodide in the presence of steam suggests that radiation-induced methyl iodide formation will generally be rapid under the postulated accident situations. Thus, in the plenum region where thermal reactions have become slow, the radiation-induced reaction can still proceed and may well become the dominant factor. The same situation probably pertains as well to the containment region. (author)

  9. ACR-1000® end-temperature peaking analysis under postulated accident conditions

    This paper presents a novel and systematic approach to conduct end-temperature peaking analysis under accident conditions for an ACR-1000 reactor, using a two-dimensional (radial and axial) finite-element computer code FEAT. In the past, end-flux peaking effects were overly conservatively assessed by including power increase in the fuel end region without accounting for heat transfer enhancement due to flow disturbance at the bundle end region, especially at the down-stream of a bundle junction. The current analysis determines the end-flux-peaking induced increase in fuel sheath and fuel centreline temperatures while accounting for all relevant key phenomena such as end-flux peaking and heat transfer characteristics including the effects of flow/thermal boundary layer redeveloping at the bundle end region. Using this method significantly reduces the fuel sheath temperature increase caused by end-flux peaking in comparison with the conservative analysis. The postulated accident events considered in this analysis include large break loss-of-coolant accident (LOCA), small break LOCA, and pressure tube rupture within an intact calandria tube. The determined temperature increases relative to the case without end-flux peaking are required to be quantitatively included in detailed safety analyses for postulated accidents. (author)

  10. On requirements to environment protection under accident conditions

    Accident situation on nuclear power plant operation is considered. Definition is given of the concept of ''Accident situation'' and recommendations are made for sequence of evaluation of such a situation. Population protection measures at an accident situation are considered depended on the level of radiation hazard. Recommendations are made for functions of accident team emergency evaluation of radiation hazard in the case of accident and recommendations on composition of equipment for mobile field dosimetric groups are also done. Requirements are given for emergency measures plan for nuclear power plant and criterions for radiation hazard estimation

  11. Failure of fretted steam generator tubes under accident conditions

    Tests were carried out with a bank of tubes in a water tunnel to determine the tolerance of flawed nuclear reactor steam generator tubes to accident conditions which would result in high cross-flow velocities. Fourteen specimen tubes were tested, each having one or two types of defect machined into the surface simulating fretting-wear type scars found in some operating steam generators. The tubes were tested at flow velocities sufficient to induce high fluid elastic-type vibrations. Seven of the tubes failed near the thinnest section of the defects during the one-hour tests, due to impacting and/or rubbing between the tube and the support. Strain gauges, displacement transducers, force gauges and an accelerometer were used on the target tube and/or the tube immediately downstream of it to measure their vibrational characteristics

  12. ORNL studies of fission product release under LWR accident conditions

    High burnup Zircaloy-clad UO2 fuel specimens have been heated to study the release of fission products in tests simulating LWR accident conditions. The dominant variable was found to be temperature, with atmosphere, time, and burnup also being significant variables. Comparison of data from tests in steam and hydrogen, at temperatures of 2000 to 2700 K, have shown that the releases of the most volatile species (Kr, Xe, I, and Cs) are relatively insensitive to atmosphere. The releases of the less-volatile species (Sr, Mo, Ru, Sb, Te, Ba, and Eu), however, may vary by orders of magnitude depending on atmosphere. In addition, the atmosphere may drastically affect the mode and extent of fuel destruction

  13. Ruthenium behaviour in severe nuclear accident conditions. Final report

    During routine nuclear reactor operations, ruthenium will accumulate in the fuel in relatively high concentrations. In a steam atmosphere, ruthenium is not volatile, and it is not likely to be released from the fuel. However, in an air ingress accident during reactor power operation or during maintenance, ruthenium may form volatile species, which may be released into the containment. Oxide forms of ruthenium are more volatile than the metallic form. Radiotoxicity of ruthenium is high both in the short and the long term. The results of this project imply that in oxidising conditions during nuclear reactor core degradation, ruthenium release increases as oxidised gaseous species Ru03 and Ru04 are formed. A significant part of the released ruthenium is then deposited on reactor coolant system piping. However, in the presence of steam and aerosol particles, a substantial amount of ruthenium may be released as gaseous Ru04 into the containment atmosphere. (au)

  14. Heat Transfer in Cane Fiberboard Exposed to Hypothetical Accident Conditions

    Gromada, R.J.

    1995-05-25

    Radioactive material packages containing fiberboard insulation have been subjected to Hypothetical Accident Condition (HAC) thermal tests for many years. Historically, the packages` thermal performance has always been difficult to grasp. A package designer needs to understand the effects of temperature and pyrolysis on the rate of heat transfer and performance. This paper describes in detail the one-dimensional HAC thermal tests performed on fiberboard to understand the effects of pyrolysis, its char and its gas products. The tests were conducted by the Packaging and Transportation Group at the Savannah River Site (SRS). Test fixtures were assembled at SRS and thermal testing conducted in the Radiant Heat Facility at the Sandia National Laboratories. Descriptions of the test fixtures are provided, as well as the time dependent temperature profiles. In addition, lessons learned are discussed.

  15. Ruthenium behaviour in severe nuclear accident conditions. Final report

    Backman, U.; Lipponen, M.; Auvinen, A.; Jokiniemi, J.; Zilliacus, R. [VVT Processes (Finland)

    2004-08-01

    During routine nuclear reactor operations, ruthenium will accumulate in the fuel in relatively high concentrations. In a steam atmosphere, ruthenium is not volatile, and it is not likely to be released from the fuel. However, in an air ingress accident during reactor power operation or during maintenance, ruthenium may form volatile species, which may be released into the containment. Oxide forms of ruthenium are more volatile than the metallic form. Radiotoxicity of ruthenium is high both in the short and the long term. The results of this project imply that in oxidising conditions during nuclear reactor core degradation, ruthenium release increases as oxidised gaseous species Ru03 and Ru04 are formed. A significant part of the released ruthenium is then deposited on reactor coolant system piping. However, in the presence of steam and aerosol particles, a substantial amount of ruthenium may be released as gaseous Ru04 into the containment atmosphere. (au)

  16. Measurement of steam condensation on aerosols und LWR accident conditions

    The report summarizes the results of experiments on steam condensation onto aerosol particles. A facility was constructed which allows the direct measurement of the condensation processes. The thermodynamic boundary conditions were typical for a core melt accident. Different aerosol species were used, especially uranium dioxide which constitutes a large fraction of the core melt aerosol. As a general result the condensation process in supersaturated atmospheres causes a drastic change in the shape of the aerosol particles. Originally fluffy chain-like aggregates are compressed to nearly spherical dense particles. A significant simplification of the NAUA-model can be used because the commonly encountered shape factor problems become non-existent. This also leads to a greater reliability of computed results. (orig./HP)

  17. A risk-based evaluation of the impact of key uncertainties on the prediction of severe accident source terms - STU

    The purpose of this project is to address the key uncertainties associated with a number of fission product release and transport phenomena in a wider context and to assess their relevance to key severe accident sequences. This project is a wide-based analysis involving eight reactor designs that are representative of the reactors currently operating in the European Union (EU). In total, 20 accident sequences covering a wide range of conditions have been chosen to provide the basis for sensitivity studies. The appraisal is achieved through a systematic risk-based framework developed within this project. Specifically, this is a quantitative interpretation of the sensitivity calculations on the basis of 'significance indicators', applied above defined threshold values. These threshold values represent a good surrogate for 'large release', which is defined in a number of EU countries. In addition, the results are placed in the context of in-containment source term limits, for advanced light water reactor designs, as defined by international guidelines. Overall, despite the phenomenological uncertainties, the predicted source terms (both into the containment, and subsequently, into the environment) do not display a high degree of sensitivity to the individual fission product issues addressed in this project. This is due, mainly, to the substantial capacity for the attenuation of airborne fission products by the designed safety provisions and the natural fission product retention mechanisms within the containment

  18. Comparison of selected U.S. highway and railway severe accidents to U.S. regulatory accident conditions and IAEA transport standards

    This paper discusses selected severe historical US highway and rail accidents and compares the mechanical and/or thermal environments associated with these accidents to the 10CFR71 Hypothetical Accident Conditions and the accident environments (both regulatory and extraregulatory) investigated in 'Shipping Container Response to Severe Highway and Railway Accident Conditions', which is commonly known as the Modal Study, and in 'Re-examination of Spent Fuel Shipment Risk Estimates', NUREG/CR-6672. Since the hypothetical accident conditions of 10CFR71 are similar to the International Atomic Energy Agency's (IAEA) package tests for accident conditions of transport, the evaluation is also valid in demonstrating the adequacy of IAEA's transport safety standard. Careful examination of the reports on the severe accidents revealed the accidents were found to be bounded by the regulatory environment described in 10CFR71. (author)

  19. Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO2–Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  20. Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

    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

  1. Identification of traffic accident risk-prone areas under low lighting conditions

    K. Ivan; I. HAIDU; J. BENEDEK; Ciobanu, S. M.

    2015-01-01

    Besides other non-behavioural factors, the low lighting conditions significantly influence the frequency of the traffic accidents in the urban environment. This paper intends to identify the impact of low lighting conditions on the traffic accidents in the city of Cluj-Napoca. The dependence degree between lighting and the number of traffic accidents was analyzed by the Pearson's correlation and the relation between the spatial distributio...

  2. Investigation of Key Factors for Accident Severity at Railroad Grade Crossings by Using a Logit Model

    Hu, Shou-Ren; Li, Chin-Shang; Lee, Chi-Kang

    2010-01-01

    Although several studies have used logit or probit models and their variants to fit data of accident severity on roadway segments, few have investigated accident severity at a railroad grade crossing (RGC). Compared to accident risk analysis in terms of accident frequency and severity of a highway system, investigation of the factors contributing to traffic accidents at an RGC may be more complicated because of additional highway–railway interactions. Because the proportional odds assumption ...

  3. Shipping container response to severe highway and railway accident conditions: Main report

    This report describes a study performed by the Lawrence Livermore National Laboratory to evaluate the level of safety provided under severe accident conditions during the shipment of spent fuel from nuclear power reactors. The evaluation is performed using data from real accident histories and using representative truck and rail cask models that likely meet 10 CFR 71 regulations. The responses of the representative casks are calculated for structural and thermal loads generated by severe highway and railway accident conditions. The cask responses are compared with those responses calculated for the 10 CFR 71 hypothetical accident conditions. By comparing the responses it is determined that most highway and railway accident conditions fall within the 10 CFR 71 hypothetical accident conditions. For those accidents that have higher responses, the probabilities anf potential radiation exposures of the accidents are compared with those identified by the assessments made in the ''Final Environmental Statement on the Transportation of Radioactive Material by Air and other Modes,'' NUREG-0170. Based on this comparison, it is concluded that the radiological risks from spent fuel under severe highway and railway accident conditions as derived in this study are less than risks previously estimated in the NUREG-0170 document

  4. Large-Scale Containment Cooler Performance Experiments under Accident Conditions

    Computational Fluid Dynamics codes are increasingly used to simulate containment conditions after various transient accident scenarios. This paper presents validation experiments, conducted in the frame of the OECD/SETH-2 project. These experiments address the combined effects of mass sources and heat sinks related to gas mixing and hydrogen transport within containment compartments. A wall jet interacts with an operating containment cooler located in the middle (M-configuration) and the top (T-configuration) of the containment vessel. The experiments are characterized by a 3-phase injection scenario. In Phase I, pure steam is injected, while in Phase II, a helium-steam mixture is injected. Finally, in Phase III, pure steam is injected again. Results for the M-configuration show helium stratification build up during Phase II. During Phase III, a positively buoyant plume emerging from the cooler housing becomes negatively buoyant once it reaches the helium-steam layer and continuously erodes the layer. For the M-configuration, a strong degradation of the cooler performance was observed during the injection of the helium/steam mixture (Phase II). For the T-configuration, we observe a mainly downwards acting cooler resulting in a combination of forced and natural convection flow patterns. The cooler performance degradation was much weaker compared with the M-configuration and a good mixing was ensured by the operation of the cooler.

  5. Fission product release from fuel under LWR accident conditions

    Three tests have provided additional data on fission product release under LWR accident conditions in a temperature range (1400 to 20000C). In the release rate data are compared with curves from a recent NRC-sponsored review of available fission product release data. Although the iodine release in test HI-3 was inexplicably low, the other data points for Kr, I, and Cs fall reasonably close to the corresponding curve, thereby tending to verify the NRC review. The limited data for antimony and silver release fall below the curves. Results of spark source mass spectrometric analyses were in agreement with the gamma spectrometric results. Nonradioactive fission products such as Rb and Br appeared to behave like their chemical analogs Cs and I. Results suggest that Te, Ag, Sn, and Sb are released from the fuel in elemental form. Analysis of the cesium and iodine profiles in the thermal gradient tube indicates that iodine was deposited as CsT along with some other less volatile cesium compound. The cesium profiles and chemical reactivity indicate the presence of more than one cesium species

  6. Should evacuation conditions after a nuclear accident be revised?

    The author proposes to draw lessons from the Fukushima accident, notably in the field of post-accident management. He discusses the definition of an as widely understandable as possible method of description of risks related to irradiations after a nuclear accident. As these irradiations are mainly low dose ones which have a carcinogenic effect, he proposes to assess the average life expectancy loss due to an irradiation. Then, this risk can be easily compared with other risks like air pollution, smoking and passive smoking, and so on. Then, once this risk assessment method is well defined, it is possible to associate the inhabitants of contaminated areas to the post-accident management. They could then decide to go back to their homes or not with full knowledge of the facts

  7. Identification of traffic accident risk-prone areas under low-light conditions

    Ivan, K.; Haidu, I.; Benedek, J.; Ciobanu, S. M.

    2015-09-01

    Besides other non-behavioural factors, low-light conditions significantly influence the frequency of traffic accidents in an urban environment. This paper intends to identify the impact of low-light conditions on traffic accidents in the city of Cluj-Napoca, Romania. The dependence degree between light and the number of traffic accidents was analysed using the Pearson correlation, and the relation between the spatial distribution of traffic accidents and the light conditions was determined by the frequency ratio model. The vulnerable areas within the city were identified based on the calculation of the injury rate for the 0.5 km2 areas uniformly distributed within the study area. The results show a strong linear correlation between the low-light conditions and the number of traffic accidents in terms of three seasonal variations and a high probability of traffic accident occurrence under the above-mentioned conditions at the city entrances/exits, which represent vulnerable areas within the study area. Knowing the linear dependence and the spatial relation between the low light and the number of traffic accidents, as well as the consequences induced by their occurrence, enabled us to identify the areas of high traffic accident risk in Cluj-Napoca.

  8. Iodine chemistry and associated interactions under severe accident conditions

    In a highly improbable severe accident wherein the core cooling is decapacitated or insufficient the scenario may lead to melting of fuel elements and fission products release. Nuclear power plants are designed with inherent engineering safety systems and associated operational procedures that provide an in-depth defence against such accidents. Iodine, one of the fission products, behaviour is required for the analysis of severe accident consequences because iodine is a chemical more active to the potential source term for release to the environment. During severe accident, Iodine is released and transported in aqueous, organic and inorganic forms. Iodine release from fuel, iodine transport in primary coolant system, containment, and reaction with control rods are some of the important phases in a severe accident scenario. The behaviour of iodine-bearing particles is governed by aerosol physics, depletion mechanisms gravitational settling, diffusiophoresis and thermophoresis. Sorption and desorption of iodine occurring on containment surface are also of importance. The presence of gaseous organic compounds and oxidizing compounds on iodine, reactions of aerosol iodine with boron and formation of cesium iodide which results in more volatile iodine release in containment plays significant roles. Water radiolysis products due to presence of dissolved impurities such as dissolved oxygen, nitrate/nitrite (NO3/NO2) produced by air radiolysis, trace metal ions such as Fe2+/Fe3+ dissolved from steel surfaces, chloride ions coming from the pyrolysis/radiolysis of polyvinyl material from cables and organic impurities from painted surfaces and polymers also inherent and should be considered while calculating iodine release. This paper elaborates stare of art on iodine chemistry and its behaviour during accident. (author)

  9. Off-gas and air cleaning systems for accident conditions in nuclear power plants

    This report surveys the design principles and strategies for mitigating the consequences of abnormal events in nuclear power plants by the use of air cleaning systems. Equipment intended for use in design basis accident and severe accident conditions is reviewed, with reference to designs used in IAEA Member States. 93 refs, 48 figs, 23 tabs

  10. Some conditions affecting the definition of design basis accidents relating to sodium/water reactions

    The possible damaging effects of large sodium/water reactions on the steam generator, IHX and secondary circuit are considered. The conditions to be considered in defining the design basis accidents for these components are discussed, together with some of the assumptions that may be associated with design assessments of the scale of the accidents. (author)

  11. Robot dispatching Scenario for Accident Condition Monitoring of NPP

    In March of 2011, unanticipated big size of tsunami attacks Fukushima NPP, this accident results in explosion of containment building. Tokyo electric power of Japan couldn't dispatch a robot for monitoring of containment inside. USA Packbot robot used for desert war in Iraq was supplied to Fukushima NPP for monitoring of high radiation area. Packbot also couldn't reach deep inside of Fukushima NPP due to short length of power cable. Japanese robot 'Queens' also failed to complete a mission due to communication problem between robot and operator. I think major reason of these robot failures is absence of robot dispatching scenario. If there was a scenario and a rehearsal for monitoring during or after accident, these unanticipated obstacles could be overcome. Robot dispatching scenario studied for accident of nuclear power plant was described herein. Study on scenario of robot dispatching is performed. Flying robot is regarded as good choice for accident monitoring. Walking robot with arm equipped is good for emergency valve close. Short time work and shift work by several robots can be a solution for high radiation area. Thin and soft cable with rolling reel can be a good solution for long time work and good communication

  12. Thermal Analysis of a H1616-1 Shipping container in Hypothetical Accident conditions

    The thermal response of the H1616 transport container is simulated to demonstrate compliance with the Federal regulations for performance during hypothetical accident conditions (HAC). The goal is to show that tests conducted for the certification of the H1616 shipping container provide conservatively high estimates of temperatures at key regions within the container. A one-dimensional computational model is developed to simulate the thermal response of the shipping container in cylindrical coordinates. The model assumes the container is axisymmetric and allows for variable thermal properties. The model is calibrated using temperature data obtained FR-om two experimental thermal tests and is then used to evaluate the thermal response of the shipping container to several different scenarios that meet or exceed the Federal regulations. A pre-heating technique, which is used to simulate the thermal effects of a radioactive heat source within the container, is also evaluated

  13. A radioactive waste transportation package monitoring system for normal transport and accident emergency response conditions

    Shipments of radioactive material (RAM) constitute but a small fraction of the total hazardous materials shipped in the United States each year. Public perception, however, of the potential consequences of a release from a transportation package containing RAM has resulted in significant regulation of transport operations, both to ensure the integrity of a package in accident conditions and to place operational constraints on the shipper. Much of this attention has focused on shipments of spent nuclear fuel and high level wastes which, although comprising a very small number of total shipments, constitute a majority of the total curies transported on an annual basis. Shipment of these highly radioactive materials is made in what is described in the regulations as a Type B packaging. Type B transportation packages are designed to withstand a sequence of accident scenarios, including drop, puncture, fire, and immersion with virtually no release of contents. Due to the quantities of spent fuel and high level wastes carried in Type B casks and the public perception and apprehension regarding the potential consequences of a release, involvement of a packaging containing spent fuel or high level wastes in any accident will result in a very cautious emergency response until it can be determined that the integrity of the cask is maintained. Typically this involves closure of the transport link or pathway, evacuation of all unnecessary personnel, diversion of traffic from the area, and subsequent investigative and mitigative procedures from trained specialists. An onboard instrumentation/communications package has been developed that, when affixed to a radioactive materials cask, can monitor key indicators of the integrity of the cask and communicate these parameters to emergency responders through modules on the vehicle. Entitled the Transportation Intelligent Monitoring System (TRANSIMS), this package enables remote monitoring of the status and integrity of the cask

  14. Numerical module for debris behavior under severe accident conditions

    The late phase of a hypothetical severe accident in a nuclear reactor is characterized by the appearance of porous debris and liquid pools in core region and lower head of the reactor vessel. Thermal hydraulics and heat transfer in these regions are very important for adequate analysis of severe accident dynamics. The purpose of this work is to develop a universal module which is able to model above-mentioned phenomena on the basis of modern physical concepts. The original approach for debris evolution is developed from classical principles using a set of parameters including debris porosity; average particle diameter; temperatures and mass fractions of solid, liquid and gas phases; specific interface areas between different phases; effective thermal conductivity of each phase, including radiative heat conductivity; mass and energy fluxes through the interfaces. The calculation results of several tests on modeling of porous debris behavior, including the MP-1 experiment, are presented in comparison with experimental data. The results are obtained using this module implemented into the Russian best estimate code, RATEG/SVECHA/HEFEST, which was developed for modeling severe accident thermal hydraulics and late phase phenomena in VVER nuclear power plants. (author)

  15. The influence of seasonal conditions on the radiological consequences of a nuclear accident

    The impact of an accidental release of radioactivity to the environment can be strongly influenced by prevailing environmental conditions. Thus, potential variations in accident consequences caused by variable seasonal, meteorological or climatic conditions are of significance to the development and application of protective measures and emergency response plans. These proceedings present the results of a workshop organized by the NEA to examine such aspects of emergency response to a nuclear accident

  16. The CRP-6 benchmark on HTGR fuel behavior under accident conditions

    National engagement as well as bilateral or multi-national cooperation in HTGR fuel development is ongoing and is expected to further improve fuel performance and the ability to make reliable predictions. The accident condition benchmark exercise, one of the key elements within the sixth IAEA-directed Coordinated Research Project (CRP) on 'Advances in HTGR Fuel Technology Development', has successfully demonstrated to be a useful basis for verification and validation in establishing the reliability of code predictions. Participants in the accident condition benchmark included France, Germany, Russia, South Africa, Korea, and the United States applying a total of eight models to all or a part of the 24 proposed benchmark cases. The benchmark consisted of three parts, a sensitivity study to examine fission product release from a fuel particle, the postcalculation of well documented irradiation and heating experiments, and finally some predictive calculations. In the sensitivity study, most codes have shown good agreement among each other. Differences can be explained by different assumptions for input data or boundary conditions. In comparison with the numerical procedure of the diffusion calculation for the kernel, the application of the analytical solution offered by the Booth model appears to be more accurate method. Time step length may also influence the calculational results. From the postcalculations of heating tests, it appears that the diffusion coefficient for cesium in silicon carbide is still varying over a broad range. In particular, strontium release data are obviously largely overpredicted and should undergo a thorough review. Silver release measurement results are often unexpected and inconsistent, and therefore extremely difficult for postcalculation. One of the most recent heating experiments, HFR-K6/3, has shown surprisingly low krypton and cesium release values, which are largely overpredicted by the model calculations. This extremely good

  17. A review of iodine chemistry under severe accident conditions

    This report reviews the progress that has been made in establishing a basic understanding of the factors which will determine the behaviour of iodine during postulated accidents in water-cooled reactors. The topics considered are thermal reactions, radiolytic reactions, impurity effects, organic iodide formation, integral models and tests and volatility control. There have been substantial gains in a number of areas, most notably in the kinetics and thermodynamics databases for thermal and radiolytic reactions of inorganic iodine in solution. However, there remains a limited understanding of the mechanisms controlling the formation of organic iodides and a need for integral tests of iodine behaviour in complex, 'dirty' systems to provide data for the validation of chemical models which are undergoing development. 81 refs

  18. Potential behavior of depleted uranium penetrators under shipping and bulk storage accident conditions

    An investigation of the potential hazard from airborne releases of depleted uranium (DU) from the Army's M829 munitions was conducted at the Pacific Northwest Laboratory. The study included: (1) assessing the characteristics of DU oxide from an April 1983 burn test, (2) postulating conditions of specific accident situations, and (3) reviewing laboratory and theoretical studies of oxidation and airborne transport of DU from accidents. Results of the experimental measurements of the DU oxides were combined with atmospheric transport models and lung and kidney exposure data to help establish reasonable exclusion boundaries to protect personnel and the public at an accident site. 121 references, 44 figures, 30 tables

  19. Key factors contributing to accident severity rate in construction industry in Iran: a regression modelling approach.

    Soltanzadeh, Ahmad; Mohammadfam, Iraj; Moghimbeigi, Abbas; Ghiasvand, Reza

    2016-03-01

    Construction industry involves the highest risk of occupational accidents and bodily injuries, which range from mild to very severe. The aim of this cross-sectional study was to identify the factors associated with accident severity rate (ASR) in the largest Iranian construction companies based on data about 500 occupational accidents recorded from 2009 to 2013. We also gathered data on safety and health risk management and training systems. Data were analysed using Pearson's chi-squared coefficient and multiple regression analysis. Median ASR (and the interquartile range) was 107.50 (57.24- 381.25). Fourteen of the 24 studied factors stood out as most affecting construction accident severity (psafety and health risk management system to reduce ASR. PMID:27092639

  20. Prediction of temperature and fission product release from HTR fuel under accident conditions

    Modern, small High-Temperature Reactors (HTRs) are designed such that maximum accident fuel temperatures remain below 1600degC without active control mechanisms. It has been demonstrated that HTR fuel remains intact and retains all fission products under these maximum accident conditions at least as well as under normal operating conditions. The accident temperature limit has been achieved by a core design with small thermal power and low power density. In the case of a loss-of-coolant accident (LOCA), the decay heat is removed from the core by passive means. The passive core temperature limitation has been demonstrated with a series of LOCA simulation tests with the AVR pebble-bed HTR in Julich, Germany. Here, the maximum core temperatures were measured to be 1080degC in agreement with predictions and, being used for code validation, in agreement with post-test calculations. (J.P.N.)

  1. Behaviour of LWR core materials under accident conditions. Proceedings of a technical committee meeting

    At the invitation of the Government of the Russian Federation, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA convened a Technical Committee Meeting on Behaviour of LWR Core Materials Under Accident Conditions from 9 to 13 October 1995 in Dimitrovgrad to analyze and evaluate the behaviour of LWR core materials under accident conditions with special emphasis on severe accidents. In-vessel severe accidents phenomena were considered in detail, but specialized thermal hydraulic aspects as well as ex-vessel phenomena were outside the scope of the meeting. Forty participants representing eight countries attended the meeting. Twenty-three papers were presented and discussed during five sessions. Refs, figs, tabs

  2. Failure strains and proposed limit strains for an reactor pressure vessel under severe accident conditions

    The local failure strains of essential design elements of a reactor vessel are investigated. The size influence of the structure is of special interest. Typical severe accident conditions including elevated temperatures and dynamic loads are considered. The main part of work consists of test families with specimens under uniaxial and biaxial load. Within one test family the specimen geometry and the load conditions are similar, but the size is varied up to reactor dimensions. Special attention is given to geometries with a hole or a notch causing non-uniform stress and strain distributions typical for the reactor vessel. A key problem is to determine the local failure strain. Here suitable methods had to be developed including the so-called 'vanishing gap method', and the 'forging die method'. They are based on post-test geometrical measurements of the fracture surfaces and reconstructions of the related strain fields using finite element models. The results indicate that stresses versus dimensionless deformations are approximately size independent up to failure for specimens of similar geometry under similar load conditions. Local failure strains could be determined. The values are rather high and size dependent. Statistical evaluation allow the proposal of limit strains which are also size dependent. If these limit strains are not exceeded, the structures will not fracture

  3. Experiences in methods to involve key players in planning protective actions in the case of a nuclear accident

    A widely used method in the planning of protective actions is to establish a stakeholder network to generate a comprehensive set of generic protective actions. The aim is to increase competence and build links for communication and coordination. The approach of this work was to systematically evaluate protective action strategies in the case of a nuclear accident. This was done in a way that the concerns and issues of all key players could be transparently and equally included in the decision taken. An approach called Facilitated Decision Analysis Workshop has been developed and tested. The work builds on case studies in which it was assumed that a hypothetical accident had led to a release of considerable amounts of radionuclides and, therefore, various types of countermeasures had to be considered. Six workshops were organised in the Nordic countries where the key players were represented, i.e. authorities, expert organisations, industry and agricultural producers. (authors)

  4. Key concepts and history of radiation protection and safety focusing on the ICRP publications. For the status after the accident of the TEPCO Fukushima Daiichi NPP

    After the accident of the Fukushima Daiichi NPP, new regulations and guidelines for radiation protection and waste management were provided by the government office concerned and local autonomous bodies. These were decided based on the recommendation of the International Commission on Radiological Protection (ICRP) and adapted to apply the recommendation to the present conditions. The concept of ICRP has been accepted in the regulations and guidelines of many countries. In this paper, we review the history of the ICRP and key points for the emergency and existing situations described in the ICRP publication 103 (2007 ICRP recommendation) and ICRP Publication 111. (author)

  5. Spherical steel containments of pressurized water reactors under accident conditions

    The investigations will deal with the mechanical behavior of a free standing spherical containment shell built for the latest type of a German pressurized water reactor. The diameter of the containment shell is 56 m. The wall thickness is 38 mm. The material used is the ferritic steel 15MnNi63. The investigation program includes theoretical as well as experimental activities and concerns four different accidents which are beyond the scope of the common design and licensing practice: containment behavior under quasi-static pressure increase up to containment failure; containment behavior under high transient pressures; containment vibrations due to earthquake loadings (consideration of shell imperfections); containment buckling due to earthquake loadings. First results concerning the containment behavior under quasi-static pressure increase are presented. It turns out that the mechanical failure of the containment shell is controlled by plastic instability. A computer program to describe this problem has been developed and membrane tests to check the computational methods have been carried out. (orig.)

  6. Monitoring of personal doses under normal and accident conditions

    Organization, legal structure and results of the control of personal doses in Lithuania are provided. Legal basis for monitoring of personal doses in Lithuania are the Basic Standards of Radiation Protection HN 73-1997 which are in force since January 1 1998. The General Order of Dosimetric Control in the Case of Radiation Accident has been approved by the Government on 12 May 1998. Dosimetric control consist of registration of equivalent dose rate and exposure rate, measurements of surface contamination, personal and group control of doses of persons taking part in remedial actions and members of public, radiometric control of foodstuff, drinking water and other samples. The Radiation Protection Centre of the Ministry of Health and Joint Research Center of the Ministry of Environment are responsible for organization, coordination and control of personal dosimetry according to their competence. Data on licensees and sources of ionizing radiation, results of monitoring of personal doses in medicine and Ignalina NPP during year 1995 - 1997 are presented as well

  7. The behaviour of spherical HTR fuel elements under accident conditions

    Hypothetical accidents may lead to significantly higher temperatures in HTR fuel than during normal operation. In order to obtain meaningful statements on fission product behaviour and release, irradiated spherical fuel elements containing a large number of coated particles (20,000-40,000) with burnups between 6 and 16% FIMA were heated at temperatures between 1400 and 2500 deg. C. HTI-pyrocarbon coating retains the gaseous fission products (e.g. Kr) very well up to about 2400 deg. C if the burnup does not exceed the specified value for THTR (11.5%). Cs diffuses through the pyrocarbon significantly faster than Kr and the diffusion is enhanced at higher fuel burnups because of irradiation induced kernel microstructure changes. Below about 1800 deg. C the Cs release rate is controlled by diffusion in the fuel kernel; above this temperature the diffusion in the pyrocarbon coating is the controlling parameter. An additional SiC coating interlayer (TRISO) ensures Cs retention up to 1600 deg. C. However, the release obtained in the examined fuel elements was only by a factor of three lower than through the HTI pyrocarbon. Solid fission products added to UO2-TRISO particles to simulate high burnup behave in various ways and migrate to attack the SiC coating. Pd migrates fastest and changes the SiC microstructure making it permeable

  8. Estimation of spray system efficiency in case of loss in coolant severe accident condition

    The results of pressurize surge line double ended break accident analysis in case of failure of ECCS at Armenian NPP are presented. Based on the analysis results the assessment of spray system efficiency on decreasing confinement pressure and amount radioactive material is carried out. Hydrogen behavior in confinement is analyzed. The occurrence of conditions for possible hydrogen burning in the confinement is assessed as well. Likelihood of accident is in the range of 10-7. However for accident analysis purposes of such kind of accidents needs to be taken into account. The analysis shows that the main contributor in release decrease is spray system availability factor. Unavailability of spray system could lead to the increase of radioactive release by factor 8

  9. Inherent safety features of the HTTR revealed in the accident condition

    The High Temperature Engineering Test Reactor (HTTR) being constructed by JAERI (Japan Atomic Energy Research Institute) is a graphite-moderated and helium-cooled reactor with an outlet gas temperature of 950degC. The inherent safety characteristics in the HTTR prevent temperature increase of reactor fuels and fission product release from the reactor core in postulated accident conditions. The reactor core can be cooled by a Vessel Cooling System (VCS) indirectly, even in the case that no forced cooling is expected during the accident such as primary pipe break. The VCS consists of independent water cooling loop and cooling panel around the reactor pressure vessel. The cooling panel whose temperature of 60-90degC cools the reactor pressure vessel by radiation and removes the decay heat from the core indirectly. Furthermore, even if failure of VCS is assumed during this accident as a severe accident, the reactor core is remained safe despite the temperature increase of biological concrete shield around the reactor pressure vessel. This paper describes the inherent safety features of the HTTR specially focused on the accident condition without forced cooling. The detailed analytical results of such an accident are described together with clarifying the role of the VCS. (author)

  10. Fuel Behaviour and Modelling under Severe Transient and Loss of Coolant Accident (LOCA) Conditions. Proceedings of a Technical Meeting

    the art of the performance of nuclear fuel for water cooled reactors under severe transients and LOCA conditions. The meeting was attended by 83 specialists representing fuel vendors, nuclear utilities, research and development institutions, and regulatory authorities from 19 Member States. The papers submitted to the meeting were organized into seven sessions covering analytical and experimental RIA and LOCA studies and international programmes, power ramp, and severe accident analysis. These proceedings contain all the papers that were presented and discussed during the meeting, and highlight key findings and recommendations based on the summaries of the session chairpersons. While the Fukushima Daiichi accident influenced the discussions, it was not directly considered because of the lack of fuel behaviour data available at the time of the technical meeting

  11. Computational analysis of the behaviour of nuclear fuel under steady state, transient and accident conditions

    Accident analysis is an important tool for ensuring the adequacy and efficiency of the provision in the defence in depth concept to cope with challenges to plant safety. Accident analysis is the milestone of the demonstration that the plant is capable of meeting any prescribed limits for radioactive releases and any other acceptable limits for the safe operation of the plant. It is used, by designers, utilities and regulators, in a number of applications such as: (a) licensing of new plants, (b) modification of existing plants, (c) analysis of operational events, (d) development, improvement or justification of the plant operational limits and conditions, and (e) safety cases. According to the defence in depth concept, the fuel rod cladding constitutes the first containment barrier of the fission products. Therefore, related safety objectives and associated criteria are defined, in order to ensure, at least for normal operation and anticipated transients, the integrity of the cladding, and for accident conditions, acceptable radiological consequences with regard to the postulated frequency of the accident, as usually identified in the safety analysis reports. Therefore, computational analysis of fuel behaviour under steady state, transient and accident conditions constitutes a major link of the safety case in order to justify the design and the safety of the fuel assemblies, as far as all relevant phenomena are correctly addressed and modelled. This publication complements the IAEA Safety Report on Accident Analysis for Nuclear Power Plants (Safety Report Series No. 23) that provides practical guidance for establishing a set of conceptual and formal methods and practices for performing accident analysis. Computational analysis of the behaviour of nuclear fuel under transient and accident conditions, including normal operation (e.g. power ramp rates) is developed in this publication. For design basis accidents, depending on the type of influence on a fuel element

  12. Preliminary Analysis of Radiation Shielding for HIC Transport Package Under the Hypothetical Accident Conditions

    A radiation shielding analysis under the hypothetical accident condition has been conducted using a computer program MCNP5 for a B-type HIC (High Integrated Container) Transport Package, which contains HIC with radioactive waste or spent resin, for transportation from nuclear power plat sites to disposal repository. Radiation source term is first carefully determined from the safety analysis reports related to HIC for appropriate calculation. And then MCNP5 is performed to obtain the minimum crevice between package lid and body, which meets the dose rate limit under the hypothetical accident conditions. Standards and codes of radiation shielding analysis related to the hypothetical accident condition are prescribed in Korea Nuclear Law, IAEA Safety Standards Series for Radioactive Material Transport and US 10CFR Part 71

  13. Instrument Fault Detection Sensitivity of an Empirical Model under Accident Condition in NPPs

    After the recent accident in Fukushima, Japan, it has been proven that we cannot obtain fully reliable information from instruments during severe accident conditions. Although the reactor core really melted down, the RV water level indicator showed a more optimistic value than the actual conditions. Accordingly, plant operators were under the misunderstanding that the core was not exposed. This caused confusion for the incident response. Therefore, it is necessary to be equipped with a function that informs operators of the status of the instrument integrity in real time. If plant operators verify that the instruments are working properly during accident conditions, they able to make safer decisions. In an effort to solve this problem, we considered an empirical model using a Process Equipment Monitoring (PEM) tool as a method of instrument diagnosis in a nuclear power plant

  14. Experience of past radiation accidents and problems of response to possible dispersion of radioactive materials in urban conditions

    The report studies into key problems associated with direct and indirect consequences of possible radiological terrorist acts committed in urban conditions. Much attention is paid to the analysis of lessons learned from elimination of consequences of past radiation accidents in the territory of the former USSR, especially as regards radioactive contamination of cities. The report contains recommendations on necessary improvements of instrumentation, methodological, legal and organizational bases of managerial decision-making to reduce a probability of radiological terrorism acts and minimize their direct and indirect consequences. (author)

  15. Behavior of HEPA filter systems under accident conditions

    With respect to the behavior of HEPA filters under high humidity conditions, emphasis was placed on the study of the differential pressure increase. Under fog conditions, the differential pressure of dust loaded filters increased within minutes, up to values sufficient to damage normal commercial filters units. The investigation into the failure mechanisms was completed with the development of an equation to calculate filter medium tensile stresses for two of the three most important modes of failure. Initial work was begun toward the development of a computer code to model transient fluiddynamic and thermodynamic conditions in complex air cleaning systems. Further investigation into the transmission of weak shock waves in air-cleaning system ductwork were carried out in branches of ducts with square cross-sections and with variable cross-sections. (orig./DG)

  16. Research progress on assessment of reactor vessel integrity under severe accident conditions

    As a representative method of reactor vessel integrity (RVI) under severe accident conditions, In-vessel retention of molten core debris (IVR) is an important severe accident management strategy employed in the AP1000 generation-3 Pressurized Water Reactor. In this paper, research progress on the test and theoretical analysis based on RVI is reviewed. Test facilities and techniques, as well as the modeling are summarized. In addition, tools for numerical simulation for RVI are evaluated. Finally, based on the applications in thermal hydraulic technology for the generation-3 Pressurized Water Reactor in China, the potential research direction of thermal-hydraulics under RVI conditions are discussed. (authors)

  17. Fuel behavior under loss-of-coolant-accident conditions

    The paper is a comprehensive summary of the main results of the KfK/PNS investigations on LWR fuel behavior under LOCA conditions. These investigations were started in 1973 and will be finished in 1983. It is shown that the dominant phenomena, such as the deformation and failure of the cladding, the high temperature steam oxidation, the interaction of the cladding with fuel and fission products, and the influence of thermohydraulics on the cladding deformation are well understood today. All results confirm that under LOCA conditions the coolability of the core is not questioned and the fission product release is well below license limits. (orig.)

  18. Determination of Optimal Flow Paths for Safety Injection According to Accident Conditions

    Yoo, Kwae Hwan; Kim, Ju Hyun; Kim, Dong Yeong; Na, Man Gyun [Chosun Univ., Gwangju (Korea, Republic of); Hur, Seop; Kim, Changhwoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In case severe accidents happen, major safety parameters of nuclear reactors are rapidly changed. Therefore, operators are unable to respond appropriately. This situation causes the human error of operators that led to serious accidents at Chernobyl. In this study, we aimed to develop an algorithm that can be used to select the optimal flow path for cold shutdown in serious accidents, and to recover an NPP quickly and efficiently from the severe accidents. In order to select the optimal flow path, we applied a Dijkstra algorithm. The Dijkstra algorithm is used to find the path of minimum total length between two given nodes and needs a weight (or length) matrix. In this study, the weight between nodes was calculated from frictional and minor losses inside pipes. That is, the optimal flow path is found so that the pressure drop between a starting node (water source) and a destination node (position that cooling water is injected) is minimized. In case a severe accident has happened, if we inject cooling water through the optimized flow path, then the nuclear reactor will be safely and effectively returned into the cold shutdown state. In this study, we have analyzed the optimal flow paths for safety injection as a preliminary study for developing an accident recovery system. After analyzing the optimal flow path using the Dijkstra algorithm, and the optimal flow paths were selected by calculating the head loss according to path conditions.

  19. Experiences in methods to involve key players in planning protective actions in a case of nuclear accident

    Full text: Openness, transparency and key players participation are all important for balanced decision making in public issues. The emergency exercises involve commonly representatives from various sectors of the society to increase competence and build links for communication and coordination. A different approach has been a set of meetings where the key players aimed to plan comprehensive set of generic protective actions. The approach of this work was to develop methods and techniques to evaluate systematically and comprehensively protective action strategies. This was done in a way that all key players' concerns and issues related to decisions on protective actions could be aggregated openly and equally. We have developed and tested an approach called facilitated workshop based an theory of decision analysis. The work builds on case studies in which it was assumed that a hypothetical accident at a nuclear power plant had led to a release of considerable amounts of radionuclides and therefore various types of protective actions should be considered. Altogether six workshops were organized where all key players were represented, i.e., authorities, expert organizations, industry and producers. The participants were those who are responsible for preparing advice or making presentations of matters to those responsible for formal decision-making. Many preparatory meetings were held with various experts. It was seen essential that the setup followed strictly the decision-making process the participants are accustomed with. The realistic nature and the disciplined process of a facilitated workshop, and committed to decision-making yielded insight on what information should be collected or studied. Information should be in the proper form needed in decision-making. For example, the study revealed the need to further develop methods to assess the radiological and cost implications of different countermeasures realistically. In order to provide consequence assessments

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

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

  1. Study on oxidation behavior of cladding for accident conditions in spent fuel pool

    In order to clarify the air oxidation behavior of the cladding at high temperatures for study on improvement of safety for accident conditions in spent fuel pool, the oxidation tests for both small specimens under constant temperature conditions and long specimens under loss of coolant simulated temperature conditions were carried out, and the knowledge for influence of both temperature gradient and preoxide film on oxidation behavior of the cladding were obtained in this study. (author)

  2. Investigation of Focusing Effect according to the Cooling Condition and Height of the Metallic layer in a Severe Accident

    Moon, Je-Young; Chung, Bum-Jin [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    The Fukushima nuclear power plant accident has led to renewed research interests in severe accidents of nuclear power plants. In-Vessel Retention (IVR) of core melt is one of key severe accident management strategies adopted in nuclear power plant design. The metallic layer is heated from below by the radioactive decay heat generated at the oxide pool, and is cooled from above and side walls. During the IVR process, reactor vessel may be cooled externally (ERVC) and the heat fluxes to the side wall increase with larger temperature difference than above. This {sup F}ocusing effect{sup i}s varied by cooling condition of upper boundary and height of the metallic layer. A sulfuric acid–copper sulfate (H{sub 2}SO{sub 4} - CuSO{sub 4}) electroplating system was adopted as the mass transfer system. Numerical analysis using the commercial CFD program FLUENT 6.3 were carried out with the same material properties and cooling conditions to examine the variation of the cell. The experimental and numerical studies were performed to investigate the focusing effect according to cooling condition of upper boundary and the height in metallic layer. The height of the side wall was varied for three different cooling conditions: top only, side only, and both top and side. Mass transfer experiments, based on the analogy concept, were carried out in order to achieve high Rayleigh number. The experimental results agreed well with the Rayleigh-Benard convection correlations of Dropkin and Somerscales and Globe and Dropkin. The heat transfer on side wall cooling condition without top cooling is highest and was enhanced by decreasing the aspect ratio. The numerical results agreed well with the experimental results. Each cell pattern (cell size, cell direction, central location of cell) differed in the cooling condition. Therefore, it is difficult to predict the internal flow due to complexity of cell formation behavior.

  3. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  4. Sleep disorders, medical conditions, and road accident risk.

    Smolensky, Michael H; Di Milia, Lee; Ohayon, Maurice M; Philip, Pierre

    2011-03-01

    Sleep disorders and various common acute and chronic medical conditions directly or indirectly affect the quality and quantity of one's sleep or otherwise cause excessive daytime fatigue. This article reviews the potential contribution of several prevalent medical conditions - allergic rhinitis, asthma, chronic obstructive pulmonary disease, rheumatoid arthritis/osteoarthritis - and chronic fatigue syndrome and clinical sleep disorders - insomnia, obstructive sleep apnea, narcolepsy, periodic limb movement of sleep, and restless legs syndrome - to the risk for drowsy-driving road crashes. It also explores the literature on the cost-benefit of preventive interventions, using obstructive sleep apnea as an example. Although numerous investigations have addressed the impact of sleep and medical disorders on quality of life, few have specifically addressed their potential deleterious effect on driving performance and road incidents. Moreover, since past studies have focused on the survivors of driver crashes, they may be biased. Representative population-based prospective multidisciplinary studies are urgently required to clarify the role of the fatigue associated with common ailments and medications on traffic crash risk of both commercial and non-commercial drivers and to comprehensively assess the cost-effectiveness of intervention strategies. PMID:21130215

  5. Thermal and hydraulic behaviour of CANDU cores under severe accident conditions

    This report summarizes the results of a study of the thermo-hydraulic behavior of CANDU cores under accident conditions more severe than those normally considered in the licensing process. A comprehensive description and complete results of the study are given in the main report

  6. PRESSURE INTEGRITY OF 3013 CONTAINER UNDER POSTULATED ACCIDENT CONDITIONS

    Rawls, G.

    2010-02-01

    A series of tests was carried out to determine the threshold for deflagration-to-detonation transition (DDT), structural loading, and structural response of the Department of Energy 3013 storage systems for the case of an accidental explosion of evolved gas within the storage containers. Three experimental fixtures were used to examine the various issues and three mixtures consisting of either stoichiometric hydrogen-oxygen, stoichiometric hydrogen-oxygen with added nitrogen, or stoichiometric hydrogen-oxygen with an added nitrogen-helium mixture were tested. Tests were carried out as a function of initial pressure from 1 to 3.5 bar and initial temperature from room temperature to 150 C. The elevated temperature tests resulted in a slight increase in the threshold pressure for DDT. The elevated temperature tests were performed to ensure the test results were bounding. Because the change was not significant the elevated temperature data are not presented in the paper. The explosions were initiated with either a small spark or a hot surface. Based on the results of these tests under the conditions investigated, it can be concluded that DDT of a stoichiometric hydrogen-oxygen mixture (and mixtures diluted with nitrogen and helium) within the 3013 containment system does not pose a threat to the structural integrity of the outer container.

  7. Study on protective layer for severe accident conditions for EC6 reactor vault structure

    The Enhanced CANDU 6 (EC6) is designed both for the prevention and mitigation of Design Basis Accidents (DBAs) as well as Beyond Design Basis Accidents (BDBAs). The foremost objective, in accordance with the safety goals specified in the CNSC Regulatory Document (RD-337), is to prevent the occurrence of any accident that could jeopardize nuclear safety, and, if an accident should occur, to limit the radiological releases resulting from the accident and minimize the impact on nearby communities. During a postulated severe core accident, Molten Core-Concrete Interaction (MCCI) may occur when molten core debris breaches the calandria vessel and contacts concrete surfaces, whereby the thermal and chemical properties of the melt contribute to the potential degradation of the concrete. The earliest phase of MCCI is characterized by very-high-temperature molten metal and oxide pouring from the calandria vessel and settling as a pool on the concrete surfaces of the vault floor. The molten material can result in spalling or fragmentation of the concrete near where the corium first contacts the concrete. As the corium settles on the concrete surface, the melt begins to react chemically with the concrete through the penetrating cracks and fragments produced on the initial contact, generating various gases including carbon monoxide and combustible hydrogen. In order to control and mitigate MCCI, a protective layer (refractory material) with suitable material properties and sufficient thickness was proposed to protect the reactor vault concrete floor. To further enhance vault floor protection and mitigate the conditions under severe accidents a special concrete composition in the upper layer of the vault floor concrete is to be provided in case the refractory material is breached. This special concrete should minimize the generation of various gases including combustible hydrogen and carbon monoxide during MCCI. As a part of research and development program an experimental

  8. Study of light water reactor containments under important severe accident conditions

    The US Nuclear Regulatory Commission has sponsored studies to develop a ''LEAKAGE-BEFORE-FAILURE'' model for use in severe accident risk assessments to provide a means of accounting for significant containment leakage prior to reaching the containment threshold pressure. Six containment types have been studied (large dry, subatmospheric, ice condenser, Mark I, II, and III). Potential leak paths through major containment penetration assemblies were investigated and upper-bound estimates of leak areas established. These leak areas may result from increasing internal pressure and degradation of nonmetallic seal materials due to severe accident conditions. This paper describes the approach and summarizes the results and conclusions of this study

  9. Retention of elemental 131I by activated carbons under accident conditions

    Under simulated accident conditions (maximum temperature: 1300C) no significant difference was found in the retention of I-131 loaded as elemental iodine, by various fresh and aged commercial activated carbons. In all the cases, the I-131 passing through deep beds of activated carbon was in a non-elemental form. It is concluded that a minimum retention of 99.99% for elemental radioiodine, as required by the RSK guidelines for PWR accident filters, can be equally well achieved with various commercial activated carbons. (orig.)

  10. Study on safety evaluation for nuclear fuel cycle facility under accident conditions

    Source term data for estimating release behavior of radioactive nuclides is necessary to evaluate synthetic safety of nuclear fuel cycle facility under accident conditions, such as fire and criticality. In JAERI, the data has been obtained by performing some demonstration tests. In this paper, the data for the criticality accident in fuel solution obtained from the TRACY experiment, will be mainly reviewed. At 4.5 h after the transient criticality, the release ratio of the iodine were about 0.2% for re-insertion of transient rod at just after transient criticality and about 0.9% for not re-insertion. Similarly the release coefficient and release ratio for Xe were estimated. It was proved that the release ratio of Xe-141 from the solution was over 90% in case that the inverse period was over about 100 1/s. Furthermore, outline of the study on the fire accident as future plan will be also mentioned. (author)

  11. analysis of reactivity accidents in MTR for various protection system parameters and core condition

    Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99Mo production. The old core comprising 29 fuel elements and one Co Irradiation Device (CID) and the new core comprising 27 fuel elements, CID, and two 99Mo production boxes. The in core irradiation has the advantage of no special cooling or irradiation loop is required. The purpose of the present work is the analysis of reactivity accidents (RIA) for ETRR-2 cores. The analysis was done to evaluate the accidents from different point of view:1- Analysis of the new core for various Reactor Protection System (RPS) parameters 2- Comparison between the two cores. 3- Analysis of the 99Mo production boxes.PARET computer code was employed to compute various parameters. Initiating events in RIA involve various modes of reactivity insertion, namely, prompt critical condition (p=1$), accidental ejection of partial and complete CID uncontrolled withdrawal of a control rod accident, and sudden cooling of the reactor core. The time histories of reactor power, energy released, and the maximum fuel, clad and coolant temperatures of fuel elements and LEU plates were calculated for each of these accidents. The results show that the maximum clad temperatures remain well below the clad melting of both fuel and uranium plates during these accidents. It is concluded that for the new core, the RIA with scram will not result in fuel or uranium plate failure.

  12. Criticality safety evaluation of spent fuel storage rack under accident condition using soluble boron credit

    A boraflex attached on a consolidated storage spent fuel rack as neutron absorber has a characteristic that silica and boron carbide(B4C) present in the boraflex are dissolved into pent fuel pool water due to the long term irradiation of boraflex by spent fuels. in this report it is analyzed how in a case of complete dissolution of boron from the boraflex into the pool water, the adapted cresit of the dissolved boron affects on the criticality of storage spent fules to compensate an excessive reactivity due to postulated accidents. For criticality analyses PHOENIX-P and SCALE4.4 were used and benchmark calculations were carried out to verify the bias and uncertainties of the codes. The result of criticality analyses for postulated accident conditions shows that most of postulated accident such as spent fuel drop did not cause reactivity to increase significantly. However, the most severe accident to increase reactivity was a postulated abnormal loading of spent fuel under checkerboard loading pattern and the maximum required soluble boron concentration to compensate the increased reactivity in this case was 698.45ppm. The soluble boron concentration to make up the uncertainty from the burnup calculation and measurement of the spent duels was 116.65ppm so that the total required soluble boron concentration for compensation of the increased reactivity due to the most severe accident could be taken 815.10ppm by arithmetic addition of 698.45 and 116.65 ppm. It can be concluded that 2,300ppm minimum soluble requirement in technology specification of spent fuel storage pool operation of Ulchin NPP No. 2 is large enough to maintain sub-critical of the spent fuel storage pool under all of postulated accidents conditions

  13. Behaviour of HTGR coated particles and fuel elements under normal and accident conditions

    Main results of testing HTGR coated particles and spheric fuel elements developed in Scientific and Industrial Association ''Lutch'' under conditions of higher level of energy release and temperature than those designed are given in the report. The summarized data on tightness and characteristic defects change, on gas and solid fission products release under model accident conditions before, during and after radiation are presented. (author). 6 refs, 9 figs, 1 tab

  14. Tests of the carbon steel containment coating systems under design basis accident conditions

    During a Design Basis Accident (DBA) in nuclear power plants, conditions in the reactor containment will be characterized by elevated temperature and pressure. The Thermohydraulic Laboratory of CDTN have done tests, for evaluating protective coating systems test specimens for the steel containment, under simulated DBA conditions. Until this moment were tested 60 specimens of 6 coating systems. This paper presents the test installation, the tests performed and the temperature pattern specified for Angra II Power Plant. (author). 3 refs, 8 figs

  15. EVALUATION OF TRAFFIC ACCIDENT CHARACTERISTICS ASSOCIATED TO WEATHER CONDITIONS IN BOTUCATU, SP

    Sergio Augusto Rodrigues

    2015-12-01

    Full Text Available Uncontrolled growth of several cities has generated major problems regarding urban environment and mobility. Among several factors affecting mobility in Brazilian cities and towns traffic accidents are increasingly common concerns and climate condition might be a possible generator for such growth. In addition there is increasing pollution and possible changes in the environment generated by the large increase in the number of vehicles in circulation. The objective of this study was to evaluate the association between traffic accidents and weather conditions in Botucatu, SP, Brazil. It was used database with information obtained from the city´s responsible bodies for traffic accidents as well as from the meteorological station. These data were analyzed using univariate statistical procedures. Charts and tables were presented for a better understanding of the behavior of each variable. Later it was used linear correlation coefficient for understanding how climate characteristics of the city are associated to some information related to traffic accidents. It was observed that there were significant correlations between some of the variables.

  16. Assessment of potential doses to workers during postulated accident conditions at the Waste Isolation Pilot Plant

    Hoover, M.D.; Farrell, R.F. [DOE, Carlsbad, NM (United States); Newton, G.J.

    1995-12-01

    The recent 1995 WIPP Safety Analysis Report (SAR) Update provided detailed analyses of potential radiation doses to members of the public at the site boundary during postulated accident scenarios at the U.S. Department of Energy`s Waste Isolation Pilot Plant (WIPP). The SAR Update addressed the complete spectrum of potential accidents associated with handling and emplacing transuranic waste at WIPP, including damage to waste drums from fires, punctures, drops, and other disruptions. The report focused on the adequacy of the multiple layers of safety practice ({open_quotes}defense-in-depth{close_quotes}) at WIPP, which are designed to (1) reduce the likelihood of accidents and (2) limit the consequences of those accidents. The safeguards which contribute to defense-in-depth at WIPP include a substantial array of inherent design features, engineered controls, and administrative procedures. The SAR Update confirmed that the defense-in-depth at WIPP is adequate to assure the protection of the public and environment. As a supplement to the 1995 SAR Update, we have conducted additional analyses to confirm that these controls will also provide adequate protection to workers at the WIPP. The approaches and results of the worker dose assessment are summarized here. In conformance with the guidance of DOE Standard 3009-94, we emphasize that use of these evaluation guidelines is not intended to imply that these numbers constitute acceptable limits for worker exposures under accident conditions. However, in conjunction with the extensive safety assessment in the 1995 SAR Update, these results indicate that the Carlsbad Area Office strategy for the assessment of hazards and accidents assures the protection of workers, members of the public, and the environment.

  17. Heat transport and afterheat removal for gas cooled reactors under accident conditions

    The Co-ordinated Research Project (CRP) on Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions was organized within the framework of the International Working Group on Gas Cooled Reactors (IWGGCR). This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs) and supports the conduct of these activities. Advanced GCR designs currently being developed are predicted to achieve a high degree of safety through reliance on inherent safety features. Such design features should permit the technical demonstration of exceptional public protection with significantly reduced emergency planning requirements. For advanced GCRs, this predicted high degree of safety largely derives from the ability of the ceramic coated fuel particles to retain the fission products under normal and accident conditions, the safe neutron physics behaviour of the core, the chemical stability of the core and the ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. Prior to licensing and commercial deployment of advanced GCRs, these features must first be demonstrated under experimental conditions representing realistic reactor conditions, and the methods used to predict the performance of the fuel and reactor must be validated against these experimental data. Within this CRP, the participants addressed the inherent mechanisms for removal of decay heat from GCRs under accident conditions. The objective of this CRP was to establish sufficient experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advance gas cooled reactors during accidents. The scope includes experimental and analytical investigations of heat transport by natural convection conduction and thermal

  18. Development of advanced claddings for suppressing the hydrogen emission in accident conditions. Development of advanced claddings for suppressing the hydrogen emission in the accident condition

    The development of accident-tolerant fuels can be a breakthrough to help solve the challenge facing nuclear fuels. One of the goals to be reached with accident-tolerant fuels is to reduce the hydrogen emission in the accident condition by improving the high-temperature oxidation resistance of claddings. KAERI launched a new project to develop the accident-tolerant fuel claddings with the primary objective to suppress the hydrogen emission even in severe accident conditions. Two concepts are now being considered as hydrogen-suppressed cladding. In concept 1, the surface modification technique was used to improve the oxidation resistance of Zr claddings. Like in concept 2, the metal-ceramic hybrid cladding which has a ceramic composite layer between the Zr inner layer and the outer surface coating is being developed. The high-temperature steam oxidation behaviour was investigated for several candidate materials for the surface modification of Zr claddings. From the oxidation tests carried out in 1 200 deg. C steam, it was found that the high-temperature steam oxidation resistance of Cr and Si was much higher than that of zircaloy-4. Al3Ti-based alloys also showed extremely low-oxidation rate compared to zircaloy-4. One important part in the surface modification is to develop the surface coating technology where the optimum process needs to be established depending on the surface layer materials. Several candidate materials were coated on the Zr alloy specimens by a laser beam scanning (LBS), a plasma spray (PS) and a PS followed by LBS and subject to the high-temperature steam oxidation test. It was found that Cr and Si coating layers were effective in protecting Zr-alloys from the oxidation. The corrosion behaviour of the candidate materials in normal reactor operation condition such as 360 deg. C water will be investigated after the screening test in the high-temperature steam. The metal-ceramic hybrid cladding consisted of three major parts; a Zr liner, a ceramic

  19. Development of a diagnostic system for identifying accident conditions in a reactor

    This report describes a methodology for identification of accident conditions in a nuclear reactor from the signals available to the operator. A large database of such signals is generated through analyses - for core, containment, environmental dispersion and radiological dose to train a computer code based on an Artificial Neural Networks (ANNs). At present, in the prediction mode, information on LOCA (location and size of break), status of availability of ECCS, and expected doses can be predicted well for a 220 MWe PHWR. (author)

  20. Estimate of radionuclide release characteristics into containment under severe accident conditions

    A detailed review of the available light water reactor source term information is presented as a technical basis for development of updated source terms into the containment under severe accident conditions. Simplified estimates of radionuclide release and transport characteristics are specified for each unique combination of the reactor coolant and containment system combinations. A quantitative uncertainty analysis in the release to the containment using NUREG-1150 methodology is also presented

  1. Estimate of radionuclide release characteristics into containment under severe accident conditions. Final report

    Nourbakhsh, H.P. [Brookhaven National Lab., Upton, NY (United States)

    1993-11-01

    A detailed review of the available light water reactor source term information is presented as a technical basis for development of updated source terms into the containment under severe accident conditions. Simplified estimates of radionuclide release and transport characteristics are specified for each unique combination of the reactor coolant and containment system combinations. A quantitative uncertainty analysis in the release to the containment using NUREG-1150 methodology is also presented.

  2. Most likely vessel lower head failure location during severe accident conditions

    The Nuclear Regulatory Commission is sponsoring a lower vessel head research program to investigate plausible modes of reactor vessel failure to determine: (a) which modes have the greatest likelihood of occurrence during a severe accident and (b) the range of core debris and accident conditions that lead to these failures. All major types of US light water reactor vessels are being considered, and both high- and low-pressure conditions are being addressed for each reactor type. The research program includes analytical and finite element calculations. In addition, high temperature creep and tensile data for predicting vessel structural response were obtained. Calculational results used to predict which failure location is more likely in a particular reactor design during a severe accident are described within this paper. Detailed analyses are being performed to investigate the relative likelihood of a BWR penetration and vessel to fail during a wide range of severe accident conditions. The analyses include applying a numerical model to obtain the penetration and vessel thermal response and applying an analytical model to investigate the relative likelihood of tube rupture and global vessel failure. Sensitivity studies considered the impact of assumptions related to debris composition, debris porosity, corium decay heat, vessel coolant mass, heat removal from the vessel, melt relocation time, and melt relocation distance on vessel and penetration response. In addition, analytically developed failure maps, which were developed in terms of dimensionless groups, are applied to extrapolate numerically-obtained results to geometries and materials occurring in PWR penetration/vessel configurations and a wide range of debris conditions

  3. Study on probability of failure for RPV nozzle region under severe accident condition

    Hwang, Il Soon; Oh, Young Jin; Sim, Sang Hoon [Seoul National University, Seoul (Korea)

    2002-04-01

    Most of previous studies for creep rupture of RPV lower head under severe accident condition, have been focused on global failure of RPV lower head. In contrast, the local failure of the RPV nozzle region has not been studied in detail. This study focused the nozzle failure analysis into creep rupture evaluation of RPV lower head under severe accident condition, and this will help improve the safety assessment of nuclear power plants under severe accident conditions. The existence and features of nozzle failure in LAVA-ICI tested vessel of Korea Atomic Energy Research Institute and LHF-4 tested vessel of SNL, are examined. To understand the basic mechanical properties of nozzle material and weld metal, the tensile tests in various temperature levels and the creep rupture tests in various temperature and stress levels, are accomplished. The stress and deformation of LAVA-ICI experiments are analysed using measured basic mechanical properties. The failure time of Advanced Power Reactor 1400 (APR1400) in nozzle region was calculated using modified TMI-2 VIP model. Nozzle region failure characteristics was studied for SNL-LHF-4 experimental case using Finite Element Method (FEM). Using characteristics of nozzle failure, a new failure prediction experimental method was proposed for RPV nozzle failure. 19 refs., 43 figs. (Author)

  4. TECHNICAL BASIS DOCUMENT FOR THE ABOVE GROUND TANK FAILURE REPRESENTATIVE ACCIDENT and ASSOCIATED REPRESENTED HAZARDOUS CONDITIONS

    This document analyzes aboveground tank failure accident scenarios. The radiological and toxicological consequences are determined for a range of aboveground tank failure accident scenarios to determine the representative accident. Based on the consequence results and accident frequency evaluations, risk bins are determined and control decisions are made. This revision deals with aboveground tank failure accidents during CH-TRUM operations

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

    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)

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

    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)

  7. Identification of the security threshold by logistic regression applied to fuel under accident conditions

    Gomes, Daniel de Souza; Baptista Filho, Benedito; Oliveira, Fabio Branco de, E-mail: dsgomes@ipen.br, E-mail: bdbfilho@ipen.br, E-mail: fabio@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    A reactivity-initiated Accident (RIA) is a disastrous failure, which occurs because of an unexpected rise in the fission rate and reactor power. This sudden increase in the reactor power may activate processes that might lead to the failure of fuel cladding. In severe accidents, a disruption of fuel and core melting can occur. The purpose of the present research is to study the patterns of such accidents using exploratory data analysis techniques. A study based on applied statistics was used for simulations. Then, we chose peak enthalpy, pulse width, burnup, fission gas release, and the oxidation of zirconium as input parameters and set the safety boundary conditions. This new approach includes the logistic regression. With this, the present research aims also to develop the ability to identify the conditions and the probability of failures. Zirconium-based alloys fabricating the cladding of the fuel rod elements with niobium 1% were analyzed for high burnup limits at 65 MWd/kgU. The data based on six decades of investigations from experimental programs. In test, perform in American reactors such as the transient reactor test (TREAT), and power Burst Facility (PBF). In experiments realized in Japanese program at nuclear in the safety research reactor (NSRR), and in Kazakhstan as impulse graphite reactor (IGR). The database obtained from the tests and served as a support for our study. (author)

  8. Identification of the security threshold by logistic regression applied to fuel under accident conditions

    A reactivity-initiated Accident (RIA) is a disastrous failure, which occurs because of an unexpected rise in the fission rate and reactor power. This sudden increase in the reactor power may activate processes that might lead to the failure of fuel cladding. In severe accidents, a disruption of fuel and core melting can occur. The purpose of the present research is to study the patterns of such accidents using exploratory data analysis techniques. A study based on applied statistics was used for simulations. Then, we chose peak enthalpy, pulse width, burnup, fission gas release, and the oxidation of zirconium as input parameters and set the safety boundary conditions. This new approach includes the logistic regression. With this, the present research aims also to develop the ability to identify the conditions and the probability of failures. Zirconium-based alloys fabricating the cladding of the fuel rod elements with niobium 1% were analyzed for high burnup limits at 65 MWd/kgU. The data based on six decades of investigations from experimental programs. In test, perform in American reactors such as the transient reactor test (TREAT), and power Burst Facility (PBF). In experiments realized in Japanese program at nuclear in the safety research reactor (NSRR), and in Kazakhstan as impulse graphite reactor (IGR). The database obtained from the tests and served as a support for our study. (author)

  9. Fission products behaviour in UO2 submitted to nuclear severe accident conditions

    Geiger, E.; Bès, R.; Martin, P.; Pontillon, Y.; Solari, P. L.; Salome, M.

    2016-05-01

    The objective of this work was to study the molybdenum chemistry in UO2 based materials, known as SIMFUELS. These materials could be used as an alternative to irradiated nuclear fuels in the study of fission products behaviour during a nuclear severe accident. UO2 samples doped with 12 stable isotopes of fission products were submitted to annealing tests in conditions representative to intermediate steps of severe accidents. Samples were characterized by SEM-EDS and XAS. It was found that Mo chemistry seems to be more complex than what is normally estimated by thermodynamic calculations: XAS spectra indicate the presence of Mo species such as metallic Mo, MoO2, MoO3 and Cs2MoO4.

  10. a Study of the Interferences with the On-Line Radioiodine Measurement Under Nuclear Accident Conditions

    Tseng, Tung-Tse

    In this research the interferences with the on -line detection of radioiodines, under nuclear accident conditions, were studied. The special tool employed for this research is the developed on-line radioiodine monitor (the Penn State Radioiodine Monitor), which is capable of detecting low levels of radioiodine on-line in air containing orders of magnitude higher levels of radioactive noble gases. Most of the data reported in this thesis were collected during a series of experiments called "Source -Term Experiment Program (STEP)." The experiments were conducted at the Argonne National Laboratory's TREAT reactor located at the Idaho National Engineering Laboratory (INEL). In these tests, fission products were released from the Light Water Reactor (LWR) test fuels as a result of simulating a reactor accident. The Penn State Monitor was then used to sample the fission products accumulated in a large container which simulated the reactor containment building. The test results proved that the Penn State Monitor was not affected significantly by the passage of large amounts of noble gases through the system. Also, it confirmed the predicted results that the operation of conventional on-line radioiodine detectors would, under nuclear accident conditions, be seriously impaired by the passage of high concentrations of radioactive noble gases through such systems. This work also demonstrated that under conditions of high noble gas concentrations and low radioiodine concentrations, the formation of noble-gas-decayed alkali metals can seriously interfere with the on-line detection of radioiodine, especially during the 24 hours immediately after the accident. The decayed alkali metal particulates were also found to be much more penetrating than the ordinary type of particulates, since a large fraction (15%) of the particulates were found to penetrate through the commonly used High Efficiency Particulate Air (HEPA) filter (rated >99.97% for 0.3 (mu)m particulate). Also, a

  11. CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

    This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

  12. Condition monitoring a key component in the preventive maintenance

    The preventive maintenance programs are necessary to ensure that nuclear safety significant equipment will function when it is supposed to. Diesel generator, pumps, motor operated valves and air operated control valves are typically operated every three months. When you drive a car, you depend on lot of sounds, the feel of the steering wheel and gauges to determine if the car is running correctly. Similarly with operating equipment for a power plant - sounds or vibration of the equipment or the gauges and test equipment indicate a problem or degradation, actions are taken to correct the deficiency. Due to safety and economical reason diagnostic and monitoring systems are of growing interest in all complex industrial production. Diagnostic systems are requested to detect, diagnose and localize faulty operating conditions at an early stage in order to prevent severe failures and to enable predictive and condition oriented maintenance. In this context it is a need for using various on-line and off-line condition monitoring and diagnostics, non-destructive inspection techniques and surveillance. The condition monitoring technique used in nuclear power plant Cernavoda are presented in this paper. The selection of components and parameters to be monitored, monitoring and diagnostics techniques used are incorporated into a preventive maintenance program. Modern measurement technique in combination with advanced computerized data processing and acquisition show new ways in the field of machine surveillance. The diagnostic capabilities of predictive maintenance technologies have increased recently year with advances made in sensor technologies. The paper will focus on the following condition monitoring technique: - oil analysis - acoustic leakage monitoring - thermography - valve diagnostics: motor operated valve, air operated valve and check valve - motor current signature - vibration monitoring and rotating machine monitoring and diagnostics For each condition monitoring

  13. Need for Replications of Key Studies in Covert Conditioning.

    Livingston, Roger H.; Elson, Steven E.

    There is a considerable body of research that involves covert antecedents and consequences of behavior, and how these factors tend to influence overt behavior. As is frequently the case in new areas of endeavor, overenthusiastic claims have been made for covert conditioning procedures, often based on poorly controlled experiments or clinical…

  14. Status report of advanced cladding modeling work to assess cladding performance under accident conditions

    B.J. Merrill; Shannon M. Bragg-Sitton

    2013-09-01

    Scoping simulations performed using a severe accident code can be applied to investigate the influence of advanced materials on beyond design basis accident progression and to identify any existing code limitations. In 2012 an effort was initiated to develop a numerical capability for understanding the potential safety advantages that might be realized during severe accident conditions by replacing Zircaloy components in light water reactors (LWRs) with silicon carbide (SiC) components. To this end, a version of the MELCOR code, under development at the Sandia National Laboratories in New Mexico (SNL/NM), was modified by replacing Zircaloy for SiC in the MELCOR reactor core oxidation and material properties routines. The modified version of MELCOR was benchmarked against available experimental data to ensure that present SiC oxidation theory in air and steam were correctly implemented in the code. Additional modifications have been implemented in the code in 2013 to improve the specificity in defining components fabricated from non-standard materials. An overview of these modifications and the status of their implementation are summarized below.

  15. Use of scale models to assess structural response of nuclear shipping containers under accident conditions

    Experimental scale modelling techniques were used to investigate the complex behaviour of truck-type high level waste and spent fuel shipping packaging during severe impact accidents. A series of experiments were conducted with distorted replica scale models fabricated with Type 304 stainless steel and unbonded lead shielding. The models were fabricated to represent typical 1/8, 1/4, and 1/2 linear scaled versions of a full scale prototype unit. Experiments were conducted for 9m (30 ft) free fall accidents onto an essentially unyielding surface with the centre-of-gravity of the model directly over the centre of its bottom plate. The test temperatures were - 40 and 1750C to cover the extreme environmental conditions that this type of packaging may encounter in its normal service life. The inertial loading of the model was controlled during the simulated impact accident by attaching a balsa wood impact limiter to the bottom of the model. Deceleration measurements obtained during the tests were in the range of 1000g. Permanent strain induced in the steel shells was in the range of 0.004 m/m with the largest strain induced at 1750C as expected. Lead slump occurred in all experiments and was in the range of 1 to 3% of the original shielded length. (author)

  16. Analysis 320 coal mine accidents using structural equation modeling with unsafe conditions of the rules and regulations as exogenous variables.

    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." PMID:27085591

  17. Design feasibility study on corium stabilization in bottom end-fitting for AHWR under accident condition

    Advanced Heavy Water Reactor (AHWR) is being designed in a robust way to cater both Design and Beyond Design Basis Accidents to meet all the safety functions. All the functions are met by passive means with special emphasis on 'residual heat removal' which is catered by passive natural circulation mode. In context to Design Basis Accidents, several features are designed to handle worst kind of scenario like Station Black Out. For Design Extension Conditions (DEC), the means of passive natural circulation is adopted as a design means to meet the DEC-A conditions like cooling of moderator by natural circulation means with GDWP inventory. Under the DEC-B condition where large scale of fuel melting is envisaged, a core catcher is designed with active/passive cooling modes to take care of the residual heat of the core. All the mentioned features utilizes the natural mode of heat transfer to meet one of the safety function i.e. 'residual heat removal'. The analysis shows that the tube sheet as well as lattice tube temperatures remain low and are able to take out the heat from corium through sub-cooled nucleate boiling. The ES cooling is sufficient to maintain the cooling water in subcooled condition. The integrity of tube sheet and lattice tube is maintained

  18. Role of Winter Weather Conditions and Slipperiness on Tourists’ Accidents in Finland

    Lépy, Élise; Rantala, Sinikka; Huusko, Antti; Nieminen, Pentti; Hippi, Marjo; Rautio, Arja

    2016-01-01

    (1) Background: In Finland, slippery snowy or icy ground surface conditions can be quite hazardous to human health during wintertime. We focused on the impacts of the variability in weather conditions on tourists’ health via documented accidents during the winter season in the Sotkamo area. We attempted to estimate the slipping hazard in a specific context of space and time focusing on the weather and other possible parameters, responsible for fluctuations in the numbers of injuries/accidents; (2) Methods: We used statistical distributions with graphical illustrations to examine the distribution of visits to Kainuu Hospital by non-local patients and their characteristics/causes; graphs to illustrate the distribution of the different characteristics of weather conditions; questionnaires and interviews conducted among health care and safety personnel in Sotkamo and Kuusamo; (3) Results: There was a clear seasonal distribution in the numbers and types of extremity injuries of non-local patients. While the risk of slipping is emphasized, other factors leading to injuries are evaluated; and (4) Conclusions: The study highlighted the clear role of wintery weather conditions as a cause of extremity injuries even though other aspects must also be considered. Future scenarios, challenges and adaptive strategies are also discussed from the viewpoint of climate change. PMID:27537899

  19. Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions

    The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

  20. Status Report on Spent Fuel Pools under Loss-of-Cooling and Loss-of-Coolant Accident Conditions - Final Report

    Following the 2011 accident at the Fukushima Daiichi Nuclear Power Station, the Nuclear Energy Agency Committee on the Safety of Nuclear Installations decided to launch several high-priority activities to address certain technical issues. Among other things, it was decided to prepare a status report on spent fuel pools (SFPs) under loss of cooling accident conditions. This activity was proposed jointly by the CSNI Working Group on Analysis and Management of Accidents (WGAMA) and the Working Group on Fuel Safety (WGFS). The main objectives, as defined by these working groups, were to: - Produce a brief summary of the status of SFP accident and mitigation strategies, to better contribute to the post-Fukushima accident decision making process; - Provide a brief assessment of current experimental and analytical knowledge about loss of cooling accidents in SFPs and their associated mitigation strategies; - Briefly describe the strengths and weaknesses of analytical methods used in codes to predict SFP accident evolution and assess the efficiency of different cooling mechanisms for mitigation of such accidents; - Identify and list additional research activities required to address gaps in the understanding of relevant phenomenological processes, to identify where analytical tool deficiencies exist, and to reduce the uncertainties in this understanding. The proposed activity was agreed and approved by CSNI in December 2012, and the first of four meetings of the appointed writing group was held in March 2013. The writing group consisted of members of the WGAMA and the WGFS, representing the European Commission and the following countries: Belgium, Canada, Czech Republic, France, Germany, Hungary, Italy, Japan, Korea, Spain, Sweden, Switzerland and the USA. This report mostly covers the information provided by these countries. The report is organised into 8 Chapters and 4 Appendices: Chapter 1: Introduction; Chapter 2: Spent fuel pools; Chapter 3: Possible accident

  1. Thermo-mechanical behaviour of coolant channels for heavy water reactors under accident conditions

    The objective of nuclear safety research programme is to develop and verify computer models to accurately predict the behavior of reactor structural components under operating and off normal conditions. Indian Pressurised Heavy Water Reactors (PHWRs) are tube type of reactors. The coolant channel assemblies, being one of the most important components, need detailed analysis under all operating conditions as well as during postulated conditions of accidents for its thermo-mechanical behaviour. One of the postulated accident scenarios for heavy water moderated pressure tube type of reactors i.e. PHWRs is Loss Of Coolant Accident (LOCA) coincident with Loss Of Emergency Core Cooling System (LOECCS). In this case, even though the reactor is tripped, the decay heat may not be removed adequately due to low or no flow condition and inventory depletion of primary side. Since the emergency core cooling system is presumed to be not available, the cooling of the fuel pins and the coolant channel assembly depends on the moderator cooling system, which is assumed to be available. Moderator cooling system is a separate system in PHWRs. In PHWRs, the fuel assembly is surrounded by pressure tube, an annulus insulating environment and a concentric calandria tube. In this postulated accident scenario, a structural integrity evaluation has been carried out to assess the modes of deformation of pressure tube-calandria tube assembly in a tube type nuclear reactor. The loading of pressure and temperature causes the pressure tube to sag/balloon and come in contact with the outer cooler calandria tube. The resulting heat transfer could cool and thus control the deformation of the pressure tube thus introducing inter-dependency between thermal and mechanical contact behaviour. The amount of heat thus expelled significantly depends on the thermal contact conductance and the nature of contact between the two tubes. Deformation of pressure tube creates a heat removal path to the relatively

  2. Development of solution behavior observation system under criticality accident conditions in TRACY

    Understanding of radiolytic gas behavior in fissile solution is very important to evaluate feedback reactivity, amount of released radioactive materials and pressure increase under criticality accident conditions. For this purpose, an observation system has been developed to observe behavior of the solution and radiolytic gas on the solution surface in TRACY (Transient Experiment Critical Facility). The system has been provided clear motion pictures, which can contribute the development of a computational code. This paper summarizes an outline of the system and experimental results. (author)

  3. Characterization and chemistry of fission products released from LWR fuel under accident conditions

    Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 20000C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab

  4. Modelling of Nuclear Fuel Under Accident Conditions by Means of Transuranus

    The TRANSURANUS fuel performance code, which is developed at the JRC-ITU and in collaboration with many partner institutes since more than three decades, has been adapted in order to be able to simulate design basis accident (DBA) conditions. In a first step, the developments and associated validation work will be summarised for LOCA conditions. This part includes modifications in the model for large strains, for the crystallographic phase transition in Zircaloy, and for burst release and large cladding deformations. In a second step, the ongoing work for simulations of RIA conditions will be outlined that include the model for the plenum temperature, along with the separate effect studies and detailed model developments made in parallel by means of multi-scale and multi-physics tools for the high burnup structure. Finally, the perspectives of model developments and needs for further verification and validation in the frame of international benchmark exercises dedicated to DBA simulations and the first phase of a severe accident, i.e. when the cylindrical fuel rod geometry is preserved, will be presented for discussion. (author)

  5. Considerations on Fail Safe Design for Design Basis Accident (DBA) vs. Design Extension Condition (DEC): Lesson Learnt from the Fukushima Accident

    The fail safety design is referred to as an inherently safe design concept where the failure of an SSC (System, Structure or Component) leads directly to a safe condition. Usually the fail safe design has been devised based on the design basis accident (DBAs), because the nuclear safety has been assured by securing the capability to safely cope with DBAs. Currently regards have been paid to the DEC (Design Extension Condition) as an extended design consideration. Hence additional attention should be paid to the concept of the fail safe design in order to consider the DEC, accordingly. In this study, a case chosen from the Fukushima accident is studied to discuss the issue associated with the fail safe design in terms of DBA and DEC standpoints. For the fail safe design to be based both on the DBA and the DEC, a Mode Changeable Fail Safe Design (MCFSD) is proposed in this study. Additional discussions on what is needed for the MCFSD to be applied in the nuclear safety are addressed as well. One of the lessons learnt from the Fukushima accident should include considerations on the fail-safe design in a changing regulatory framework. Currently the design extension condition (DEC) including severe accidents should be considered during designing and licensing NPPs. Hence concepts on the fail safe design need to be changed to be based on not only the DBA but also the DEC. In this study, a case on a fail-safe design chosen from the Fukushima accident is studied to discuss the issue associated with the fail safe design in terms of DBA and DEC conditions. For the fail safe design to be based both on the DBA and the DEC, a Mode Changeable Fail Safe Design (MCFSD) is proposed in this study. Additional discussions on what is needed for the MCFSD to be applied in the nuclear safety are addressed as well

  6. Suppression pool testing at the SIET laboratory (4). Release of fission products into the environment under severe accident conditions

    Long term effects of radioactive cesium on the environments and the contaminated water are one of the key issues for restoration of the Fukushima Daiichi nuclear power plant (NPP) Accident. In order to evaluate the cesium sources and their behaviors, source terms under the severe accident at Fukushima Daiichi NPP were discussed from both viewpoints of short and long term fission product (FP) sources. The former was evaluated by analyzing radioactive species based on monitoring post data, which suggested that one of major FP sources was from wet-well venting for decreasing the primary containment vessel (PCV) pressure. The latter was evaluated by analyzing long term trends of the contaminated water in the reactor and turbine buildings, which suggested that FP concentrations in the contaminated water during the 2 years since the accident were determined by the short term FP sources, while their saturated concentrations, due to a balance between the release from the reactor and the clean-up, were determined by the long term FP sources. In order to determine the PCV water scrubbing effects on FP removal, two kinds of experiment were carried out. A mini scale scrubbing tests based on a 1 L of grass made pool with 0.02 kg/h of steam, 0.04 kg/h of carrier gas and cesium, iodine and hematite tracers and a large scale mock-up tests based on 1000L of transparent pool with 360 kg/h of steam and hematite tracer. As a result of the mini scale scrubbing tests, it was evaluated that the steam carry over rates of cesium during pool scrubbing around the boiling temperature was 50% and that during sub-cooled boiling it was about 30%, which was also confirmed by the mock-up experiments. The chemical forms of the long term cesium source in the reactor have not been determined yet. Survey of core debris and cesium remained in the reactor and the PCV is one of most importance issues to understand the FP source term behaviors during the severe accident conditions. (author)

  7. Damage analysis of coolant piping due to local creep under severe accident condition

    During severe accident of a light water reactor, the failure of piping of the reactor cooling system could occur due to a thermal load, resulted from the heat transfer from a high temperature gas generated in the reactor core and decay heat released from fission product deposits. It is considered that, under such a condition, the short-term creep at a high temperature causes the piping failure. The objective of the present study is to predict the piping failure quantitatively. For this purpose, the development of an analytical method for the accurate prediction of the creep deformation is required, in which a creep constitutive equation taking the creep damage into account should be used, in order to evaluate the structural integrity of the piping during the severe accident. In this paper, creep constitutive equations considering the tertiary creep was fabricated for cold-drawn type 316 stainless steel (SUS316) based on the isotropic damage rule proposed by Kachanov-Rabotnov. In addition, creep analyses were performed for a pipe made of cold-drawn SUS316 under a condition that elevated temperature distribution was established in the pipe wall. The numerical results show that the damage of the pipe is quantitatively described by the damage variable introduced in the finite element analyses, and the failure characteristics are in reasonable agreement with those observed in a piping failure test. The failure time does not agree well with that the time of the piping failure test. It is, however, indicated that we can estimate the state of the failure of the coolant piping under severe accident by the accurate estimation of the temperature. (author)

  8. Experimental results from containment piping bellows subjected to severe accident conditions: Results from bellows tested in corroded conditions. Volume 2

    Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of nineteen bellows have been tested. Thirteen bellows were tested in ''like-new'' condition (results reported in Volume 1), and six were tested in a corroded condition. The tests showed that bellows in ''like-new'' condition are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage, while those in a corroded condition did not perform as well, depending on the amount of corrosion. The corroded bellows test program and results are presented in this report

  9. Heat transfer from fuel rod surface under reactivity-initiated accident conditions. NSRR experiments under varied cooling conditions

    The temperature evolution of fuel cladding during a reactivity-initiated accident (RIA) involves rapid changes in the mechanical properties of the cladding tube and is believed to play the primary role in fuel behaviors such as deformation and failure. Cladding-temperature behavior accompanied by boiling of coolant water, which is the case of an RIA in light-water reactors, is influenced by cooling conditions such as subcooling, pressure, and flow velocity. In order to study the effects of cooling conditions on the boiling heat transfer from the fuel rod surface to the coolant water, RIA-simulating experiments with fresh fuels had been conducted in the nuclear safety research reactor (NSRR) under cooling conditions with subcoolings of ∼10 to 80 K, flow velocities of 0 to ∼3 m/s, pressures of 0.1 to ∼16 MPa. In addition, pre-irradiated fuels had been subjected to the NSRR experiments under cooling conditions with subcoolings of ∼80 K, stagnant water, and atmospheric pressure. Out of the NSRR experiments, this report presents the fuel specifications, the test conditions, and the transient records during the pulse operations for the cases that the cladding temperature had been successfully measured. Characteristic parameters such as cladding peak temperatures were extracted from the transient records for summarizing the effects of cooling conditions and pre-irradiation on the heat transfer from the cladding surface. A CD-ROM's attached as an appendix. (J.P.N.)

  10. Key factors affecting urban runoff pollution under cold climatic conditions

    Valtanen, Marjo; Sillanpää, Nora; Setälä, Heikki

    2015-10-01

    Urban runoff contains various pollutants and has the potential of deteriorating the quality of aquatic ecosystems. In this study our objective is to shed light on the factors that control the runoff water quality in urbanized catchments. The effects of runoff event characteristics, land use type and catchment imperviousness on event mass loads (EML) and event mean concentrations (EMC) were studied during warm and cold periods in three study catchments (6.1, 6.5 and 12.6 ha in size) in the city of Lahti, Finland. Runoff and rainfall were measured continuously for two years at each catchment. Runoff samples were taken for total nutrients (tot-P and tot-N), total suspended solids (TSS), heavy metals (Zn, Cr, Al, Co, Ni, Cu, Pb, Mn) and total organic carbon (TOC). Stepwise multiple linear regression analysis (SMLR) was used to identify general relationships between the following variables: event water quality, runoff event characteristics and catchment characteristics. In general, the studied variables explained 50-90% of the EMLs but only 30-60% of the EMCs, with runoff duration having an important role in most of the SMLR models. Mean runoff intensity or peak flow was also often included in the runoff quality models. Yet, the importance (being the first, second or third best) and role (negative or positive impact) of the explanatory variables varied between the cold and warm period. Land use type often explained cold period concentrations, but imperviousness alone explained EMCs weakly. As for EMLs, the influence of imperviousness and/or land use was season and pollutant dependent. The study suggests that pollutant loads can be - throughout the year - adequately predicted by runoff characteristics given that seasonal differences are taken into account. Although pollutant concentrations were sensitive to variation in seasonal and catchment conditions as well, the accurate estimation of EMCs would require a more complete set of explanatory factors than used in this

  11. A guide for thermal testing transport package for radioactive material: Hypothetical accident conditions

    This document provides guidelines for planning, conducting, and reporting thermal tests on transport packages for radioactive material. Test conditions and acceptance criteria are for the hypothetical accident conditions specified in Part 71 of Title 10 Code of Federal Regulations (10 CFR 71). All type B packages for transport of radioactive material must be tested to these conditions, by physical or analytical test, and meet the acceptance criteria for certification by a U.S. regulatory agency as being in compliance with federal safety standards. The principal objective of this Thermal Test Guide (TTG) is to provide an applicant with general recommendations for development of a physical test program. Also, the TTG is in accord with the general philosophy for reviewing safety analysis reports for packaging and provides a common basis for applicants and reviewers. As there can be a large variety of package designs, the TTG is not all-inclusive. An applicant should appropriately apply the TTG to ensure acceptable test conditions and results for a particular package design. Recommended test conditions are based on a proposed ruling of 10 CFR 71 that was published in the US Federal Register. Thermal test conditions in the proposed ruling exceed those specified in the current ruling. Also, they are effectively equivalent to thermal test conditions specified by the International Atomic Energy Agency (IAEA) for Type B packages. Thus, by following the TTG recommendations, an applicant would be assured of meeting future 10 CFR 71 requirements and also complying with IAEA requirements

  12. Iodine behaviour under LWR accident conditions: Lessons learnt from analyses of the first two Phebus FP tests

    The International Phebus Fission Product programme, initiated in 1988 and performed by the French 'Institut de Radioprotection et de Surete Nucleaire' (IRSN), investigates through a series of in-pile integral experiments, key phenomena involved in light water reactor (LWR) severe accidents. The tests cover fuel rod degradation and the behaviour of fission products released via the primary coolant circuit into the containment building. The results of the first two tests, called FPT0 and Ftp, carried out under low pressure, in a steam rich atmosphere and using fresh fuel for Ftp and fuel burned in a reactor at 23 GWdt-1 for Ftp, were immensely challenging, especially with regard to the iodine radiochemistry. Some of the most important observed phenomena with regard to the chemistry of iodine were indeed neither predicted nor pre-calculated, which clearly shows the interest and the need for carrying out integral experiments to study the complex phenomena governing fission product behaviour in a PWR in accident conditions. The three most unexpected results in the iodine behaviour related to early detection during fuel degradation of a weak but significant fraction of volatile iodine in the containment, the key role played by silver rapidly binding iodine to form insoluble AgI in the containment sump and the importance of painted surfaces in the containment atmosphere for the formation of a large quantity of volatile organic iodides. To support the Phebus test interpretation small-scale analytical experiments and computer code analyses were carried out. The former, helping towards a better understanding of overall iodine behaviour, were used to develop or improve models while the latter mainly aimed at identifying relevant key phenomena and at modelling weaknesses. Specific efforts were devoted to exploring the potential origins of the early-detected volatile iodine in the containment building. If a clear explanation has not yet been found, the non-equilibrium chemical

  13. Development of solution behavior observation system under criticality accident conditions in TRACY

    An observation system has been developed as a new instrumentation of TRACY (Transient Experiment Critical Facility) in order to observe the behavior of uranyl nitrate solution and radiolytic gas voids under criticality accident conditions. The system consists of a radiation-resistive optical fiberscope, a light source and a radiation-resistive video camera. The severe radiation environment in TRACY and safety functions as the primary boundary of TRACY were considered in the design of the system. The system has been successfully utilized in the recent TRACY experiments, and provided clear color motion pictures showing the behavior of the solution and radiolytic gas voids. As a result, it was visually confirmed that there is the difference in the behavior of the solution and radiolytic gas voids depending on the conditions of the reactivity addition. The system provides detailed information on the behavior of the solution and voids, and will contribute to the development of a computational kinetics model. (author)

  14. Metallurgical study of failed specimen and piping under LWR severe accident conditions

    The WIND (Wide range piping INtegrity Demonstration) project is being performed at JAERI in order to demonstrate the integrity of LWRs piping under the severe accident conditions. The hot tensile test, the short-term creep rupture test and the piping failure test of the piping materials which are the same as those of a reactor coolant system piping used in Japanese LWRs, have been performed in order to identify the failure conditions of RCS pipings. The 0.2% proof stress and the ultimate tensile strength above 800degC are given as a function of the temperature for served piping materials. The effect of the remicrostructure, i.e., the precipitation, the growth and the resolution of the precipitates, under high-temperature and high-stress on the integrity, is discussed with the results of the metallurgical study. (author)

  15. Investigation of the behaviour of packaged radioactive waste under fire accident conditions

    A study has been made of the behaviour of packaged intermediate level waste (ILW) when exposed to fire conditions so as to provide information to support safety cases for ILW transport and disposal. The temperatures used in the study were selected to exceed those that the waste might be subject to in fire accidents during the transport and handling of ILW. Four waste materials, immobilised in cement or in organic resin, with properties representative of a wide range of waste streams were included in the study. Tests were carried out on samples of both real waste materials and non-radioactive simulants, and also on full-scale (500 litre) drums of simulant wastes. The overall release fractions were low, even for external temperatures of up to 1000oC. Examination showed that the stainless steel drums were still in good condition and on sectioning, little damage to the matrix or decrease in its strength was evident. (author)

  16. Comparison and verification of two computer programs used to analyze ventilation systems under accident conditions

    Two computer codes, TVENT and EVENT, which were developed at the Los Alamos National Laboratory (LANL) for the analysis of ventilation systems, have been modified to model air-cleaning systems that include active components with time-dependent flow-resistance characteristics. With both modified programs, fluid-dynamic transients were calculated for a test facility used to simulate accident conditions in air-cleaning systems. Experiments were performed in the test facility whereby flow and pressure transients were generated with the help of two quick-actuating air-stream control valves. The numerical calculations are compared with the test results. Although EVENT makes use of a more complex theoretical flow model than TVENT, the numerical simulations of both codes were found to be very similar for the flow conditions studied and to closely follow the experimental results

  17. Analyses of conditions in a large, dry PWR containment during an TMLB' accident sequence

    The aim of the paper is to give an assessment of the conditions which would develop in the large, dry containment of a modern Westinghouse-type PWR during a severe accident where all safety systems are unavailable. The analysis is based principally on the results of calculations using the CONTAIN code, with a 4 cell model of the containment, for a station blackout (TMLB') scenario in which the vessel is assumed to fail at high pressure. In particular, the following are noted: (i) If much of the debris is in contact with water, so that decay heat can boil water directly, then the pressure rises steadily to reach the assumed containment failure point after 11/2 to 2 days. If most of the debris becomes isolated from water, for example, because of water is held up on the containment floors and in sumps and drains, the pressure rises too slowly to threaten the containment on this timescale. (ii) If a core-concrete interaction occurs, most of the associated fission product release takes place soon after relocation of molten fuel to the containment. The aerosols which transport these (and other non-gaseous fission products released earlier in the accident) in the containment agglomerate and settle. As a result, 0.1% or less of the aerosols remain airborne a day after the start of the accident. (iii) Hydrogen and carbon monoxide, which would accumulate in the containment are not expected to burn because the atmosphere would be inerted by steam. If, however, enough of the steam is condensed, for example, by recovering the containment sprays, a burn could occur but the resulting pressure spike is unlikely to threaten the containment unless a transition to detonation occurs. 6 refs., 6 tabs., 12 figs

  18. Insights on fission products behaviour in nuclear severe accident conditions by X-ray absorption spectroscopy

    Geiger, E.; Bès, R.; Martin, Ph; Pontillon, Y.; Ducros, G.; Solari, P. L.

    2016-04-01

    Many research programs have been carried out aiming to understand the fission products behaviour during a Nuclear Severe Accident. Most of these programs used highly radioactive irradiated nuclear fuel, which requires complex instrumentation. Moreover, the radioactive character of samples hinders an accurate chemical characterisation. In order to overcome these difficulties, SIMFUEL stand out as an alternative to perform complementary tests. A sample made of UO2 doped with 11 fission products was submitted to an annealing test up to 1973 K in reducing atmosphere. The sample was characterized before and after the annealing test using SEM-EDS and XAS at the MARS beam-line, SOLEIL Synchrotron. It was found that the overall behaviour of several fission products (such as Mo, Ba, Pd and Ru) was similar to that observed experimentally in irradiated fuels and consistent with thermodynamic estimations. The experimental approach presented in this work has allowed obtaining information on chemical phases evolution under nuclear severe accident conditions, that are yet difficult to obtain using irradiated nuclear fuel samples.

  19. Fission product aerosol removal test by containment spray under accident management conditions (2)

    In order to demonstrate effective fission product (FP) aerosol removal and pressure suppression effects by containment spray under Japanese accident management (AM) conditions, system integral tests simulating typical BWR accident sequences have been carried out using a full-height simulation containment vessel test facility. In case of 10% reduction spray flow rate comparing with a reference test case, aerosol concentration in the entire drywell (D/W) decreased rapidly about 1/10 of initial concentration within 30 min after the spray initiation and remained low through 12 hours test period similar to the reference test case. The maximum pressure was slightly higher in this case. Both the existence of non-condensable gas and the location of aerosol injection did not affect both pressure suppression effect and aerosol removal effect. In case of aerosol injection into the middle D/W, aerosol concentration in the upper D/W was relatively high, but the concentrations in the middle and the lower D/W were extremely low. The degradation of FP removal due to the existence of non-condensable gas was supplemented by FP removal of pool scrubbing in suppression chamber. After the modification of FP removal model in MELCOR, calculated time dependency of CsI aerosol concentration and pressure in the D/W agreed well with the test data. (author)

  20. Performance Analysis Review of Thorium TRISO Coated Particles during Manufacture, Irradiation and Accident Condition Heating Tests

    Thorium, in combination with high enriched uranium, was used in all early high temperature reactors (HTRs). Initially, the fuel was contained in a kernel of coated particles. However, particle quality was low in the 1960s and early 1970s. Modern, high quality, tristructural isotropic (TRISO) fuel particles with thorium oxide and uranium dioxide (UO2) had been manufactured since 1978 and were successfully demonstrated in irradiation and accident tests. In 1980, HTR fuels changed to low enriched uranium UO2 TRISO fuels. The wide ranging development and demonstration programme was successful, and it established a worldwide standard that is still valid today. During the process, results of the thorium work with high quality TRISO fuel particles had not been fully evaluated or documented. This publication collects and presents the information and demonstrates the performance of thorium TRISO fuels.This publication is an outcome of the technical contract awarded under the IAEA Coordinated Research Project on Near Term and Promising Long Term Options for Deployment of Thorium Based Nuclear Energy, initiated in 2012. It is based on the compilation and analysis of available results on thorium TRISO coated particle performance in manufacturing and during irradiation and accident condition heating tests

  1. Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

    Xu Wu; Piyush Sabharwall; Jason Hales

    2014-07-01

    This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO2/zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

  2. Containment performance of prototypical reactor containments subjected to severe accident conditions

    In SECY-90-016, the NTRC proposed a safety goal of a conditional containment failure probability (CCFP) of 0.1 and the alternative acceptance criteria allowed for steel containments, which specifies that the stresses should not exceed ASNE Level C allowables for severe accident pressures and temperatures. In this work, the need for an equivalent criterion for concrete containments was studied. Six surrogate containments were designed and analyzed in order to compare the margins between design pressure, pressure resulting in exceedance of Level C (or yield) stress limits, and ultimate pressure. For comparability, each containment has an identical internal volume and design pressure. Results from the analysis showed margins to yield are comparable and display a similar margin for both steel and concrete containments. In addition, the margin to failure, although slightly higher in the steel containments, were also comparable. Finally, a CCFP for code design was determined based on general membrane behavior and imposing an upper bound severe accident curve developed in the DCH studies. The resulting CCFP's were less then 0.02 (or 2%) for all the surrogate containments studied, showing that these containment designs all achieved the NRC safety goal

  3. Containment performance of prototypical reactor containments subjected to severe accident conditions

    Klamerus, E.W.; Bohn, M.P. [Sandia National Labs., Albuquerque, NM (United States); Wesley, D.A. [EQE International, Irvine, CA (United States); Krishnaswamy, C.N. [Sargent & Lundy, Chicago, IL (United States)

    1996-12-01

    In SECY-90-016, the NTRC proposed a safety goal of a conditional containment failure probability (CCFP) of 0.1 and the alternative acceptance criteria allowed for steel containments, which specifies that the stresses should not exceed ASNE Level C allowables for severe accident pressures and temperatures. In this work, the need for an equivalent criterion for concrete containments was studied. Six surrogate containments were designed and analyzed in order to compare the margins between design pressure, pressure resulting in exceedance of Level C (or yield) stress limits, and ultimate pressure. For comparability, each containment has an identical internal volume and design pressure. Results from the analysis showed margins to yield are comparable and display a similar margin for both steel and concrete containments. In addition, the margin to failure, although slightly higher in the steel containments, were also comparable. Finally, a CCFP for code design was determined based on general membrane behavior and imposing an upper bound severe accident curve developed in the DCH studies. The resulting CCFP`s were less then 0.02 (or 2%) for all the surrogate containments studied, showing that these containment designs all achieved the NRC safety goal.

  4. Nuclear waste shipping container response to severe accident conditions, A brief critique of the modal study

    Audin, L.

    1990-12-01

    The Modal Study (NUREG/CR-4829) attempts to upgrade the analysis of spent nuclear fuel transportation accidents, and to verify the validity of the present regulatory scheme of cask performance standards as a means to minimize risk. While an improvement over many prior efforts in this area (such as NUREG-0170), it unfortunately fails to create a realistic simulation either of a shipping cask, the severe conditions to which it could be subjected, or the potential damage to the spent fuel cargo during an accident. There are too many deficiencies in its analysis to allow acceptance of its results for the presumed cask design, and many pending changes in new containers, cargoes and shipping patterns will limit applicability of the Modal Study to future shipments. In essence, the Modal Study is a good start, but is too simplistic, incomplete, outdated and open to serious question to be used as the basis for any present-day environmental or risk assessment of spent fuel transportation. It needs to be redone, with peer review during its production and experimental verification of its assumptions, before it has any relevance to the shipments planned to Yucca Mountain. Finally, it must be expanded into a full risk assessment by inputing its radiological release fractions and probabilities into a valid dispersal simulation to properly determine the impact of its results. 51 refs.

  5. Measurement of neutron dose under criticality accident conditions at TRACY using ebonites

    Neutron doses under criticality accident conditions at TRACY were measured using ebonites, which are hard rubber containing sulfur. Ebonites can be easily available and are inexpensive since they are generally used as an insulator. To evaluate a neutron dose, beta rays emitted from 32P induced by 32S(n, p) reaction in an ebonite disc are measured with Geiger-Mueller (GM) counter. Then, a calibration factor (Gy/cpm), which is pre-determined using a 252Cf source, is applied to the count rates to obtain neutron doses. Factors to correct for the difference between responses of 32S(n, p) induced in an ebonite to the spontaneous fission spectrum of 252Cf calibration source and to spectra of TRACY were calculated using MCNP5, and applied to the doses. In the experiments, ebonites placed in free air and on phantom were exposed by TRACY with and without the water reflector. Neutron doses measured with ebonites in TRACY without a reflector were evaluated with an uncertainty of less than about 40%, and were consistently overestimated. On the other hand, average of neutron doses measured with ebonites in TRACY with the water reflector were accurate; however, the disparsion of neutron dose per integrated power of TRACY was large. By these measurements it was found that ebonites can be used as a neutron dosimeter for criticality accidents. (author)

  6. Radioecological estimation of the condition of wild fauna in the zone of Chernobyl nuclear accident

    As the result of long time of wildlife radioecological monitoring in the zone of Chernobyl nuclear accident the main trends in radioactive contamination of the animals of different taxones, the condition of fauna biodiversity have been shown. After a noticeable decrease of the radionuclide contents observed in the period immediately following the accident which was mainly caused by decay of short-living isotopes, in recent years a tendency of stabilising the radionuclide accumulation was found in the majority of the animal groups. The dynamics and state of the fauna depends more on the secondary effects of human evacuation than on direct radioecological impact. Natural ecological succession may have accelerated due to the post-evacuation removal of human pressure on contaminated habitats. Cessation of economic activity had the greatest effect on the structure and number of ornithocomplexes and populations of commercial game mammals. Changes in aquatic animals are expressed to a smaller extent and follow the laws of natural development to a greater extent. These dynamics processes of transformation of wildlife communities offer a unique opportunity to study the development and conservation of wild animal biodiversity within the context of specific land use and landscape ecological changes. (authors)

  7. Nuclear waste shipping container response to severe accident conditions, A brief critique of the modal study

    The Modal Study (NUREG/CR-4829) attempts to upgrade the analysis of spent nuclear fuel transportation accidents, and to verify the validity of the present regulatory scheme of cask performance standards as a means to minimize risk. While an improvement over many prior efforts in this area (such as NUREG-0170), it unfortunately fails to create a realistic simulation either of a shipping cask, the severe conditions to which it could be subjected, or the potential damage to the spent fuel cargo during an accident. There are too many deficiencies in its analysis to allow acceptance of its results for the presumed cask design, and many pending changes in new containers, cargoes and shipping patterns will limit applicability of the Modal Study to future shipments. In essence, the Modal Study is a good start, but is too simplistic, incomplete, outdated and open to serious question to be used as the basis for any present-day environmental or risk assessment of spent fuel transportation. It needs to be redone, with peer review during its production and experimental verification of its assumptions, before it has any relevance to the shipments planned to Yucca Mountain. Finally, it must be expanded into a full risk assessment by inputing its radiological release fractions and probabilities into a valid dispersal simulation to properly determine the impact of its results. 51 refs

  8. RADIATION CONDITIONS IN KALUGA REGION 30 YEARS AFTER CHERNOBYL NPP ACCIDENT

    A. G. Ashitko

    2016-01-01

    Full Text Available The article describes radiation conditions in the Kaluga region 30 years after the Chernobyl NPP accident. The Chernobyl NPP accident caused radioactive contamination of nine Kaluga region territories: Duminichsky, Zhizdrinsky, Kuibyshevsky, Kirovsky, Kozelsky, Ludinovsky, Meshchovsky, Ulyanovsky and Hvastovichsky districts. Radioactive fallout was the strongest in three southern districts: Zhizdrinsky, Ulyanovsky and Hvastovichsky, over there cesium-137 contamination density is from 1 to 15Ci/km. According to the Russian Federation Government Order in 2015 there are 300 settlements (S in the radioactive contamination zone, including 14 settlements with caesium-137 soil contamination density from 5 to 15 Ci/ km2 and 286 settlements with the contamination density ranging from 1 to 5 Ci/km2. In the first years after the Chernobyl NPP accident in Kaluga region territories, contaminated with caesium-137, there were introduced restrictive land usage, were carried out agrochemical activities (ploughing, mineral fertilizer dressing, there was toughened laboratory radiation control over the main doze-forming foodstuff. All these measures facilitated considerable decrease of caesium-137 content in local agricultural produce. Proceeding from the achieved result, in 2002 there took place the transition to more tough requirements SanPiN 2.3.2.1078-01. Analysis of investigated samples from Zhizdrinsky, Ulyanovsky and Hvastovichsky districts demonstrated that since 2005 meat samples didn’t exceed the standard values, same for milk samples since 2007. Till the present time, the use of wild-growing mushrooms, berries and wild animals meat involves radiation issues. It was demonstrated that average specific activity of caesium-137 in milk samples keeps decreasing year after year. Long after the Chernobyl NPP accident, the main products forming internal irradiation doses in population are the wild-growing mushrooms and berries. Population average annual

  9. Study of heat and mass transfer phenomena in fuel assembly models under accident conditions

    The majority of the material in support of the thermal - hydraulic safety of WWER core was obtained on single - assembly models containing a relatively small number of elements - heater rods. Upgrading the requirements to the reactor safety leads to the necessity for studying phenomena in channels representing the cross - sectional core dimensions and non - uniform radial power generation. Under such conditions, the contribution of natural convection can be significant in some core zones, including the occurrence of reverse flows and interchannel instability. These phenomena can have an important influence on heat transfer processes. Such influence is especially drastical under accident conditions associated with ceasing the forced circulation over the circuit. A number of urgent reactor safety problems at low operating parameters is related with the computer code verification and certification. One of the important trends in the reactor safety research is concerned with the rod bundle reflooding and verificational calculations of this phenomenon. To assess the water cooled reactor safety, the best fit computer codes are employed, which make it possible to simulate accident and transient operating conditions in a reactor installation. One of the most widely known computer codes is the RELAP5/MOD3 Code. The paper presents the comparison of the results calculated using this computer code with the test data on 4 - rod bundle quenching, which were obtained at the SSCRF-IPPE. Recently, the investigations on the steam - zirconium reaction kinetics have been performed at the SSCFR-IPPE and are being presently performed for the purpose of developing new and verifying available computer codes. (author). 3 refs, 6 figs

  10. Pretest analysis of containment studies facility model for simulated loss of coolant accident conditions

    An experimental facility called Containment Studies Facility (CSF) has been constructed at Bhabha Atomic Research Centre (BARC), Trombay for the purpose of research and development in the area of nuclear reactor containment thermal hydraulics. The facility consists of reinforced concrete containment structural model and a Primary Heat Transport Model (PHTM) vessel. The containment model is approximately 1:250 volumetrically scaled down model of a 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR) containment system and the PHTM represents the primary heat transport system of the prototype reactor. The PHTM with a pressure vessel and associated pump and piping system is designed for simulating the Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB) conditions within the containment model. As part of CSF project thermal hydraulic analysis, a pretest analysis was carried out for simulated LOCA conditions. Blow down mass and energy discharge data were obtained using Relap/MOD3.2 code for different blow down conditions and were used as inputs to CONTRAN code for simulating LOCA or main steam line break (MSLB) conditions in the containment model. Pressure and temperature transients in the CSF model for different blow down conditions and a number of parametric studies were conducted to assess the influence of a large number of thermodynamic and geometrical parameters which are known to affect the transients and alter the peak pressure and temperature values. (author)

  11. Isotopic fission product release from nuclear fuel under severe core damage accident conditions

    Isotopic fission gas release behavior during SFD tests 1-1, 1-3, and 1-4 is strongly dependent on the pre-test fuel history. For SFD 1-1, where the majority of all the fission products were generated during the preconditioning period, very little difference in isotopic release behavior between short- and long-lived species is predicted. For the SFD 1-3 and 1-4 tests, where the majority of all short-lived fission gases decayed away during the 4-year cooling period, differences between the behavior of long- and short-lived species are predicted. Most of the intragranular fission product release has been shown to be due to a grain-growth/grain-boundary-sweeping mechanism. In addition, fuel liquefaction/dissolution processes can lead to increased release under these degraded-core-accident conditions. These predictions follow the trend of the observed phenomena

  12. Vaporization of low-volatile fission products under severe CANDU reactor accident conditions

    An analytical model has been developed to describe the release behaviour of low-volatile fission products from uranium dioxide fuel under severe reactor accident conditions. The effect of the oxygen potential on the chemical form and volatility of fission products is determined by Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix or fission product vaporization from the fuel surface. The effect of fuel volatilization (i.e., matrix stripping) on the release behaviour is also considered. The model has been compared to data from an out-of-pile annealing experiment performed in steam at the Chalk River Laboratories. (author)

  13. Testing of irradiated spherical fuel elements at HTR MODUL relevant accident conditions

    It is reported that the German 200 MWth MODUL HTR uses spherical fuel elements having 10% enriched UO2 TRISO coated particles. Since 1984 the behavior of such elements of modern design under accident conditions has been studied at the Research Centre Juelich, FRG. By help of the Cold Finger Apparatus even the smallest release of fission products during testing up to 1800 deg. C can be analysed. Post heating examinations allowed important correlations between the distribution within the fuel element and the measured sphere release. The results of heating tests are described. Further work was carried out to simulate water and air ingress in a HTR. AN apparatus was built and is now commissioned. Tests with special samples and fuel spheres, and also with USA fuel are planned, to examine the influence of humidity on the fission product release. 14 refs, 13 figs, 7 tabs

  14. Rehabilitation of life conditions in territories contaminated by Chernobyl accident. ETHOS project in Ukraine

    This article presents the ETHOS project funded by European Union and whose aim is to stimulate a lasting rehabilitation of life conditions in the territories contaminated by Chernobyl nuclear accident. The daily life of people living in the contaminated regions has been affected not only on medical aspect but also on economic, ecological, social and cultural aspects. The strict regulations imposed by radiation protection authorities have been a major element to the degradation of the standard of living. ETHOS project is based on a cooperation between the authorities and the inhabitants and on a strong motivation of the people, for instance in the Olmany village 6 work groups have been organized around themes such as: the improvement of the quality of the milk and meat produced in the village, the radiation protection of children, the practical basics to know when living in a contaminated area, and the right management of home wastes like ashes that are particularly contaminated. (A.C.)

  15. Criticality safety analysis of IRT-200 storage pool under normal and accident conditions

    In the paper some results of nuclear safety analysis of the research reactor IRT-200 storage pool with IRT-4M fuel assemblies, during storage and fuel assembly manipulations are presented. The calculations have been performed by the modular code system SCALE, which is world widely used for criticality safety analysis of facilities for transport and storage of spent nuclear fuel. Conservative evaluation of the effective multiplication factor Keff of the storage pool for both: normal operation and assembly drop accident, is made. The analysis of the obtained results shows that the technological equipment and the storage conditions assure safety during the storage and manipulations of IRT-4M fuel assemblies in accordance with the requirements of the Bulgarian norms and standards, Keff < 0.95. (authors)

  16. A model for non-volatile fission product release during reactor accident conditions

    An analytical model has been developed to describe the release kinetics of non-volatile fission products (e.g., Mo, Ce, Ru and Ba) from uranium dioxide fuel under severe reactor accident conditions. The present treatment considers the rate-controlling process of release in accordance with diffusional transport in the fuel matrix and fission product vaporization from the fuel surface into the surrounding gas atmosphere. The effect of the oxygen potential in the gas atmosphere on the chemical form and volatility of the fission product is considered. A correlation is also developed to account for the trapping effects of Sb and Te in the Zircaloy cladding. This model has been used to interpret the release behaviour of fission products observed in the CEA experiments conducted in the HEVA/VERCORS facility at high temperature in a hydrogen and steam atmosphere. (author)

  17. Experimental investigation of symmetric and asymmetric heating of pressure tube under accident conditions for Indian PHWR

    Highlights: ► Circumferential temperature gradient for asymmetric heat-up was 400 °C. ► At same pressure ballooning initiates at lower temperature in asymmetrical heat-up. ► At 1 MPa ballooning initiated at 408 °C and with expansion rate of 0.005 mm/s. ► At 2 MPa ballooning initiation at 330 °C and with expansion rate of 0.0056 mm/s. ► For symmetrical heat-up strain rate was 10 times faster than asymmetric heat-up. - Abstract: In pressurized heavy water reactor (PHWR), under postulated scenario of small break Loss of Coolant Accident (LOCA) coincident with the failure of Emergency Core Cooling System (ECCS), a situation may arise under which reduction in mass flow rate of coolant through individual reactor channel can lead to stratified flow. Such stratified flow condition creates partial uncover of fuel bundle, which creates a circumferential temperature gradient over PT. The present investigation has been carried out to study thermo-mechanical behaviour of PT under asymmetric heating conditions for a 220 MWe PHWR. A 19-pin fuel simulator has been developed in which preferential heating of elements could be done by supplying power to the selected pins. The asymmetric heating of PT has been carried out at pressure 2 MPa and 1 MPa, respectively, by supplying power to upper region heating elements thus creating an half filled stratified flow conditions. The temperature difference up to 425 °C has been observed along top to bottom periphery of PT. A comparison is made between thermo-mechanical behaviour of PT under asymmetrical and symmetrical heat-up, expected from a large break LOCA condition. The radial expansion rate during symmetrical heating is found to be much faster as compared to that for asymmetric ballooning of PT at the same internal pressure. Integrity of PT is found to be maintained under both loading conditions. Heat sink around of test section, simulating moderator is found to be helpful in arresting the rise in temperature for both fuel

  18. Experimental investigation of symmetric and asymmetric heating of pressure tube under accident conditions for Indian PHWR

    Yadav, Ashwini K., E-mail: ashwinikumaryadav@gmail.com [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee-247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai-400085 (India); Kumar, Ravi, E-mail: ravikfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee-247667 (India); Chatterjee, B., E-mail: barun@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai-400085 (India); Gupta, Akhilesh, E-mail: akhilfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee-247667 (India); Lele, H.G., E-mail: hglele@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai-400085 (India)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Circumferential temperature gradient for asymmetric heat-up was 400 Degree-Sign C. Black-Right-Pointing-Pointer At same pressure ballooning initiates at lower temperature in asymmetrical heat-up. Black-Right-Pointing-Pointer At 1 MPa ballooning initiated at 408 Degree-Sign C and with expansion rate of 0.005 mm/s. Black-Right-Pointing-Pointer At 2 MPa ballooning initiation at 330 Degree-Sign C and with expansion rate of 0.0056 mm/s. Black-Right-Pointing-Pointer For symmetrical heat-up strain rate was 10 times faster than asymmetric heat-up. - Abstract: In pressurized heavy water reactor (PHWR), under postulated scenario of small break Loss of Coolant Accident (LOCA) coincident with the failure of Emergency Core Cooling System (ECCS), a situation may arise under which reduction in mass flow rate of coolant through individual reactor channel can lead to stratified flow. Such stratified flow condition creates partial uncover of fuel bundle, which creates a circumferential temperature gradient over PT. The present investigation has been carried out to study thermo-mechanical behaviour of PT under asymmetric heating conditions for a 220 MWe PHWR. A 19-pin fuel simulator has been developed in which preferential heating of elements could be done by supplying power to the selected pins. The asymmetric heating of PT has been carried out at pressure 2 MPa and 1 MPa, respectively, by supplying power to upper region heating elements thus creating an half filled stratified flow conditions. The temperature difference up to 425 Degree-Sign C has been observed along top to bottom periphery of PT. A comparison is made between thermo-mechanical behaviour of PT under asymmetrical and symmetrical heat-up, expected from a large break LOCA condition. The radial expansion rate during symmetrical heating is found to be much faster as compared to that for asymmetric ballooning of PT at the same internal pressure. Integrity of PT is found to be

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

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

  20. Hydrogen-management in beyond design accident conditions in NPP Neckar 2

    Neckar 2 is a 1340 MWE 4-loop pressurized water reactor (PWR) of Siemens KONVOI type, located in the south of Germany. It was first connected to the grid in January 1989. Commercial operation started in April 1989. Task assignment: In Germany it was recommended by the Reactor Safety Commission (RSK) on December 17, 1997, to reequip passive autocatalytic recombiners for the controlling of the hydrogen problem. The removal of the hydrogen is an essential part which guarantees the integrity of the containment. The implementation of the recombiners is a further step for the decrease of the nuclear rest risk. The RSK confirmed, that the implementation of the passive autocatalytic recombiners is a safety measure for the controlled removal of the hydrogen in beyond design accident conditions. Assumption : Failure of the whole residual heat removal system (RHRS) and non sufficient effect of the systems which have been installed for beyond design accident conditions. Effect on the reactor coolant system (RCS): The reactor core will be damaged by non sufficient cooling with the output of hydrogen because all the specified emergency actions have failed. The overheating of the core is responsible for the production of hydrogen by the reaction of zirconium of the fuel-rod cladding with the water vapour. In case of nuclear superheating it would be possible that the reactor vessel would start smelting. The interacting between the core and the concrete, together with the armouring of the biological shield would also produce hydrogen. The hydrogen would escape together with the water vapour out of the leak and would spread out into the whole containment. Results : the number and the position of the different sized recombiners were determined on engineering judgement. the following 4 scenarios are representatively. The 4 scenarios were analyzed for in beyond design accident conditions with the MELCOR-Code: No. 1: Loss of main feedwater supply with primary feed and bleed. No. 2

  1. Aging, condition monitoring, and loss-of-coolant accident (LOCA) tests of Class 1E electrical cables: Summary of results

    This paper summarizes the results of aging, condition monitoring, and accident testing of Class 1E cables used in nuclear power generating stations. Three sets of cables were aged for up to 9 months under simultaneous thermal (≅ 100 degrees C) and radiation (≅0.10 kGy/hr) conditions. After the aging, the cables were exposed to a simulated accident consisting of high dose rate irradiation (≅6 kGy/hr) followed by a high temperature steam exposure. A fourth set of cables, which were unaged, were also exposed to the accident conditions. The cables that were aged for 3 months and then accident tested were subsequently exposed to a high temperature steam fragility test (up to 400 degrees C), while the cables that were aged for 6 months and then accident tested were subsequently exposed to a 1000-hour submergence test in a chemical solution. The results of the tests indicate that the feasibility of life extension of many popular nuclear power plant cable products is promising and that mechanical measurements (primarily elongation, modulus, and density) were more effective than electrical measurements for monitoring age-related degradation. In the high temperature steam test, ethylene propylene rubber (EPR) cable materials generally survived to higher temperatures than crosslinked polyolefin (XLPO) cable materials. In dielectric testing after the submergence testing, the XLPO materials performed better than the EPR materials. This paper presents some recent experimental data that are not yet available elsewhere and a summary of findings from the entire experimental program

  2. Aging, condition monitoring, and loss-of-coolant accident (LOCA) tests of Class 1E electrical cables: Summary of results

    This paper summarizes the results of aging, condition monitoring, and accident testing of Class 1E cables used in nuclear power generating stations. Three sets of cables were aged for up to 9 months under simultaneous thermal (≅ 100C) and radiation (≅ 0.10 kGy/hr) conditions. After the aging, the cables were exposed to a simulated accident consisting of high dose rate irradiation (≅ 6 kGy/hr) followed by a high temperature steam exposure. A fourth set of cables, which were unaged, were also exposed to the accident conditions. The cables that were aged for 3 months and then accident tested were subsequently exposed to a high temperature steam fragility test (up to 400C), while the cables that were aged for 6 months and then accident tested were subsequently exposed to a 1,000-hour submergence test in a chemical solution. The results of the tests indicate that the feasibility of life extension of many popular nuclear power plant cable products is promising and that mechanical measurements (primarily elongation, modulus, and density) were more effective than electrical measurements for monitoring age-related degradation. In the high temperature steam test, ethylene propylene rubber (EPR) cable materials generally survived to higher temperatures than crosslinked polyolefin (XLPO) cable materials. In dielectric testing after the submergence testing, the XLPO materials performed better than the EPR materials. This paper presents some recent experimental data that are not yet available elsewhere and a summary of findings from the entire experimental program

  3. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  4. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made

  5. A RESEARCH ON WORKING CONDITIONS AND OCCUPATIONAL ACCIDENTS AT THE FOREST HARVESTING ACTIVITIES

    MENEMENCİOĞLU, Kayhan

    2009-01-01

    This study was condcuted to have some information about the occupational accident ratio and some habits of forest harvesting workers, and the reasons of the accidents. The data obtained from a total of 250 forest workers responded to a self-administered questionnaire working at the harvesting practises at Directorates of Forest Enterprises Adana, Oltu, Ilgaz, Pazar and Sındırgı were discussed. According to findings; 47 % of forestry workers had at least an occupational accident and the cause...

  6. Development of Instrument Transmitter Protecting Device against High-Temperature Condition during Severe Accidents

    Min Yoo; Sung Min Shin; Hyun Gook Kang

    2014-01-01

    Reliable information through instrumentation systems is essential in mitigating severe accidents such as the one that occurred at the Fukushima Daiichi nuclear power plant. There are five elements which might pose a potential threat to the reliability of parameter detection at nuclear power plants during a severe accident: high temperature, high pressure, high humidity, high radiation, and missiles generated during the evolution of a severe accident. Of these, high temperature apparently pose...

  7. Fission product aerosol removal test by containment spray under accident management conditions

    This paper summarizes the test results of fission product (FP) removal by containment spray simulating accident management (AM) condition. The features of AM conditions concerning FP transport are characterized by (1) low flow spray affecting the steam condensation degradation due to larger water droplets, (2) high humidity condition due to steam generation as a result of debris cooling and (3) continual fresh water supply from outside water source. The objectives of the test program are to provide data demonstrating the effective aerosol removal by the containment spray and to provide the data for qualification of the integral system analysis code such as MELCOR. The Tests were conducted using full-height simulation containment vessels of GIRAFFE (1/720 volumetric scaling ratio) so that real FP removal phenomena was preserved as in a reactor. Vessel heat loss was compensated by heaters on the outer surface of the vessels. CsI was selected as a typical FP aerosol. Steam generated by decay heat, CsI aerosol and spray water were supplied continuously to the drywell as transient boundary conditions. A system integration test simulating BWR low pressure vessel failure sequence during about 10 hours were successfully accomplished. Even under low spray flow condition, maximum drywell pressure was kept relatively low, though it was a little bit higher than the design pressure. After spray initiation, aerosol concentration decreased rapidly in the entire region of drywell. In the upper drywell, aerosol was removed by diffusiophoresis associated with steam condensation, while in the lower drywell it was removed by impaction. By modifying the FP removal model in the MELCOR, calculated FP concentration transient as well as pressure transient agreed well with test data. (J.P.N.)

  8. DNBR analyses under steady-state and accident conditions for a double-flat-core high conversion light water reactor

    A double-flat-core high conversion light water reactor (HCLWR-JDF1) has been developed at JAERI aiming at better fuel utilization and higher safety margin. The HCLWR has two pancake type cores piled up with lower, internal and upper axial blankets. Fuel rods are arranged in a triangular lattice with p/d = 1.23. The lengths of each core part and each blanket part are 60 cm and 30 cm, respectively. Departure from nucleate boiling (DNB) analyses were performed under steady-state operational condition and accident conditions. The primary coolant pump trip accident and locked rotor accident were selected for the transient analyses. The primary system transient calculations under accident conditions were performed with a best-estimate code J-TRAC using the same conservative assumptions as in the licensing calculation for a current LWR. The KfK critical heat flux (CHF) correlation coupled with the COBRA-IV-I subchannel analysis was used to evaluate the DNB ratio (DNBR). The KfK correlation was verified with the data from small scale (4 and 7 rods) CHF experiments at JAERI and 20-rods CHF experiments at Bettis Atomic Power Laboratory. The mixing coefficient and grid spacer loss coefficient used in the subchannel analyses were experimentally determined. Based on the criterion that no fuel rod in the core experiences DNB with 95 % probability at 95 % confidence level, which is used in the current LWR licensing procedure, the minimum DNBR was determined to be 1.28 with the KfK correlation. The estimated minimum DNBR's were 1.66 for the steady-state condition, 1.56 for the pump trip accident and 1.34 for the locked rotor accident. These minimum DNBR's are larger than the minimum allowable DNBR limit. The results indicate that the present HCLWR design is acceptable from a view point of the DNBR criterion. (author)

  9. The influence of simultaneous or sequential test conditions in the properties of industrial polymers, submitted to PWR accident simulations

    The effect of PWR plant normal and accident operating conditions on polymers forms the basis of nuclear qualification of safety-related containment equipment. This study was carried out on the request of safety organizations. Its purpose was to check whether accident simulations carried out sequentially during equipment qualification tests would lead to the same deterioration as that caused by an accident involving simultaneous irradiation and thermodynamic effects. The IPSN, DAS and the United States NRC have collaborated in preparing this study. The work carried out by ORIS Company as well as the results obtained from measurement of the mechanical properties of 8 industrial polymers are described in this report. The results are given in the conclusion. They tend to show that, overall, the most suitable test cycle for simulating accident operating conditions would be one which included irradiation and consecutive thermodynamic shock. The results of this study and the results obtained in a previous study, which included the same test cycles, except for more severe thermo-ageing, have been compared. This comparison, which was made on three elastomers, shows that ageing after the accident has a different effect on each material

  10. Water reactor fuel behaviour and fission products release in off-normal and accident conditions

    The present meeting was scheduled by the International Atomic Energy Agency upon the proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology and held at the IAEA Headquarters in Vienna from 10 to 13 November 1986. Thirty participants from 17 countries and an international organization attended the meeting. Eighteen papers were presented from 13 countries and one international organization. The meeting was composed of four sessions and covered subjects related to: physico-chemical properties of core materials under off-normal conditions, and their interactions up to and after melt-down (5 papers); core materials deformation, relocation and core coolability under (severe) accident conditions (4 papers); fission products release: including experience, mechanisms and modelling (5 papers); power plant experience (4 papers). A separate abstract was prepared for each of these 18 papers. Four working groups covering the above-mentioned topics were held to discuss the present status of the knowledge and to develop recommendations for future activities in this field. Refs, figs and tabs

  11. Safety studies on heat transport and afterheat removal for GCR accident conditions

    The IAEA coordinated an international research program on 'Heat Transport and Afterheat Removal for GCRs under Accident Conditions (CRP-3)'. America, China, France, Germany, Japan, Netherlands and Russia participate the program. Final goal of the program is to show clearly to the world one of the most important salient features of the HTGR, that is the HTGR reactor can be cooled down by passive measures without causing any damage to the nuclear reactor system even in accidental conditions, and to make clear the boundaries (or restrictions) for the passive cooling regime. The first 5 year term of the coordinate program started in 1993 and established a goal to improve common knowledge for decay heat removal and to improve our tools, like computer codes and analytical models for the prediction of the performance of decay heat removal system. We are now performing benchmark problems for these purposes. The present efforts are concentrated on the benchmark for the passive heat removal performance outside the reactor vessel, partly because we have two different type of the HTGR in the world, the pebble bed type and the block type reactor. They have quite different heat dissipation behavior inside the reactor vessel. However, they have quite similar residual heat removal process outside the reactor vessel. For the first step of the international cooperation, we selected the common problem. After finishing the present benchmark we are planning to proceed to tackle the inside heat removal problem. (J.P.N.)

  12. Advances in fracture mechanics analyses of primary system performance under operating and accident conditions

    Safety research sponsored by the Nuclear Regulatory Commission, Division of Reactor Safety Research, has resulted in notable advances in several areas of importance in the safety evaluation of reactor primary systems under normal operations and accident situations. First, the methods of linear elastic fracture mechanics and of elastic plastic fracture mechanics have been validated for prediction of pressure vessel performance by the Intermediate Vessel Test program results at the Oak Ridge National Laboratory. The ability confidently to predict vessel performance under realistic service conditions has permitted development of the computer program OCTAVIA which computes failure curves for a range of flaw sizes in terms of pressure and temperature for specified presure vessel material at specific neutron fluence levels. It then considers the probability of occurrence of flaw sizes and magnitude of pressure during an operational, overpressurization transient and determines the probability of failure, for both individual flaw sizes and for the full spectrum. This advance has been verified by the confirmatory results of testing small thick-walled cylinders under thermal shock conditions in the Heavy Section Steel Technology program, and of warm prestressing tests at the US Navel Research Laboratory. Thirdly, the technology of crack arrest has reached a level wherein standardization of test specimens and testing methods is now possible and, indeed, is underway. (Auth.)

  13. Silicon Carbide Behaviour Under Prototypic LWR Chemistry/Neutron Flux and Accident Conditions

    The Accident Tolerant Fuels (ATF) programme was started with the goal of finding alternative fuel and cladding materials that perform as well as or better than current materials under normal operating conditions with enhanced performance during accidents. SiC has been a long-time candidate for this programme. Here we will discuss both out-of-core and in-core test results that are incorporated into analytical models developed at MIT. Out-of-core research has subjected un- irradiated SiC ceramic matrix composite (CMC) samples to a series of tests to gauge performance in a loss-of-coolant accident (LOCA). Previous work by the group with monolithic ot SiC has exposed samples to steam temperatures up to 1,500°C for 8 to 48 hours at various steam flow rates. Additionally, monolith samples have been quenched from up to 1,300°C into both room temperature and saturated water at atmospheric pressure. Following this testing, burst testing was performed to quantify strength degradation and ceramographic analysis was completed to determine microstructural effects. Using the same techniques as this previous monolithic work, testing of tube samples composed of three integrated layers have been tested. The three layers are an outer environmental barrier, a ceramic matrix composite, and an inner monolith layer; all layers are composed of SiC. These SiC/SiC CMC samples have been exposed to 1,400°C steam at ~6 g/min for 48 hours. Quench testing was performed from 1,200°C into saturated water to simulate the condition of reflood water after a LOCA, and burst and ceramographic analysis performed. Several different architectures of the CMC layer are in the process of being tested. Initial results for one CMC architecture indicate that the CMC samples exhibit the expected low oxidation rate of SiC in steam, and improved burst performance of both quenched and as-received samples. In the future two additional fibre architectures and sealed tubes will be examined, with the results being

  14. Thermal and hydraulic behaviour of CANDU cores under severe accident conditions

    This report describes work performed for the Atomic Energy Control Board on a) Formation and rewetting of dry patches on CANDU reactor calandria tubes during a Loss-of-Coolant Accident, and b) Analysis of accident sequence S11: Loss-of-Coolant Accident plus Loss-of-Emergency Core Cooling plus loss of moderator cooling system. For part (a), it is concluded that any dry patches which form on calandria tubes as a result of local heating to the critical heat flux will rewet in a short time (10 to 30 seconds for a Bruce-type reactor, 90 seconds for a Douglas Point-type reactor), with negligible effects on fuel sheath and maximum pressure tube temperatures. Pressure tube integrity is not predicted to be threatened. For part (b), preliminary analysis of the S11 accident sequence is presented. The complete analysis follows in the final report on the effects of severe accidents on CANDU cores

  15. Generation IV benchmarking of TRISO fuel performance models under accident conditions. Modeling input data

    Blaise Collin

    2014-09-01

    This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: the modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release; the modeling of the AGR-1 and HFR-EU1bis safety testing experiments; and, the comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from ''Case 5'' of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. ''Case 5'' of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to ''effects of the numerical calculation method rather than the physical model''[IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison

  16. Carbon monoxide - hydrogen combustion characteristics in severe accident containment conditions. Final report

    burning velocity is high for the range of mixtures relevant to containment accident conditions, the gap in knowledge is significant. - Large-scale data on combustion pressure development in closed and vented vessels is unavailable to validate predictions of combustion models applicable to CO-H2-H2O-CO2-air mixtures, resulting in significant uncertainties in predicted pressure loads from ignition. - Experimental data on the detonation cell sizes (detonability) of CO-H2 mixtures is unavailable to validate theoretical models. Since detonability is one aspect that appears sensitive to CO addition to the containment atmosphere, there are implications for reactor safety assessments. - Theoretical studies indicate that addition of steam and CO2 reduces the detonation sensitivity of CO-H2 mixtures (i.e., increases the cell widths) in agreement with experimental studies in H2,. - The effect of carbon dioxide addition on cell width appears to depend on hydrogen stoichiometry for lean hydrogen-air mixtures (the most relevant case) the cell size decreases as the CO concentration increases. For rich mixtures, the opposite is true. - The present results indicate that the cell widths for a hydrogen-carbon monoxide-air-steam mixture can be deduced from the measured (or calculated) cell widths for a corresponding hydrogen-air-steam mixture but supporting data in CO-H2 mixtures are lacking

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

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

  18. Aging, condition monitoring, and loss-of-coolant accident (LOCA) tests of class 1E electrical cables

    This report describes the results of aging, condition monitoring, and accident testing of ethylene propylene rubber (EPR) cables. Three sets of cables were aged for up to 9 months under simultaneous thermal (≅100 degrees C) and radiation (≅0.10 kGy/hr) conditions. A sequential accident consisting of high dose rate irradiation (≅6 kGy/hr) and high temperature steam followed the aging. Also exposed to the accident conditions was a fourth set of cables, which were unaged. The test results indicate that most properly installed EPR cables should be able to survive an accident after 60 years for total aging doses of at least 150--200 kGy and for moderate ambient temperatures on the order of 45--55 degrees C (potentially higher or lower, depending on material specific activation energies and total radiation doses). Mechanical measurements (primarily elongation, modulus, and density) were more effective than electrical measurements for monitoring age-related degradation

  19. Thermochemistry of Ruthenium Oxyhydroxide Species and Their Impact on Volatile Speciations in Severe Nuclear Accident Conditions.

    Miradji, Faoulat; Virot, François; Souvi, Sidi; Cantrel, Laurent; Louis, Florent; Vallet, Valérie

    2016-02-01

    Literature thermodynamic data of ruthenium oxyhydroxides reveal large uncertainties in some of the standard enthalpies of formation, motivating the use of high-level relativistic correlated quantum chemical methods to reduce the level of discrepancies. Reaction energies leading to the formation of all possible oxyhydroxide species RuOx(OH)y(H2O)z have been calculated for a series of reactions combining DFT (TPSSh-5%HF) geometries and partition functions, CCSD(T) energies extrapolated to the complete basis set limits. The highly accurate ab initio thermodynamic data were used as input data of thermodynamic equilibrium computations to derive the speciation of gaseous ruthenium species in the temperature, pressure and concentration conditions of severe nuclear accidents occurring in pressurized water reactors. At temperatures lower than 1000 K, gaseous ruthenium tetraoxide is the dominating species, between 1000 and 2000 K ruthenium trioxide becomes preponderant, whereas at higher temperatures gaseous ruthenium oxide, dioxide and even Ru in gaseous phase are formed. Although earlier studies predicted the formation of oxyhydroxides in significant quantities, the use of highly accurate ab initio thermodynamic data for ruthenium gaseous species leads to a more reliable inventory of gaseous ruthenium species in which gaseous oxyhydroxide ruthenium molecules are formed only in negligible amounts. PMID:26789932

  20. Investigation program on PWR-steel-containment behavior under accident conditions

    This report is a first documentation of the KfK/PNS activities and plans to investigate the behaviour of steel containments under accident conditions. The investigations will deal with a free standing spherical containment shell built for the latest type of a German pressurized water reactor. The diameter of the containment shell is 56 m. The minimum wall thickness is 38 mm. The material used is the ferritic steel 15MnNi63. According to the actual planning the program is concerned with four different problems which are beyond the common design and licensing practice: Containment behavior under quasi-static pressure increase up to containment failure. Containment behavior under high transient pressures. Containment oscillations due to earthquake loadings; consideration of shell imperfections. Containment buckling due to earthquake loadings. The investigation program consists of both theoretical and experimental activities including membrane tests allowing for very high plastic strains and oscillation tests with a thin-walled, high-accurate spherical shell. (orig.)

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

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

  2. Basic study on PWR plant behavior under the condition of severe accident

    In this paper, we report on the core cooling effect by natural circulation cooling of the primary cooling system in all core cooling function loss accidents caused by SBO in PWR plant compared with BWR. We also report on the core cooling effect by using air as the final heat sink in place of the seawater by opening the main steam valve of the steam generator. On the other hand, we discuss the behavior of PWR plant in the most serious case that the damage such as LOCA is caused by earthquake and that SBO due to the subsequent tsunami causes the reactor isolation and all function of reactor core cooling system loss. That is the case that LOCA and SBO occur in superimposed manner. We can show the results from the simulation experiments that, in PWR plant, even if it is fell into the reactor core cooling function loss due to SBO, natural circulation cooling can keep the reactor core cool down as long as the feed water is supplied to SG by the turbine-driven auxiliary feed-water pump and also that the cooling effect of even more is expected by ensuring the heat-pass to the atmosphere by opening the main steam valve. We also clarify the plant behaviors under the condition that LOCA and SBO occur in superimposed manner in PWR through the simulation experiments. (author)

  3. Perspectives on Severe Accident Management by Depressurization and External Water Injection under Extended SBO Conditions

    Three major issues of severe accident management guideline (SAMG) after this sort of extended SBO would be depressurization of the primary system, external water injection and hydrogen management inside a containment. Under this situation, typical SAM actions would be depressurization and external water delivery into the core. However, limited amount of external water would necessitate optimization between core cooling, containment integrity and fission product removal. In this paper, effects of SAM actions such as depressurization and external water injection on the reactor and containment conditions after extended SBO are analyzed using MAAP4 code. Positive and negative aspects are discussed with respect to core cooling and fission product retention inside a primary system. Conclusions are made as following: Firstly, early depressurization action itself has two-faces: positive with respect to delay of the reactor vessel failure but negative with respect to the containment failure and fission product retention inside the primary system. Secondly, in order to prevent containment overpressure failure after external water injection, re-closing of PORV later should be considered in SAM, which has never been considered in the previous SAMG. Finally, in case of external water injection, the flow rate should be optimized considering not only the cooling effect but also the long term fission product retention inside the primary system

  4. Behaviour of rock-like oxide fuels under reactivity-initiated accident conditions

    Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl2O4) homogeneous mixture and particle-dispersed YSZ/spinel - were conducted in the Nuclear Safety Research Reactor to investigate the fuel behaviour under reactivity-initiated accident conditions. The ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m3, which was comparable to that of un-irradiated UO2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO2 fuel. The ROX fuels failed with fuel pellet melting and a part of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below∼12 GJ/m3. The YSZ type and homogenous YSZ/spinel type ROX fuels failed by cladding burst when their temperatures peaked, while the particle-dispersed YSZ/spinel type ROX fuel seemed to have failed by cladding local melting. (author)

  5. Aerosol resuspension in the reactor cooling system of LWR's under severe accident conditions

    Aerosol resuspension from the pipes of the RCS under severe accident conditions happens when the carrier gas flow is turbulent. The origin of such phenomenon seems to be the existence of turbulent bursts in the neighbourhood of the pipe wall. These bursts are of random nature, in space and time. Three theoretical models have been found in available literature; those are: Cleaver and Yates', RESUS and Reeks' models. The first two of them are force balance models, in which particle detachment is supposed whenever aerodynamic lift or drag forces, respectively exceed adhesive forces, and the third one is an energy balance model in which resuspension happens when particle vibrational energy exceeds adhesive potential. From experimental evidence it seems that the studied phenomenon is a force balance problem and RESUS seems to be the most appropriate to it, among the available ones. Small-scale experiments have shown, as main parameters affecting resuspension, the Reynolds number of the flow, aerosol composition and initial loading per unit of area. Moreover, the resuspension rate decreases with time in all experiments where temporal measurements were taken

  6. Thermal and hydraulic behaviour of CANDU cores under severe accident conditions - final report. Vol. 1

    This report gives the results of a study of the thermo-hydraulic aspects of severe accident sequences in CANDU reactors. The accident sequences considered are the loss of the moderator cooling system and the loss of the moderator heat sink, each following a large loss-of-coolant accident accompanied by loss of emergency coolant injection. Factors considered include expulsion and boil-off of the moderator, uncovery, overheating and disintegration of the fuel channels, quenching of channel debris, re-heating of channel debris following complete moderator expulsion, formation and possible boiling of a molten pool of core debris and the effectiveness of the cooling of the calandria wall by the shield tank water during the accident sequences. The effects of these accident sequences on the reactor containment are also considered. Results show that there would be no gross melting of fuel during moderator expulsion from the calandria, and for a considerable time thereafter, as quenched core debris re-heats. Core melting would not begin until about 135 minutes after accident initiation in a loss of the moderator cooling system and until about 30 minutes in a loss of the moderator heat sink. Eventually, a pool of molten material would form in the bottom of the calandria, which may or may not boil, depending on property values. In all cases, the molten core would be contained within the calandria, as long as the shield tank water cooling system remains operational. Finally, in the period from 8 to 50 hours after the initiation of the accident, the molten core would re-solidify within the calandria. There would be no consequent damage to containment resulting from these accident sequences, nor would there be a significant increase in fission product releases from containment above those that would otherwise occur in a dual failure LOCA plus LOECI

  7. Radionuclide releases from UO2 and MOX fuel under severe accident conditions

    Radionuclide release from fuel under severe accident conditions was investigated in VEGA (Verification Experiments of radionuclides Gas/Aerosol release) program at Japan Atomic Energy Agency (JAEA). This study compares the results of tests on PWR-UO2 fuel, BWR-UO2 fuel and ATR (Advanced Thermal Reactor)-MOX(mixed oxide) fuel. The three types of fuels have burnup of 47, 56 and 43 GWd/t, respectively. Each fuel without cladding was set in a tungsten crucible and heated up to about 3130 K in helium atmosphere at 0.1 MPa. The fuel temperature was kept constant for 10 to 20 minuets at four plateaus during the heat up. The total fractional releases of high volatile Cs were 100% for the PWR-UO2 fuel, 97% for the BWR-UO2 fuel and 97% for the ART-MOX fuel. The Cs release with the heatup was different among three fuels for the temperature range below 2310 K, while the difference became small for the higher temperature range. The difference for the lower temperature range is considered to be caused by difference of irradiation histories, which varies migration states of the high volatile element. The total fractional releases of Mo and U were in the order of 0.1% and those of Sr and Pu were in the order of 1% both the tests with the BWR-UO2 and the ATR-MOX fuels. Release of low volatiles, U, Pu, Sr and Mo were dependent strongly on their chemical states, suggesting that vaporization was the controlling process. Namely, release of Pu and Sr was enhanced by the reduction of oxide, while it was largely decreased for Mo even at higher temperatures in the same atmosphere. (author)

  8. Behavior of small-sized BWR fuel under reactivity initiated accident conditions

    The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

  9. Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

    Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

  10. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

  11. Experimental analysis of heat transfer within the AP600 containment under postulated accident conditions

    The new AP600 reactor designed by Westinghouse uses a passive safety system relying on heat removal by condensation to keep the containment within the design limits of pressure and temperature. Even though some research has been done so far in this regard, there are some uncertainties concerning the behavior of the system under postulated accident conditions. In this paper, steam condensation onto the internal surfaces of the AP600 containment walls has been investigated in two scaled vessels with similar aspect ratios to the actual AP600. The heat transfer degradation in the presence of noncondensable gas has been analyzed for different noncondensable mixtures of air and helium (hydrogen simulant). Molar fractions of noncondensables/steam ranged from (0.4-4.0) and helium concentrations in the noncondensable mixture were 0-50% by volume. In addition, the effects of the bulk temperatures, the mass fraction of noncondensable/steam, the cold wall surface temperature, the pressure, noncondensable composition, and the inclination of the condensing surface were studied. It was found that the heat transfer coefficients ranged from 50 to 800 J s-1 K-1 m-2 with the highest for high wall temperatures at high pressure and low noncondensable molar fractions. The effect of a light gas (helium) in the noncondensable mixture were found to be negligible for concentrations less than approximately 35 molar percent but could result in stratification at higher concentrations. The complete study gives a large and relatively complete data base on condensation within a scaled AP600 containment structure, providing an invaluable set of data against which to validate models. In addition, specific areas requiring further investigation are summarized. (orig.)

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

    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.

  13. Simulations of the design basis accident at conditions of power increase and the o transient of MSIV at overpressure conditions of the Laguna Verde Power Station

    This document presents the analysis of the simulation of the loss of coolant accident at uprate power conditions, that is 2027 MWt (105% of the current rated power of 1931MWt). This power was reached allowing an increase in the turbine steam flow rate without changing the steam dome pressure value at its rated conditions (1020 psiaJ. There are also presented the results of the simulation of the main steam isolation va/ve transient at overpressure conditions 1065 psia and 1067 MWt), for Laguna Verde Nuclear Power Station. Both simulations were performed with the best estimate computer code TRA C BF1. The results obtained in the loss of coolant accident show that the emergency core coolant systems can recover the water level in the core before fuel temperature increases excessively, and that the peak pressure reached in the drywell is always below its design pressure. Therefore it is concluded that the integrity of the containment is not challenged during a loss of coolant accident at uprate power conditions.The analysis of the main steam isolation valve transients at overpressure conditions, and the analysis of the particular cases of the failure of one to six safety relief valves to open, show that the vessel peak pressures are below the design pressure and have no significant effect on vessel integrity. (Author)

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

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

  15. Mitigative techniques and analysis of generic site conditions for ground-water contamination associated with severe accidents

    The purpose of this study is to evaluate the feasibility of using ground-water contaminant mitigation techniques to control radionuclide migration following a severe commercial nuclear power reactor accident. The two types of severe commercial reactor accidents investigated are: (1) containment basemat penetration of core melt debris which slowly cools and leaches radionuclides to the subsurface environment, and (2) containment basemat penetration of sump water without full penetration of the core mass. Six generic hydrogeologic site classifications are developed from an evaluation of reported data pertaining to the hydrogeologic properties of all existing and proposed commercial reactor sites. One-dimensional radionuclide transport analyses are conducted on each of the individual reactor sites to determine the generic characteristics of a radionuclide discharge to an accessible environment. Ground-water contaminant mitigation techniques that may be suitable, depending on specific site and accident conditions, for severe power plant accidents are identified and evaluated. Feasible mitigative techniques and associated constraints on feasibility are determined for each of the six hydrogeologic site classifications. The first of three case studies is conducted on a site located on the Texas Gulf Coastal Plain. Mitigative strategies are evaluated for their impact on contaminant transport and results show that the techniques evaluated significantly increased ground-water travel times. 31 references, 118 figures, 62 tables

  16. Mitigative techniques and analysis of generic site conditions for ground-water contamination associated with severe accidents

    Shafer, J.M.; Oberlander, P.L.; Skaggs, R.L.

    1984-04-01

    The purpose of this study is to evaluate the feasibility of using ground-water contaminant mitigation techniques to control radionuclide migration following a severe commercial nuclear power reactor accident. The two types of severe commercial reactor accidents investigated are: (1) containment basemat penetration of core melt debris which slowly cools and leaches radionuclides to the subsurface environment, and (2) containment basemat penetration of sump water without full penetration of the core mass. Six generic hydrogeologic site classifications are developed from an evaluation of reported data pertaining to the hydrogeologic properties of all existing and proposed commercial reactor sites. One-dimensional radionuclide transport analyses are conducted on each of the individual reactor sites to determine the generic characteristics of a radionuclide discharge to an accessible environment. Ground-water contaminant mitigation techniques that may be suitable, depending on specific site and accident conditions, for severe power plant accidents are identified and evaluated. Feasible mitigative techniques and associated constraints on feasibility are determined for each of the six hydrogeologic site classifications. The first of three case studies is conducted on a site located on the Texas Gulf Coastal Plain. Mitigative strategies are evaluated for their impact on contaminant transport and results show that the techniques evaluated significantly increased ground-water travel times. 31 references, 118 figures, 62 tables.

  17. How to identify the key factors that affect driver perception of accident risk. A comparison between Italian and Spanish driver behavior.

    de Oña, Juan; de Oña, Rocio; Eboli, Laura; Forciniti, Carmen; Mazzulla, Gabriella

    2014-12-01

    Road crashes can be caused by different factors, including infrastructure, vehicles, and human variables. Many research studies have focused solely on identifying the key factors that cause road crashes. From these studies, it emerged that human factors have the most relevant impact on accident severity. More specifically, accident severity depends on several factors related directly to the driver, i.e., driving experience, driver's socio-economic characteristics, and driving behavior and attitudes. In this paper, we investigate driver behaviors and attitudes while driving and specifically focus on different methods for identifying the factors that most affect the driver's perception of accident risk. To this end, we designed and conducted a survey in two different European contexts: the city of Cosenza, which is located in the south of Italy, and the city of Granada, which is located in the south of Spain. Samples of drivers were contacted for their opinions on certain aspects of driving rules and attitudes while driving, and different types of questions were addressed to the drivers to assess their judgments of these aspects. Consequently, different methods of data analysis were applied to determine the aspects that heavily influence driver perception of accident risk. An experiment based on the stated preferences (SP) was carried out with the drivers, and the SP data were analyzed using an ordered probit (OP) model. Interesting findings emerged from different analyses of the data and from the comparisons among the data collected in the two different territorial contexts. We found that both Italian and Spanish drivers consider driving in an altered psychophysical state and violating the overtaking rules to be the most risky behaviors. PMID:25247553

  18. Development of Instrument Transmitter Protecting Device against High-Temperature Condition during Severe Accidents

    Min Yoo

    2014-01-01

    Full Text Available Reliable information through instrumentation systems is essential in mitigating severe accidents such as the one that occurred at the Fukushima Daiichi nuclear power plant. There are five elements which might pose a potential threat to the reliability of parameter detection at nuclear power plants during a severe accident: high temperature, high pressure, high humidity, high radiation, and missiles generated during the evolution of a severe accident. Of these, high temperature apparently poses the most serious threat, since thin shielding can get rid of pressure, humidity, radiation (specifically, alpha and beta radiations, and missile effects. In view of this fact, our study focused on designing an instrument transmitter protecting device that can eliminate the high-temperature effect on transmitters to maintain their functional integrity. We present herein a novel concept for designing such a device in terms of heat transfer model that takes into account various heat transfer mechanisms associated with the device.

  19. Study on safety evaluation for nuclear fuel cycle facility under fire accident conditions

    Hot test at Rokkasho Reprocessing plant has been started since last year. In addition, construction of the MOX fuel fabrication facility at Rokkasho site is planning. So, the importance of safety evaluation of the nuclear fuel cycle facility is increasing. Under the fire accident, one of the serious postulated accidents in the nuclear fuel cycle facility, the equipments (glove-box, ventilation system, ventilation filters etc.) for the confinement of the radioactive materials within the facility could be damaged by a large amount of heat and smoke released from the combustion source. Therefore, the fundamental data and models calculating for the amount of heat and smoke released from the combustion source under such accident are important for the safety evaluation of the facility. In JAERI, the study focused on the evaluation of amount of heat and smoke released from the combustion source is planning. In this paper, the outline of experimental apparatus, measurement items and evaluation terms are described. (author)

  20. New paradigm of research on the condition of spent nuclear fuel in accident and dumping sites

    For the first time we present the results of long-term experimental researches (1800 days) of the process of fission products (FP) of 235U release from SNF into the sea water and development of methodology of search of anomalies (plume) of distribution of gaseous fission products, released from SNF in zones of accident and dumping. Development of shipboard technology of 85 Kr and 3H measuring is made and a model of existential structure of distribution of a passive impurity taking into account local heterogeneities in concentration is developed. The received results for FP release from small samples (0.2 and 0.3 G of 235UO2) in 2002-2007 have shown: 1) Kinetics of release of gaseous FP (85Kr) from SNF strongly differs from kinetics of release others FP (137Cs). 85Kr release rate dozens of times exceeds those for 137Cs. Thus, it is proved, that 85Kr is the best indicator of the beginning of fuel rods cladding failure and the following corrosion process 235UO2. 2) Time of 85Kr-output from SNF fragments (16,7% from total saved up 85Kr for 1300 days of corrosion in sea water) allow us to propose reliable and rather inexpensive methodology of periodic (once for 2-3 years) monitoring of SNF condition on the sea bottom in an accident and dumping zone. The carried out stage of researches is the development of a new paradigm of sea radioecology based on a preliminary experimental research of kinetics of release FP (85Kr and 137Cs) from SNF, with the subsequent realization in expedition (including preliminary radionuclide measurements on a vessel) and modeling with use of a fractal formalism. Conservative estimation of risk to the population and sea environment using: 1) Experimental data about the release of fission products from the spent nuclear fuel. 2) The development of a ship low-level background complex (85Kr, tritium etc.) for operative measurements of radionuclide anomalies in the sea and hydrographic observation including current field of near-bottom water layer

  1. Fission product chemistry and aerosol behaviour in the primary circuit of a pressurised water reactor under severe accident conditions

    Three key accident sequences are considered covering a representative range of different environments of pressure, flow, temperature history and degree of zircaloy oxidation, and their principle thermal hydraulic and physical characteristics affecting chemistry behaviour are identified. Inventories, chemical forms and timing of fission product release are summarized together with the major sources of structural materials and their release characteristics. Chemistry of each main fission product species is reviewed from available experimental and/or theoretical data. Studies modelling primary circuit fission product behaviour are reviewed. Requirements for further study are assessed. (UK)

  2. Serpentine tube heat transfer characteristic under accident condition in gas cooled reactors

    In nuclear reactors of the Magnox or advanced gas Cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accident conditions using two phase flow codes requires knowledge of the heat transfer behavior of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear Electric. The tests were carried out on the Thermal Hydraulics Experimental Research Assembly (THERA) loop at Manchester University. The Thermal Hydraulic Experimental Research Assembly was designed to operate with pressures up to 180 bar and temperatures of 450degC. The geometry and dimensions of this test section were similar to part of a gas cooled reactor boiler of the Hinkley Point design. Blowdown from a pressure of 60 bar with subcoolings of 5degC, 50degC, 100degC formed the main part of the programme. A set of tests was conducted using discharge orifices of different sizes to produce depressurization times from 30 s to 10 mins, and in a few cases, the duration of blowdown approached 1 hour. These times were defined using the criterion of blowdown end as a final pressure of 10% of the initial pressure. Pressures, wall and fluid temperatures were all measured at average time intervals of 1.1s during the excursion and an inventory of the remaining water content in the serpentine was taken when the blowdown ended. Some tests were also conducted at an initial pressure of 30 bar. The results obtained show interesting stratification effects for the relatively fast discharge, with substantial wall circumferential temperature variations. For these tests, a relatively small water inventory remained after blowdown. The discharge characteristics of the serpentine in terms of orifice size have been mapped, and tests at 30 bar show the equivalence in terms of orifice size have been mapped

  3. Containment severe accident thermohydraulic phenomena

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

  4. Effect of Initial Coolant Temperature on Mechanical Fuel Failure under Reactivity-Initiated Accident Conditions

    A series of pulse irradiation tests, which simulated reactivity-initiated accidents (RIAs), were performed on high burnup light water reactor fuels at high temperature (HT) in the Nuclear Safety Research Reactor (NSRR). The NSRR tests with high burnup fuels have provided data of the fuel failure limit against the pellet-cladding mechanical interaction (PCMI) under RIA conditions, where the failure limit is quantified as the fuel enthalpy increase until the cladding failure. The failure limit depends on the cladding mechanical properties which are functions of cladding hydrogen content, cladding temperature and so on. Regarding the temperature condition, the previous NSRR experiments have been conducted at room temperature (RT) of ∼20 deg. C. Therefore, the obtained failure limits are suitable for the cold zero power condition, but could be very conservative for RIAs at hot zero power or at operation. In order to investigate the possible effect of initial coolant temperature on the PCMI failure limit, the NSRR HT test was launched using a newly developed test capsule, which can achieve coolant temperature up to ∼290 degrees C at the corresponding saturation pressure of ∼7 MPa. Three types of fuels were subjected to the tests; PWR fuel rods with ZIRLO cladding at a burnup of 71 GWd/t and with MDA cladding at 77 GWd/t, both of which were irradiated at the Vandellos 2 in Spain, and a BWR fuel rod with Zircaloy-2 (LK3) cladding, which was irradiated up to 69 GWd/t at the Leibstadt in Switzerland. For each fuel, two test rods were fabricated from an identical fuel segment to be used for the HT and RT tests. All the RT tests with the three fuels resulted in the PCMI failure at a similar level of fuel enthalpy, around 60 cal/g. Metallo-graphs of the failed claddings showed that the hydride precipitates at the cladding periphery, so-called hydride rim, played an important role of inducing cladding radial cracks which caused the stress concentration at crack tips and

  5. Nuclear power plant accident simulations of gasket materials under simultaneous radiation plus thermal plus mechanical stress conditions

    In order to probe the response of silicone door gasket materials to a postulated severe accident in an Italian nuclear power plant, compression stress relaxation (CSR) and compression set (CS) measurements were conducted under combined radiation (approximately 6 kGy/h) and temperature (up to 230 degrees C) conditions. By making some reasonable initial assumptions, simplified constant temperature and dose rates were derived that should do a reasonable job of simulating the complex environments for worst-case severe events that combine overall aging plus accidents. Further simplification coupled with thermal-only experiments allowed us to derive thermal-only conditions that can be used to achieve CSR and CS responses similar to those expected from the combined environments that are more difficult to simulate. Although the thermal-only simulations should lead to sealing forces similar to those expected during a severe accident, modulus and density results indicate that significant differences in underlying chemistry are expected for the thermal-only and the combined environment simulations. 15 refs., 31 figs., 15 tabs

  6. A simple mass and heat balance model for estimating plant conditions during the Fukushima Dai-ichi NPP accident

    A simple evaluation method for the analysis of thermal-hydraulic transients in reactor pressure vessel (RPV) and primary containment vessel (PCV) is proposed to support understanding the accident behaviors of the Fukushima Dai-ichi nuclear power plant (NPP). Since most of the measurements of the plants were unavailable especially in the early stage of the accident, and the accessibility to the plants had been limited by radiation, analytical investigation for the plant was required to understand the plant conditions such as the magnitude of the damages. In order to provide easy-to-use technical tools to support the analytical investigation, we developed a simplified analysis code, named 'HOTCB', based on total mass and heat balances in a lamped parameter system. The HOTCB code has capabilities to treat two-phase fluid including water, steam, and non-condensable gas in a wide range of temperatures up to highly superheated conditions, and to consider heat structures, i.e. heat capacities and heat transfer to the fluid. The code was provided to Tokyo Electric Power Company (TEPCO) and was practically used for the analysis on the accident. This paper provides the details of the code and simulations of Unit 1 and Unit 2 reactors of Fukushima Dai-ichi nuclear power plant (NPP) as examples to show the usefulness of the code. (author)

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

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

  8. Current status of research on FBR fuel behavior under accident conditions and the relevant NSRR program plan

    In the situation that the development of demonstration FBR is being materialized, a substantial research on safety of core fuels under accident conditions is required as the part of the research and development program. The experimental study of fuel integrity against over power accidents etc. and failure behavior is important to establish a criteria for safety evaluation of FBR's. In this report, the scope of the program which is planned in NSRR is shown after reviewing other related experiments and examining the research region left undone. Major in-core experiments on fuel failure are surveyed wide in the view point of experimental region and the inquired results are summarized. Subsequently, the items and methods due to the NSRR experiment program is discussed. The experimental facility plan and the results of preliminary analysis on the fuel energy deposition and temperature behavior are also introduced. (author)

  9. Preclosure radiological safety analysis for accident conditions of the potential Yucca Mountain Repository: Underground facilities

    This preliminary preclosure radiological safety analysis assesses the scenarios, probabilities, and potential radiological consequences associated with postulated accidents in the underground facility of the potential Yucca Mountain repository. The analysis follows a probabilistic-risk-assessment approach. Twenty-one event trees resulting in 129 accident scenarios are developed. Most of the scenarios have estimated annual probabilities ranging from 10-11/yr to 10-5/yr. The study identifies 33 scenarios that could result in offsite doses over 50 mrem and that have annual probabilities greater than 10-9/yr. The largest offsite dose is calculated to be 220 mrem, which is less than the 500 mrem value used to define items important to safety in 10 CFR 60. The study does not address an estimate of uncertainties, therefore conclusions or decisions made as a result of this report should be made with caution

  10. HTGR fuel elements operating conditions during accidents with abrupt power raise

    The necessity of the investigations for developing of HTGR fuel elements operability criteria, connected with the specific energy release values and the rates of its change in fuel is demonstrated in the paper on the example of the accident with positive reactivity increase at VGM reactor pebble bed compression as a result of seismic impact. It is shown, that the average fuel enthalpy over the core in this accident with the emergency protection failure may reach ∼24 Kj/g U02, and the maximum rate of its increase is about 0.14 Kj/g.s. It considerably exceeds the established limit of fuel enthalpy for LWR fuel elements. (author). 5 refs, 2 figs

  11. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided

  12. Study on chemical condition of transported radioactive materials from the Fukushima Daiichi Nuclear Power Plant accident

    A tremendous amount of radioactivity in the atmosphere was discharged due to Fukushima dai-ichi nuclear power plant accident. We collected from March 15 to May 15, 2011, on glass-cellulose fiber filters (HE-40T) and charcoal filters (CP-20) with low-volume air samplers at radioisotope center, University of Tsukuba. And we have been reported the time variation of the radioactivity concentrations of several radionuclides in atmosphere. In this work, we observed IP image of the glass-cellulose fiber filters after over 10 months at the accident. The IP image was showed clearly a lot of particulate matter. The divided filters were leached several solutions (dilute hydrochloric acid, dilute sodium oxalate aqueous solution, and distilled water), and were observed IP image and gamma-ray spectrometry. It is showed to be different solubility of filter samples at collected date. (author)

  13. Studies on the role of molybdenum on iodine transport in the RCS in nuclear severe accident conditions

    Highlights: • In oxidising conditions, Mo reacts with Cs and thus promotes gaseous iodine release. • In reducing conditions, CsI remains the dominant form for released iodine. • The nature of released iodine is well reproduced by the ASTEC code. - Abstract: The effect of molybdenum on iodine transport in the reactor coolant system (RCS) under PWR severe accident conditions was investigated in the framework of the EU SARNET project. Experiments were conducted at the VTT-Institute and at IRSN and simulations of the experimental results were performed with the ASTEC severe accident simulation code. As molybdenum affects caesium chemistry by formation of molybdates, it may have a significant impact on iodine transport in the RCS. Experimentally it has been shown that the formation of gaseous iodine is promoted in oxidising conditions, as caesium can be completely consumed to form caesium polymolybdates and is thus not available for reacting with gaseous iodine and leading to CsI aerosols. In reducing conditions, CsI remains the dominant form of iodine, as the amount of oxygen is not sufficient to allow formation of quantitative caesium polymolybdates. An I–Mo–Cs model has been developed and it reproduces well the experimental trends on iodine transport

  14. Proceedings of a specialist meeting on the behaviour of water reactor fuel elements under accident conditions

    The contributions of this meeting report experimental, numerical and research investigations on the oxidation behaviour of zircaloy in case of a loss-of-coolant accident (LOCA), analysis of the kinetics of the oxidation rate, very high temperature behaviour of fuel rod claddings (failure mechanics, ballooning), the interaction between cladding and fuel, the mechanical behaviour of zircaloy, etc. Numerous experimental and computer code analysis results are given

  15. Thermodynamic analysis of spent pyrochemical salts in the stored condition and in viable accident scenarios

    This study involves examining ''spent'' electrorefining (ER) salts in the form present after usage (as stored), and then after exposure to water in a proposed accident scenario. Additionally, the equilibrium composition of the salt after extended exposure to air was also calculated by computer modeling and those results are also presented herein. It should be noted that these salts are extremely similar to spent MSE salts from the Rocky Flats MSE campaigns using NaCl-KCl- MgCl2

  16. Preliminary assessment of the impact of candidate accident-tolerant fuels/cladding on the predicted reactor behaviour at normal operating conditions and under DB (LOCA and RIA) and BDB (STSBO and LTSBO) accident conditions

    Currently, the United States Department of Energy (DOE) has initiated the study of advanced accident-tolerant fuel/cladding (ATF) configurations that exhibit 1) slower reaction kinetics with steam, 2) lower enthalpy of oxidation, 3) less susceptibility to unfavourable core material interactions, and 4) provision of additional barriers to fission product release. Whenever changes, whether minor or major, are made to commercial NPP fuel/cladding systems; then the effect of these changes must be evaluated on all phases of the fuel/cladding lifetime (from fabrication through operation through eventual storage and reprocessing). This presentation focuses on preliminary assessments of several potential ATFs on the impact of these materials on predicted reactor behaviour 1) at normal operating conditions, 2) under postulated design basis (DB) accidents (LOCAs and RIAs), and 3) under beyond design basis (BDB) accident conditions [for short- and long-term station blackouts(SBO)]. These preliminary reactor response predictions are compared against the responses of UO2/Zr cores. For the ATFs evaluated, during normal operation, the most significant features are much lower fuel centerline temperatures and fission gas releases; and for LOCAs the peak cladding temperatures are lower with significantly lower hydrogen generation rates and for a RIA the ATF ejected worth is very similar to the UO2 ejected worth. The use of higher melting/lower hydrogen producing core components (ATFs) will not preclude a BDB accident. Without core cooling the severe accident will march-on; however, the ATFs do allow an increase in margin (time) to initiation of core component degradation - although this may be measured in minutes rather than hours. The ATF core responses (with oxidation kinetics about two orders of magnitude lower than that for Zr) are nearly the same as for components with no oxidation (for a STSBO, the increased time to vessel dry-out is approximately 4.5 hours). There is a need

  17. The scrubbing of fission product aerosols in LWR water pools under severe accident conditions

    The analysis of source teams for core damage accidents requires that the accident progression, the plant thermal hydraulic response and the associated radionuclide release and transport be analyzed for the identified escape pathways. These pathways are identified by examining the plant design, the events which constitute the hypothetical accident sequences, and the analyzed or assumed plant response. Many of the pathways currently identified for what are believed to be the risk significant sequences involve pathway segments through water pools. Specifically, transport through the BWR pressure suppression pool and the PWR quench tank represent examples of segments of the in-plant pathways for many risk significant sequences in these plants. In order to properly analyze the fission product transport for these sequences, it is therefore necessary to describe the effect of these water pools on the steam/non-condensible gas mixtures and associated entrained radionuclides. This paper describes Phase I of the experimental pool scrubbing work that is being conducted in conjunction with the development and validation of the SUPRA code

  18. Accident scenario diagnostics with neural networks

    Nuclear power plants are very complex systems. The diagnoses of transients or accident conditions is very difficult because a large amount of information, which is often noisy, or intermittent, or even incomplete, need to be processed in real time. To demonstrate their potential application to nuclear power plants, neural networks axe used to monitor the accident scenarios simulated by the training simulator of TVA's Watts Bar Nuclear Power Plant. A self-organization network is used to compress original data to reduce the total number of training patterns. Different accident scenarios are closely related to different key parameters which distinguish one accident scenario from another. Therefore, the accident scenarios can be monitored by a set of small size neural networks, called modular networks, each one of which monitors only one assigned accident scenario, to obtain fast training and recall. Sensitivity analysis is applied to select proper input variables for modular networks

  19. Development of simplified 1D and 2D models for studying a PWR lower head failure under severe accident conditions

    In the study of severe accidents of nuclear pressurized water reactors, the scenarios that describe the relocation of significant quantities of liquid corium at the bottom of the lower head are investigated from the mechanical point of view. In these scenarios, the risk of a breach and the possibility of a large quantity of corium being released from the lower head exist. This may lead to direct heating of the containment or outer vessel steam explosion. These issues are important due to their early containment failure potential. Since the TMI-2 accident, many theoretical and experimental investigations, relating to lower head mechanical behaviour under severe thermo-mechanical loading in the event of a core meltdown accident have been performed. IRSN participated actively in the one-fifth scale USNRC/SNL LHF and OECD LHF (OLHF) programs. Within the framework of these programs, two simplified models were developed by IRSN: the first is a simplified 1D approach based on the theory of pressurized spherical shells and the second is a simplified 2D model based on the theory of shells of revolution under symmetric loading. The mathematical formulation of both models and the creep constitutive equations used are presented in detail in this paper. The corresponding models were used to interpret some of the OLHF program experiments and the calculation results were quite consistent with the experimental data. The two simplified models have been used to simulate the thermo-mechanical behaviour of a 900 MWe pressurized water reactor lower head under severe accident conditions leading to failure. The average transient heat flux produced by the corium relocated at the bottom of the lower head has been determined using the IRSN HARAR code. Two different methods, both taking into account the ablation of the internal surface, are used to determine the temperature profiles across the lower head wall and their effect on the time to failure is discussed. Using these simplified models

  20. Khorasan wheat population researching (Triticum turgidum, ssp. Turanicum (McKey) in the minimum tillage conditions

    Ikanović Jela; Popović Vera; Janković Snežana; Živanović Ljubiša; Rakić Sveto; Dončić Dalibor

    2014-01-01

    Khorasan wheat occupies a special place in the group of new-old cereals (Triticum turgidum, ssp. Turanicum McKey). It is an ancient species, native to eastern Persia, that is very close to durum wheat by morphological characteristics. Investigations were carried out in agro ecological conditions of the eastern Srem, with two wheat populations with dark and bright awns as objects of study. The following morphological and productive characteristics were inves...

  1. Uncertainties under emergency conditions in Hiroshima and Nagasaki in 1945 and Bikini accident in 1954

    In exploding an atomic bomb, in addition to ionizing radiation, strong non-ionizing radiation, such as infrared, ultraviolet light, visible light, electromagnetic pulse radiation, as well as heat and shock waves are produced. The survivors and those who visited Hiroshima immediately after the atomic bombing could have been subjected to a number of other possible noxious effects in addition to atomic radiation. Hospitals, laboratories, drugstores, pharmaceutical works, storehouses of chemicals, factories, etc. that were situated close to the hypocenter were all completely destroyed and various mutagenic, carcinogenic or teratogenic substances must have been released, many doctors, nurses and chemists were killed. There was no medical care and no food in the region of high dose exposure and the drinking water was contaminated. There would have been various possibilities of infection. Mental stress would also have been much higher in the survivors closer to the hypocenter. It is confusing which factor played a dominant role. In addition, there would be problems in accurately identifying the position of the exposed persons at the time of the atomic bombing and also in estimating the shielding factors. There may be considerable uncertainty in human memory under such conditions. It is also possible that there could have been a large storage of gasoline to be used for transportation of the army corps in Hiroshima. Therefore there is a possibility that various toxic substances, mutagenic or carcinogenic agents such as benzopyrene and other radiomimetic substances, chemical weapons (Yperit, Lewisite, etc.) could have been released from various facilities which were destroyed at the time of the atomic bombing. After the German surrender, in May 1945, it was reported in June, in Japan, that the USA might attempt landing on Japan mainland, and that they might be planning massive use of chemical weapons all over Japan on that occasion. Preparing for such case chemical officers

  2. Uncertainties under emergency conditions in Hiroshima and Nagasaki in 1945 and Bikini accident in 1954

    Nishiwaki, Y. [Universitaet Wien, Institut fuer Medizinische Physik, Vienna (Austria); Kawai, H. [Atomic Energy Research Institute of Kinki Univ., Osaka (Japan); Shono, N. [Hiroshima Jogakuin Univ., Hiroshima (Japan); Fujita, S. [Radiation Effects Research Foundation, Department of Statistics, Hiroshima (Japan); Matsuoka, H. [Japan Atomic Energy Research Institute, Earth Simulator Research and Development Center, Tokyo (Japan); Fujiwara, S. [Japan Nuclear Cycle Development Institute, Ibaraki (Japan); Hosoda, T. [Chiyoda Technol Corporation, Tokyo (Japan)

    2000-05-01

    In exploding an atomic bomb, in addition to ionizing radiation, strong non-ionizing radiation, such as infrared, ultraviolet light, visible light, electromagnetic pulse radiation, as well as heat and shock waves are produced. The survivors and those who visited Hiroshima immediately after the atomic bombing could have been subjected to a number of other possible noxious effects in addition to atomic radiation. Hospitals, laboratories, drugstores, pharmaceutical works, storehouses of chemicals, factories, etc. that were situated close to the hypocenter were all completely destroyed and various mutagenic, carcinogenic or teratogenic substances must have been released, many doctors, nurses and chemists were killed. There was no medical care and no food in the region of high dose exposure and the drinking water was contaminated. There would have been various possibilities of infection. Mental stress would also have been much higher in the survivors closer to the hypocenter. It is confusing which factor played a dominant role. In addition, there would be problems in accurately identifying the position of the exposed persons at the time of the atomic bombing and also in estimating the shielding factors. There may be considerable uncertainty in human memory under such conditions. It is also possible that there could have been a large storage of gasoline to be used for transportation of the army corps in Hiroshima. Therefore there is a possibility that various toxic substances, mutagenic or carcinogenic agents such as benzopyrene and other radiomimetic substances, chemical weapons (Yperit, Lewisite, etc.) could have been released from various facilities which were destroyed at the time of the atomic bombing. After the German surrender, in May 1945, it was reported in June, in Japan, that the USA might attempt landing on Japan mainland, and that they might be planning massive use of chemical weapons all over Japan on that occasion. Preparing for such case chemical officers

  3. Assessing veld condition in the Kruger National Park using key grass species

    W.S.W. Trollope

    1989-10-01

    Full Text Available Veld condition refers to the condition of the vegetation in relation to some functional characteristic. In the Kruger National Park important functional characteristics are the potential of the veld to produce grass forage and fuel and to resist soil erosion. Consequently a simplified technique based on 18 key grass species was developed for assessing veld conditon and monitoring the effects of wild life management practices like veld burning, development of watering points and culling. The technique has been specifically developed for use by wildlife managers and has the ability to indicate the potential of the veld to support bulk grazing animals, to carry a fire and to resist soil erosion.

  4. Which key properties controls the preferential transport in the vadose zone under transient hydrological conditions

    Groh, J.; Vanderborght, J.; Puetz, T.; Gerke, H. H.; Rupp, H.; Wollschlaeger, U.; Stumpp, C.; Priesack, E.; Vereecken, H.

    2015-12-01

    % arrival time and potential key soil properties, site factors and boundary conditions will be presented in order to identify key properties which control the preferential transport in the vadose zone under transient hydrological conditions.

  5. Coolability of corium debris under severe accident conditions in light water reactors

    Rahman, Saidur

    2013-01-01

    The debris bed which may be formed in different stages of a severe accident will be hot and heated by decay heat from the radioactive fission products. In order to establish a steady state of long-term cooling, this hot debris needs to be quenched at first. If quenching by water ingression into the dry bed is not rapid enough then heat-up by decay heat in still dry regions may again yield melting. Thus, chances of coolability must be investigated considering quenching against heat-up due to d...

  6. Criticality safety assessment of a TRIGA reactor spent fuel pool under accident conditions

    An overview paper on the criticality safety analysis of a pool type storage for a TRIGA spent fuel at the ''Jozef Stefan'' Institute in Ljubljana, Slovenia, is presented. It was shown in that subcriticality is not guaranteed for some postulated accidents (an earthquake with subsequent fuel rack disintegration resulting in contact fuel pitch). To mitigate this deficiency, a study was made about replacing a certain number of fuel elements in the rack with absorber rods in order to lower the probability for supercriticality to acceptable level. (author)

  7. Study on light water reactor fuel behavior under reactivity initiated accident condition in TREAT

    This report reviews the results of the fuel failure experiments performed in TREAT in the U.S.A. simulating Reactivity Initiated Accidents. One of the main purposes of the TREAT experiments is the study of the fuel failure behavior, and the other is the study of the molten fuel-water coolant interaction and the consequent hydrogen behavior. This report mainly shows the results of the TREAT experiments studying the fuel failure behavior in Light Water Reactor, and then it describes the fuel failure threshold and the fuel failure mechanism, considering the results of the photographic experiments of the fuel failure behavior with transparent capsules. (author)

  8. ANL proposal for investigaton of concepts for hydrogen control under LWR accident conditions

    A proposal for investigation of concepts for hydrogen control in degraded core LWR accidents which could release significant quantities of hydrogen to the containment building is presented. This proposed work includes studies on the combustion suppression mechanisms which would involve reaction kinetics experiments with addition of chemicals or ions to scavenge the intermediate (free-radical or other) combustion products. Several additives appear promising. Also several condensed-phase system concepts presently appear attractive for the longer-term hydrogen removal systems. The work would be complementary to other current programs on hydrogen control

  9. Fission product release from UO/sub 2/ under LWR accident conditions: recent data compared with review values

    Studies of fission product release from commercial LWR fuel at temperatures up to 20000C in steam have shown that >50% of the Kr, I, and Cs may be released within 20 min. These data are in fairly good agreement with the results of a previous NRC review, but the influences of specific test/accident conditions other than temperature (which is the most important variable) on the behavior of these and other fission products are apparent. In particular, chemical effects related to the extent of cladding oxidation may dominate, as in the case of tellurium. 14 refs., 5 figs., 2 tabs

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

    NONE

    2004-07-01

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

  11. Aging, condition monitoring, and loss-of-coolant accident (LOCA) tests of Class 1E electrical cables: Summary of results

    This paper summarizes the results of aging, condition monitoring, and accident testing of various nuclear power plant cable products. Four sets of cables were aged under simultaneous thermal (≅95C) and radiation (≅0.10 kGy/hr) conditions. One set of cables was aged for 3 months, a second set was aged for 6 months, a third set was aged for 9 months, and a fourth set was not aged. A sequential accident consisting of high dose rate irradiation (≅6 kGy/hr) and high temperature steam was then performed on each set of cables. The results of the tests indicate that the feasibility of life extension of some popular cable products is promising. Mechanical measurements, primarily elongation, modulus, and density, were more effective than electrical measurements for monitoring age-related degradation. The broad objectives of this experimental program were twofold: (a) to determine the life extension potential of popular cable products used in nuclear power plants and (b) to determine the potential of condition monitoring for residual life assessment

  12. Behavior of LWR/MOX Fuels under Reactivity-Initiated Accident Conditions

    Fuketa, Toyoshi; Sugiyama, Tomoyuki; Umeda, Miki; Sasajima, Hide; Nagase, Fumihisa [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Ibaraki-ken, 319-1195 (Japan)

    2009-06-15

    Utilization of MOX fuels in power-producing light water reactors (LWRs) in Japan is now firmly scheduled, although it is behind its originally aimed time. The Nuclear and Industrial Safety Agency in Ministry of Economy, Trade and Industry (METI/NISA) licensed the MOX fuel loading up to one-fourth of their cores in four PWRs (Takahama Units 3 and 4, Genkai Unit 3, and Ikata Unit3) and that up to one-third in four BWRs (Fukushima-First Unit 3, Kashiwazaki-Kariwa Unit 3, Hamaoka Unit 4, Shimane Unit 2), and the Nuclear Safety Commission (NSC) of Japan reviewed those licensing. In this situation, safety of the MOX utilization in LWRs is one of primary concerns in the country. In order to provide a data base for the regulatory guide of power-producing light water reactors (LWRs) and to proceed in the processes with a better understanding, behavior of LWR fuels during reactivity-initiated accident (RIA) is being studied with pulse-irradiation experiments in the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Agency (JAEA). To subject LWR/MOX fuels to the experiments, JAEA shipped high burnup MOX fuels from the Beznau NPP in Switzerland to JAEA-Tokai. The tests BZ-1 and BZ-2 were performed on 14x14 PWR-MOX fuel rods with Zircaloy-4 cladding, which were irradiated in the Beznau NPP. The differences in test conditions between the two tests are the pellet producing process, fuel burnup, oxide thickness and the peak fuel enthalpy to be reached in case of non-failure. The BZ-1 test fuel rod contained pellets produced with the Short Binder-less Route (SBR) process. The local burnup was 48 GWd/t. The cladding oxide thickness was approximately 30 {mu}m and the hydrogen content was evaluated as 340 ppm. On the other hand, the pellets of the BZ-2 test fuel rod were produced with the Micronized Master blend (MIMAS) process. The local burnup was 59 GWd/t. The cladding oxide thickness was about 20 {mu}m and the hydrogen content was 160 ppm. As the BZ-2 test fuel rod

  13. Reactivity analysis of a Savannah River Site reactor under severe accident conditions

    An analysis of the reactivity changes in a Savannah River Site reactor tritium-producing charge during a postulated severe fuel damage accident has been performed. Possible in- and ex-vessel configurations were recriticality could occur have been identified and analyzed using Monte Carlo techniques. The results of the analyses indicate that recriticality is possible if fuel debris collects within the assembly bottom end-fittings (BEFs) in a postulated accident scenario where moderator is retained in the vessel. All other credible debris configurations identified were found to be subcritical. In the BEF, recriticality is possible only if the target melt fraction is less than 70% and moderator is present in the vessel. Given that recriticality in the BEF occurred, the resulting power transient was analyzed using point kinetics coupled with a linear feedback kernel. The calculated final debris temperatures suggest the potential for a fluid coolant interaction following recriticality; however, no aluminum vapor production is predicted to occur. The sensitivity of the final debris temperature to initial debris temperature, target melt fraction, reactivity insertion rate (i.e., fuel melt rate), and initial neutron power were included in the evaluation

  14. Radionuclide release under specific accident conditions for the Zion plant - BMI-2104 update

    The possibility of radioactive material being released to the environment from LWRs has long been the impetus for considerable concern and research. Most reactors in the United States were designed, and their sites were chosen, on the basis of research report TID-14844. Published in 1962, TID-14844 makes certain assumptions about the release of fission products to the reactor containment area during a hypothetical severe accident. Although these assumptions are representative of the state of knowledge at the time, the behavior of fission products has become better understood in the intervening years. Accordingly, the Nuclear Regulatory Commission conducted the Reactor Safety Study to reassess the accident risks in U.S. commercial nuclear power plants. The report of that study, known as WASH-1400, was published in 1975 and provided a more comprehensive and physically accurate description of fission product behavior. The amount of fission product release to the environment (the source term) estimated in WASH-1400 has since been used extensively in planning and evaluating reactor operations

  15. Fission-product transport and retention in the PHTS under accident conditions

    The current CANDU safety analysis methodology for predicting the release of radionuclides into containment is based on the bounding assumption that fission products released from the fuel go directly into containment. Allowing for FP retention in the PHTS will help achieve the following objectives: (1) improved estimates of doses to safety equipment in environmental qualification (EQ) analyses, (2) improved estimates of public and operator doses from an improved assessment of less volatile radionuclide behaviour, (3) improved ability to perform best-estimate safety analyses, (4) improved post-accident management plans from a better knowledge of FP location, and (5) less restrictive exclusion area boundary (EAB) designs from better source term estimates. Two LWR fission-product behaviour codes, VICTORIA and SOPHAEROS, have been assessed for their ability to provide a CANDU PHTS FP transport and retention modelling capability. The assessment of VICTORIA and SOPHAEROS was conducted by comparing the features of the two codes with the requirements for CANDU PHTS fission-product transport software, and performing simulations representative of the Loss-of-Coolant Accident with additional Loss of Emergency Coolant Injection (LOCA/LOECI) and stagnation feeder break scenarios with both codes. Based on this assessment, SOPHAEROS is better suited for simulating fission-product transport and retention in the PHTS for CANDU safety and licensing analysis, and VICTORIA should be retained to support more detailed calculations and R and D activities. (author)

  16. Radiation conditions in the Oryol region territory impacted by radioactive contamination caused by the Chernobyl NPP accident

    G. L. Zakharchenko

    2016-01-01

    Full Text Available Research objective is retrospective analysis of radiation conditions in the Oryol region during 1986- 2015 and assessment of efficacy of the carried out sanitary and preventive activities for population protection against radiation contamination caused by the Chernobyl NPP accident.Article materials were own memoirs of events participants, analysis of federal state statistic surveillance forms 3-DOZ across the Oryol region, f-35 “Data on patients with malignant neoplasms, f-12 “Report on MPI activities”. Risk assessment of oncological diseases occurrence is carried out on the basis of AAED for 1986- 2014 using the method of population exposure risk assessment due to long uniform man-made irradiation in small doses. Results of medical and sociological research of genetic, environmental, professional and lifestyle factors were obtained using the method of cancer patients’ anonymous survey. Data on "risk" factors were obtained from 467 patients hospitalized at the Budgetary Health Care Institution of the Oryol region “Oryol oncology clinic”; a specially developed questionnaire with 60 questions was filled out.The article employs the method of retrospective analysis of laboratory and tool research and calculation of dose loads on the Oryol region population, executed throughout the whole period after the accident.This article provides results of the carried out laboratory research of foodstuff, environment objects describing the radiation conditions in the Oryol region since the first days after the Chernobyl NPP accident in 1986 till 2015.We presented a number of activities aimed at liquidation of man-caused radiation accident consequences which were developed and executed by the experts of the Oryol region sanitary and epidemiology service in 1986-2015. On the basis of the above-stated one may draw the conclusions listed below. Due to interdepartmental interaction and active work of executive authorities in the Oryol region, the

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

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

    2010-07-01

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

  18. Experiments to quantify airborne release from packages with dispersible radioactive materials under accident conditions

    Martens, R.; Lange, F. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Schwertnergasse 1, 50667 Koeln (Germany); Koch, W.; Nolte, O. [Fraunhofer-Institut fuer Toxikologie und Experimentelle Medizin (ITEM), Nikolai-Fuchs-Str.1, 30625 Hannover (Germany)

    2005-07-01

    For transport or handling accidents involving packages with radioactive materials and the assessment of potential radiological consequences, for the review of current requirements of the IAEA Transport Regulations, and for their possible further development reliable release data following mechanical impact are required. Within this context a research project was carried out which extends the basis for a well-founded examination of the contemporary system of requirements of 'Low Specific Activity' (LSA)-type materials and allows for its further development where appropriate. This project comprises a prior system-analytical examination and an experimental programme aiming at improving the general physical understanding of the release process as well as the quantity and the characteristics of airborne released material for non-fixed dispersible LSA-II material upon mechanical impact. Impaction experiments applying small, medium and real sized specimens of different dispersible materials revealed that the release behaviour of dispersible powders strongly depends upon material properties, e.g. particle size distribution and cohesion forces. The highest experimentally determined release fraction of respirable mass (AED < 10 {mu}m) amounted to about 2 % and was obtained for 2 kg of un-contained easily dispersible pulverized fly ash (PFA). For larger un-contained PFA specimen the release fraction decreases. However, packaging containing powdery material substantially reduces the airborne release fraction. The measured airborne release fractions for a 200 l drum with Type A certificate containing PFA were about a factor of 50 to 100 lower than for un-contained material. For a drop height of 9 m the airborne release fraction amounted to about 4 x 10{sup -5}. This value should be applicable for most of transport and handling accidents with mechanical impact. For a metal container of Type IP-2 or better which contains powder masses of 100 kg or more this release

  19. Methodological procedures for the description of the behaviour of primary circuit components under severe accident conditions

    Comprehensive work is being carried out for extending our understanding of severe accidents in the framework of the research programme on reactor safety supported by the German ministry BMBF as well as the reactor safety research programmes of the EU. A number of different individual projects are being worked on to better describe load, stress and material behaviour with regard to the question of the load carrying capacity of structures of the pressure bearing containment. There are reports on what has been discovered to date, with special emphasis on what improvements have been achieved in extending the material database and describing structural behaviour. Furthermore, information is given on experiments presently planned for verification.(orig.)

  20. The reaction between iodine and organic coatings under severe PWR accident conditions. An experimental parameter study

    Hellmann, S.; Funke, F.; Greger, G.U.; Bleier, A.; Morell, W. [Siemens AG, Power Generation Group, Erlangen (Germany)

    1996-12-01

    An extensive experimental parameter study was performed on the deposition and on the resuspension kinetics in the reaction system iodine/organically coated surfaces. Both reactions in the gas phase and in the liquid phase were investigated and kinetic rate constants suitable for modelling were derived. Previous experimental studies on the reaction of iodine with organic coated surfaces were mostly limited to temperatures below 100{sup o}C. Thus, this parameter study aims at filling a gap and providing kinetic data on heterogeneous reactions with organic surfaces in the accident-relevant temperature range of 100-160{sup o}C. Two types of laboratory experiments carried out at Siemens/KWU using coatings representative for German power plants (epoxy-tape paint), namely gas phase tests and liquid phase tests. (author) 6 figs., 6 tabs., 5 refs.

  1. Formation and characterization of fission-product aerosols under postulated HTGR accident conditions

    The paper presents the results of an experimental investigation on the formation mechanism and physical characterization of simulated nuclear aerosols that could likely be released during an HTGR core heat-up accident. Experiments were carried out in a high-temperature flow system consisting essentially of an inductively heated release source, a vapor deposition tube, and a filter assembly for collecting particulate matter. Simulated fission products Sr and Ba as oxides are separately impregnated in H451 graphite wafers and released at elevated temperatures into a dry helium flow. In the presence of graphite, the oxides are quantitatively reduced to metals, which subsequently vaporize at temperatures much lower than required for the oxides alone to vaporize in the absence of graphite. A substantial fraction of the released material is associated with particulate matter, which is collected on filters located downstream at ambient temperature. The release and transport of simulated fission product Ag as metal are also investigated

  2. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    Heames, T.J. (Science Applications International Corp., Albuquerque, NM (USA)); Williams, D.A.; Johns, N.A.; Chown, N.M. (UKAEA Atomic Energy Establishment, Winfrith (UK)); Bixler, N.E.; Grimley, A.J. (Sandia National Labs., Albuquerque, NM (USA)); Wheatley, C.J. (UKAEA Safety and Reliability Directorate, Culcheth (UK))

    1990-10-01

    This document provides a description of a model of the radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident. This document serves as the user's manual for the computer code called VICTORIA, based upon the model. The VICTORIA code predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating. This document provides a detailed description of each part of the implementation of the model into VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided. The VICTORIA code was developed upon a CRAY-XMP at Sandia National Laboratories in the USA and a CRAY-2 and various SUN workstations at the Winfrith Technology Centre in England. 60 refs.

  3. Health conditions among workers who participated in the cleanup of the Chernobyl accident

    People who took part in the Chernobyl accident cleanup have been registered upon their return to Kyrgyzstan since 1991, and their children since 1992. Later, citizens affected by the Semipalatinsk and Chelyabinsk contamination incidents were included for registration and health care purposes. The effects of the nuclear waste depositories in the Mailuu-Suu region were examined with the assistance of the Kansas University Medical Center (United States of America). All these investigations of affected people indicated apparent increases in a number of symptoms and illnesses when compared to the rest of the population. Samples sizes ranged from several hundred to several thousand. Above-normal radiation levels and/or the stress and fear of living in contaminated areas can lead to significant increases in nervous disorders, cardiovascular diseases and other problems. The most significant increase was in the suicide rate. 6 refs, 2 figs, 1 tab

  4. The reaction between iodine and organic coatings under severe PWR accident conditions. An experimental parameter study

    An extensive experimental parameter study was performed on the deposition and on the resuspension kinetics in the reaction system iodine/organically coated surfaces. Both reactions in the gas phase and in the liquid phase were investigated and kinetic rate constants suitable for modelling were derived. Previous experimental studies on the reaction of iodine with organic coated surfaces were mostly limited to temperatures below 100oC. Thus, this parameter study aims at filling a gap and providing kinetic data on heterogeneous reactions with organic surfaces in the accident-relevant temperature range of 100-160oC. Two types of laboratory experiments carried out at Siemens/KWU using coatings representative for German power plants (epoxy-tape paint), namely gas phase tests and liquid phase tests. (author) 6 figs., 6 tabs., 5 refs

  5. Radiolysis of cesium iodide solutions in conditions prevailing in a pressurized water reactor severe accident

    Measurements were made of I/sub 2/ formed when aqueous cesium iodide (CsI) solutions were exposed to two temperatures, 43 and 950C, with irradiation. Iodine partition coefficients were obtained from the experiments. The parameters varied were dose, CsI concentration, and Cs/sub 2/CO/sub 3/ concentration, in the presence of air-carbon dioxide and air-carbon dioxide-hydrogen mixtures, to provide information to calculate the form in which iodine released from fuel as CsI in a reactor accident might reach the environment. In a series of experiments, a two-compartment cell was used to trap the gaseous iodine produced. In this case, it was found that the quantity of gaseous iodine produced increased approximately linearly with the dose (at the dose rate used)

  6. Fuel pin behaviour under conditions of control rod withdrawal accident in CABRI-2 experiments

    Simulation of the control rod withdrawal accident has been performed in the international CABRI-2 experimental programme. The tests realized with industrial pins led to clarification of the influence of the pellet design and have shown the important role of fission products on the solid fuel swelling which promotes early pin failure with solid fuel pellet. With annular pellet design, large fuel swelling combined to low smear density leads to degradation of fuel thermal conductivity and thus reduces power to melt. However, the high margin to deterministic failure is confirmed with hollow pellets. Improvements of the modelling were necessary to describe such behaviours in computer codes as SAS-4A, PAPAS-2S and PHYSURAC. (author)

  7. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    This document provides a description of a model of the radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident. This document serves as the user's manual for the computer code called VICTORIA, based upon the model. The VICTORIA code predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating. This document provides a detailed description of each part of the implementation of the model into VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided. The VICTORIA code was developed upon a CRAY-XMP at Sandia National Laboratories in the USA and a CRAY-2 and various SUN workstations at the Winfrith Technology Centre in England. 60 refs

  8. Modelling of plate-out under gas-cooled reactor (GCR) accident conditions

    The importance of plate-out in mitigating consequences of gas-cooled reactor accidents, and its place in assessing these consequences, are discussed. The data requirements of a plate-out modelling program are discussed, and a brief description is given of parallel work programs on thermal/hydraulic reactor behaviour and fuel modelling, both of which will provide inputs to the plate-out program under development. The representation of a GCR system used in SRD studies is presented, and the equations governing iodine adsorption, desorption and transport round the circuit are derived. The status of SRD's plate-out program is described, and the type of sensitivity studies to be undertaken with the partially-developed computer program in order to identify the most useful lines for future research is discussed. (author)

  9. Effectiveness and adverse effects of reactor coolant system depressurization strategy with various severe accident management guidance entry conditions for OPR1000

    Severe accident analysis for Korean OPR1000 with MELCOR 1.8.6 was performed by adapting a mitigation strategy under different entry conditions of Severe Accident Management Guidance (SAMG). The analysis was focused on the effectiveness of the mitigation strategy and its adverse effects. Four core exit temperatures (CETs) were selected as SAMG entry conditions, and Small Break Loss of Coolant Accident (SBLOCA), Station Blackout (SBO), and Total Loss of Feed Water (TLOFW) were selected as postulated scenarios that may propagate into severe accidents. In order to delay reactor pressure vessel (RPV) failure, entering the SAMG when the CET reached 923 K, 923 K, and 753 K resulted in the best results for SBLOCA, SBO, and TLOFW scenarios, respectively. This implies that using event-based diagnosis for severe accidents may be more beneficial than using symptom-based diagnosis. There is no significant difference among selected SAMG entry conditions in light of the operator's available action time before the RPV failure. Potential vulnerability of the RPV due to hydrogen generation was analyzed to investigate the foreseeable adverse effects that act against the accident mitigation strategies. For the SBLOCA cases, mitigation cases generated more hydrogen than the base case. However, the amount of hydrogen generated was similar between the base and mitigation cases for SBO and TLOFW. Hydrogen concentrations of containment were less than 5% before RPV failure for most cases. (author)

  10. Creep failure of a reactor pressure vessel lower head under severe accident conditions

    A severe accident in a nuclear power plant could result in the relocation of large quantities of molten core material onto the lower head of he reactor pressure vessel (RPV). In the absence of inherent cooling mechanisms, failure of the RPV ultimately becomes possible under the combined effects of system pressure and the thermal heat-up of the lower head. Sandia National Laboratories has performed seven experiments at 1:5th scale simulating creep failure of a RPV lower head. This paper describes a modeling program that complements the experimental program. Analyses have been performed using the general-purpose finite-element code ABAQUS-5.6. In order to make ABAQUS solve the specific problem at hand, a material constitutive model that utilizes temperature dependent properties has been developed and attached to ABAQUS-executable through its UMAT utility. Analyses of the LHF-1 experiment predict instability-type failure. Predicted strains are delayed relative to the observed strain histories. Parametric variations on either the yield stress, creep rate, or both (within the range of material property data) can bring predictions into agreement with experiment. The analysis indicates that it is necessary to conduct material property tests on the actual material used in the experimental program. The constitutive model employed in the present analyses is the subject of a separate publication

  11. Performance evaluation of control room HVAC and air cleaning systems under accident conditions

    In light water reactors, control rooms and technical support centers must be designed to provide habitable environments in accordance with the requirements specified in General Design Criterion 19 of Appendix A, 10 CFR Part 50. Therefore, the effectiveness of HVAC and air cleaning system designs with respect to plant operator protection has to be evaluated by the system designer. Guidance for performing the analysis has been previously given in ANSI/ASME N509-1980 as well as in presentations at past Air Cleaning Conferences. The previous work is extended and the methodology used in a generic, interactive computer program that performs Main Control Room and Technical Support Center (TSC) habitability analyses for LWR nuclear power plants is presented. For given accident concentrations of radionuclides or hazardous gases in the outdoor air intakes and plant spaces surrounding the Main Control Room (or TSC), the program models the performance of the HVAC and air cleaning system designs, and determines control room (or TSC) contaminant concentrations and plant operator protection factors. Calculated or actual duct leakage, air cleaning efficiency, and airborne contamination are taken into account. Flexibility of the model allows for the representation of most control rooms (or TSC) and associated HVAC and air cleaning system conceptual designs that have been used by the US architect/engineers. The program replaced tedious calculations to determine the effects of HVAC ductwork and equipment leakage and permits (1) parametric analyses of various HVAC system design options early in the conceptual phase of a project, and (2) analysis of the effects of leakage test results on contaminant room concentrations, and therefore operator doses

  12. Experimental investigations of BWR pressure suppression pool behavior under loss of coolant accident conditions

    The experiments discussed in this paper look into different processes which may occur during a loss-of-coolant accident in the pressure suppression pool of a Boiling Water Reactor (BWR). These processes include: a) development of a thermal stratification, b) bubble dynamics and related water flow during continuous release of air and c) air blowdown and associated water slug phenomenon in the water pool. The experiments have been performed in the THAI test facility, which is a cylindrical vessel of 9.2 m height, 3.2 m diameter and with a gas volume of 60 m3. The variation in the investigated test parameters included, steam and air mass flux, initial water pool temperature, blowdown pressures, downcomer submergence, etc. A systematic variation of the test parameters allowed better understanding of the phenomena. Experiments discussed in this paper were performed with a vertical downcomer of 0.1 m diameter and 2 m submergence depth in the water pool. For the blowdown experiments, a separate interconnecting vessel of 1 m3 volume was used to inject air at pressures between 3 bar and 10 bar. A high speed camera (1000 fps) was installed to visualize the formation and propagation of air bubbles in the suppression pool and the resulting pool swelling phenomena. Customized instrumentation applied during the tests included grids of densely spaced thermocouples and of pressure transducers at various locations in order to capture the temperature distribution in the pool and the water slug induced pressure loadings, respectively. The present paper discusses the main outcome of the selected experiments. On the whole the experimental data may be very useful for code validation. (authors)

  13. Study of Air Ingress Across the Duct During the Accident Conditions

    Hassan, Yassin [Texas A & M Univ., College Station, TX (United States)

    2013-05-06

    The goal of this project is to study the fundamental physical phenomena associated with air ingress in very high temperature reactors (VHTRs). Air ingress may occur due to a rupture of primary piping and a subsequent breach in the primary pressure boundary in helium-cooled and graphite-moderated VHTRs. Significant air ingress is a concern because it introduces potential to expose the fuel, graphite support rods, and core to a risk of severe graphite oxidation. Two of the most probable air ingress scenarios involve rupture of a control rod or fuel access standpipe, and rupture in the main coolant pipe on the lower part of the reactor pressure vessel. Therefore, establishing a fundamental understanding of air ingress phenomena is critical in order to rationally evaluate safety of existing VHTRs and develop new designs that minimize these risks. But despite this importance, progress toward development these predictive capabilities has been slowed by the complex nature of the underlying phenomena. The combination of inter-diffusion among multiple species, molecular diffusion, natural convection, and complex geometries, as well as the multiple chemical reactions involved, impose significant roadblocks to both modeling and experiment design. The project team will employ a coordinated experimental and computational effort that will help gain a deeper understanding of multiphased air ingress phenomena. This project will enhance advanced modeling and simulation methods, enabling calculation of nuclear power plant transients and accident scenarios with a high degree of confidence. The following are the project tasks: Perform particle image velocimetry measurement of multiphase air ingresses; and, Perform computational fluid dynamics analysis of air ingress phenomena.

  14. Study of Air Ingress Across the Duct During the Accident Conditions

    The goal of this project is to study the fundamental physical phenomena associated with air ingress in very high temperature reactors (VHTRs). Air ingress may occur due to a rupture of primary piping and a subsequent breach in the primary pressure boundary in helium-cooled and graphite-moderated VHTRs. Significant air ingress is a concern because it introduces potential to expose the fuel, graphite support rods, and core to a risk of severe graphite oxidation. Two of the most probable air ingress scenarios involve rupture of a control rod or fuel access standpipe, and rupture in the main coolant pipe on the lower part of the reactor pressure vessel. Therefore, establishing a fundamental understanding of air ingress phenomena is critical in order to rationally evaluate safety of existing VHTRs and develop new designs that minimize these risks. But despite this importance, progress toward development these predictive capabilities has been slowed by the complex nature of the underlying phenomena. The combination of inter-diffusion among multiple species, molecular diffusion, natural convection, and complex geometries, as well as the multiple chemical reactions involved, impose significant roadblocks to both modeling and experiment design. The project team will employ a coordinated experimental and computational effort that will help gain a deeper understanding of multiphased air ingress phenomena. This project will enhance advanced modeling and simulation methods, enabling calculation of nuclear power plant transients and accident scenarios with a high degree of confidence. The following are the project tasks: Perform particle image velocimetry measurement of multiphase air ingresses; and, Perform computational fluid dynamics analysis of air ingress phenomena.

  15. Condition of organ of vision and free radical process parameters in liquidators of the Chernobyl accident

    84 liquidators of consequences of Chernobyl APS accident from the age of 28 to 58 were examined. The control group was made with 22 men from the age of 28 to 52. A certain increase of infringement of a transparency of lens without typical attributes of radiating cataract is revealed in the experimental group. Electrophysiological investigation (EPI) shows a certain reduction of amplitude of a wave ''a'' of macular electroretinogram (ERG) on green stimulus, amplitude of a main component and lengthening of an interpeak time interval of flicker ERG 10 Hz is revealed. These changes indicate the tendency to reduction of functional activity of a retina (first of all at a level of photoreceptors) in paramacular and in a smaller degree in peripheral zones among liquidators. The parameters of contrast sensitivity are definitely reduced in the experimental group for all stimuli on all spatial frequencies. Luminous and colour sensitivity to stimuli of different colour in the experimental group is definitely reduced in all central field of sight, but in paracentral zone the degree of reduction is higher. We investigated the parameters of oxidative stress in both groups. Definite increase of production of the reactive oxygen species and disbalance of a glutathione link of antioxidant protection are revealed. Authentic correlation dependences are revealed: moderate direct correlation - between a level of glutathione reductase and amplitude of a main component of flicker ERG 10 Hz, between a level of oxidized glutathione and interpeak time interval of flicker ERG 10 Hz, inverse correlation - between the level of oxidized glutathione and amplitude of a main component of flicker ERG 10 Hz. In view of large spontaneous activity of free radical processes in a retina in norm the received results can explain revealed changes of an organ of vision. (author)

  16. CANSWEL-2: a computer model of the creep deformation of Zircaloy cladding under loss-of-coolant accident conditions

    The CANSWEL-2 code models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised water reactor (PWR). It considers in detail the centre rod of a 3 x 3 nominally square array, taking into account azimuthal non-uniformities in cladding thickness and temperature, and the mechanical restraint imposed on contact with neighbouring rods. Any of the rods in the array may assume a non-circular shape. Models are included for primary and secondary creep, dynamic phase change and superplasticity when both alpha- and beta-phase Zircaloy are present. A simple treatment of oxidation strengthening is incorporated. Account is taken of the anisotropic creep behaviour of alpha-phase Zircaloy which leads to cladding bowing. The CANSWEL-2 model is used both as a stand-alone code and also as part of the LOCA analysis code MABEL-2. (author)

  17. Reassessment of fuel failure behavior in the SPERT and PBF experiments for irradiated fuel rods under reactivity initiated accident conditions

    The current safety guideline for the evaluation of postulated reactivity initiated events in light water reactors was established by the Nuclear Safety Commission in January, 1984 on the basis of the experimental results from the NSRR program using fresh fuels. As for the burnup effects on fuel failure, the results of the previous American SPERT-CDC experiments were considered in the guideline. However, failure threshold and failure mechanism for preirradiated fuel rods were not established because only a few irradiated fuel rods were tested. Experiments with preirradiated fuel rods are now in progress as the next major research items in the NSRR program. This paper presents behavior of fuel failure for irradiated fuel rods under reactivity initiated accident conditions. Results from the previous SPERT and PBF experiments which should be compared with the experiments of the NSRR program are reviewed. The modes of fuel failure in the SPERT and PBF experiments were different from those in the experiments with fresh fuels. Cladding rupture and PCMI failure came out in the SPERT experiments, Cladding rupture in the SPERT experiments might be related to a FP gas release during both preirradiation and power burst. The rod with burnup of 31,800 MWd/t and total energy of 190 cal/g·UO2 in the SPERT experiments failed at low energy deposition (85 cal/g·UO2) with PCMI. The observed cracks appeared to be brittle fractures along the whole active length of the rod. The failure of this ROd was probably related to the cladding embrittlement by the excessive corrosion during preirradiation. Moreover, relationship between supposed failure mechanisms and influencing factor for generally irradiated fuel rod under reactivity initiated accident conditions is discussed. (author)

  18. Recent condition of Fukushima-Daiichi nuclear plant accident in Japan

    Ohnishi, Takeo

    2012-07-01

    Japanese government pronounced that the second step had been succeeded in the cooling down of the reactors on the middle of Dec 2011 at Fukushima-Daiichi nuclear power plant. In future, government aims to take out fuels from 4 reactors and shields their units. The nuclear power plants in Japan are gradually decreasing, because the checking for them has been performed and the permission of the re-start of them are difficult to be gained. On January 1st 2012, only 7 units are operating in Japan, though the about 54 units were set before the accident. At the end of December 2011, most radiations are emitted from cesium. The radioactivity in air and land around the plant was daily reported in newspaper. Government often gave the information about some RI-contamination in foods. They were taken off from the markets. At now stage, the most important project is the decontamination of radioactive materials from houses, schools, public facilities and industries. Government will newly classify three evacuation areas from April 2012. At the end of March, evacuees under 20 mSv/year possibly can go back their homes (evacuation-free area). The environmental doses will be depressed by decontamination under 10 mSv/year. At the range of 20-50 mSv, people will be controlled to live these area, they can go back their houses temporally (evacuation area). Over 50 mSv/year, however, people can go back house until 5 years at least (prohibited area). In new radiation limitation for a risk of human health, government made 100 mSv and 20 mSv for life span for one year, respectively. The aim of decontamination was set up to 10 mSv for 1 year and 5 mSv for next stage. A target at school is under1 mSv for children. Government accepted a new severe limitation per1 Kg at four groups; milk of baby (100 Bq) and milk (100 Bq), drinking water (10 Bq) and food (100 Bq). Tokyo electric Power Company and government should pay the sufficient compensation to evacuees. In future, they should keep health

  19. CHF experiments under realistic severe accident condition for IVR-ERVC strategy

    This study describes critical heat flux (CHF) experiments using a 2-D curved test section with trisodium phosphate (TSP : Na3PO4) and boric acid (BA : H3BO5). The CHF values of TSP solution, BA solution, and TSP + BA solution were enhanced by as much as 50% for all experimental conditions except the condition of 150 mm radius with BA solution. The enhancement can be explained by wettability enhancement and decrease of bubble departure diameter. This CHF enhancement could provide additional thermal margin for the IVR-ERVC strategy. (author)

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

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

  1. Mechanisms of damage to the oxide layer of cladding of fuel rods under accident conditions like RI

    During reactivity initiated accident, the importance of cladding tube oxidation on its thermomechanical behavior has been investigated. After RIA tests in experimental reactors oxide damage including radial cracking and spallation of the outer oxide layer has been evidenced. This work aims at better understanding the key mechanisms controlling these phenomena. Laboratory air-oxidation of Zircaloy-4 cladding tubes has been performed at 470 C. SEM micrographs show that radial cracks are initiated from the outer surface of the oxide layer and propagated radially towards the oxide-metal interface. A model predicting the stress evolution within the oxide and the depth of crack has been developed and validated on literature tests and tests of this study. Ring compression tests were used for the experimental study of the oxide degradation under mechanical loading. Experimental data revealed three mechanisms: densification of the radial crack network, propagation of these radial cracks, branching and spallation of oxide fragments. The influence of the circumferential cracks, periodically distributed in the oxide layer, on the stress distribution in oxide fragments has been analysed using finite element modelling. The determining influence of these cracks on the maximum stress oxide fragments has been demonstrated. (author)

  2. Experimental results from containment piping bellows subjected to severe accident conditions. Volume 1, Results from bellows tested in 'like-new' conditions

    Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted under the sponsorship of the US Nuclear Regulatory Commission at Sandia National Laboratories. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of thirteen bellows have been tested, all in the 'like-new' condition. (Additional tests are planned of bellows that have been subjected to corrosion.) The tests showed that bellows are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage. The test data is presented and discussed

  3. Pre-conception counselling for key cardiovascular conditions in Africa: optimising pregnancy outcomes.

    Zühlke, Liesl; Acquah, Letitia

    2016-01-01

    The World Health Organisation (WHO) supports pre-conception care (PCC) towards improving health and pregnancy outcomes. PPC entails a continuum of promotive, preventative and curative health and social interventions. PPC identifies current and potential medical problems of women of childbearing age towards strategising optimal pregnancy outcomes, whereas antenatal care constitutes the care provided during pregnancy. Optimised PPC and antenatal care would improve civil society and maternal, child and public health. Multiple factors bar most African women from receiving antenatal care. Additionally, PPC is rarely available as a standard of care in many African settings, despite the high maternal mortality rate throughout Africa. African women and healthcare facilitators must cooperate to strategise cost-effective and cost-efficient PPC. This should streamline their limited resources within their socio-cultural preferences, towards short- and long-term improvement of pregnancy outcomes. This review discusses the relevance of and need for PPC in resource-challenged African settings, and emphasises preventative and curative health interventions for congenital and acquired heart disease. We also consider two additional conditions, HIV/AIDS and hypertension, as these are two of the most important co-morbidities encountered in Africa, with significant burden of disease. Finally we advocate strongly for PPC to be considered as a key intervention for reducing maternal mortality rates on the African continent. PMID:27213854

  4. Multi-rod burst test under a loss-of-coolant accident condition, (2)

    Multi-rod burst test No. 7806 was performed with a views to estimating the quantitative channel blockage caused by the ballooning of fuel assembly during a postulated LOCA. The test was conducted under conditions of initial internal pressure 20 kg/cm2, steam flow rate 0.4 g/cm2min and heating rate 90C/sec. Following results were obtained: (1) Internal pressure increased up to 28 kg/cm2 during the heat-up stage, the average burst pressure of 49 rods being measured to be about 26 kg/cm2. (2) Almost all ruptured claddings had relatively short ballooned region length expanded above 34%. The length ranged from 0 to 40 mm, being much shorter than those measured in other tests performed under different conditions. (3) Maximum channel blockage of the assembly(7 x 7) was measured to be 36.2%, while was 43.4% in the interior rods(5 x 5) which had relatively uniform temperature distribution in the radial direction of the rod. These values were also smaller than those measured in other tests. (author)

  5. Kinetics of iodine and cesium reactions in the CANDU reactor primary heat transport system under accident conditions

    Gas-phase reaction kinetics have been modelled for the release of cesium and iodine into steam and steam/hydrogen atmospheres. The conditions are those anticipated in a CANDU reactor fuel channel following some postulated loss-of-coolant accidents. A total of seventeen chemical species were used in the model, including all important cesium and iodine species. Reaction rate constants were taken from the literature, or calculated where possible, or estimated. The composition evolution of the system was calculated, following a burst release of cesium and iodine, as a function of total iodine and cesium concentrations, cesium/iodine release ratio, iodine release form (atomic I or CsI), fuel channel atmosphere, and radiolysis effects. In general, the calculation demonstrates that CsI and CsOH rapidly (-2 s) become the most important species in the system for virtually all conditions. Atomic I is found to be significant only for very low release concentrations, or for Cs:I ratios less than unity. The main body of the modelling was performed at 1000 K. Some calculations were also performed for a three-node temperature system - 1500 K, 1000 K and 750 K - with the fission products being transported from high to low temperature. Thus, a qualitative picture is provided of the evolution of the chemistry in the fuel channel as the fission products are swept out by the residual steam flow

  6. Strategies for operation of containment related ESFs in managing activity release to the environment during accident conditions

    In Indian PHWR design, a double containment concept with passive vapour suppression pool (to limit peak pressure) system has been adopted. In addition to it, various Engineered Safety Features (ESFs) have been incorporated to limit the release of radioactivity to the environment. They are: Reactor building emergency coolers for cooling which results in fast reduction of overpressure; Primary Containment Filtration and Pump Back System (PCFPBS) for reduction in iodine concentration inside RB atmosphere during post LOCA period; and, Primary Containment Controlled Discharge System (PCCDS) for the rapid reduction of over-pressure tail. Due to operation of secondary containment purge system, which maintain negative pressure in the annulus, the ground level release is negligibly small. However, if non- availability of negative pressure in secondary containment space is assumed, then operation of PCFPBS and PCCDS system reduces the ground level release significantly. In this situation, depending upon time of operation of the PCFPBS, it can effectively reduce the iodine release, both in stack level and ground level by trapping it in charcoal filters. It is seen that delay time of PCFPBS operation in conjunction with prevailing weather condition can be manipulated to reduce the effect of stack level release of iodine. In this paper the containment related ESFs used in Indian PHWR is discussed in brief and the effectiveness of operator actions and management strategies in actuation of the ESFs in reducing the activity release to environment (during postulated accident conditions) will be brought out. (author)

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

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

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

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

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

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

    2015-03-15

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

  10. Accidents - Chernobyl accident

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

  11. Retention of iodine and other airborne radionuclides in nuclear facilities during abnormal and accident conditions

    Extensive research efforts have been undertaken in the world scientific community advancing the status of systems to maintain high air cleaning efficiency under the extreme abnormal conditions. The IAEA Co-ordinated Research Programme to upgrade technology in the area started in 1983 on the recommendations of a previous programme and the development covering a five year term is described in this document. Research laboratories from ten Member States participated, Belgium, German Democratic Republic, Hungary, India and Yugoslavia for three years with Austria, Canada, Federal Republic of Germany, Republic of Korea and UK for lesser periods. Research co-ordination meetings were held in Belgium (1984), Canada (1986) and Hungary (1988). A separate abstract was prepared for each of the 9 presentations from experts from the above mentioned Member States who participated in this research programme. Refs, figs and tabs

  12. Microcomputer based program for predicting heat transfer under reactor accident conditions. Volume II

    A microcomputer based program called Heat Transfer Prediction Software (HTPS) has been developed. It calculates the heat transfer for tube and bundle geometries for steady state and transient conditions. This program is capable of providing the best estimated of the hot pin temperatures during slow transients for 37- and 28-element CANDU type fuel bundles. The program is designed for an IBM-PC AT/XT (or IBM-PC compatible computer) equipped with a Math Co-processor. The following input parameters are required: pressure, mass flux, hydraulic diameter, and quality. For the steady state case, the critical heat flux (CHF), the critical heat flux temperature, the minimum film boiling temperature, and the minimum film boiling heat flux are the primary outputs. With either the surface heat flux or wall temperature specified, the program determines the heat transfer regime and calculates the surface heat flux, wall temperature and heat transfer coefficient. For the slow transient case, the pressure, mass flux, quality, and volumetric heat generation rate are the time dependent input parameters are required to calculate the hot pin sheath temperatures and surface heat fluxes. A simple routine for generating properties has been developed for light water to support the above program. It contains correlations that have been verified for pressures ranging from 0.6kPa to 30 MPa, and temperatures up to 1100 degrees Celcius. The thermodynamic and transport properties that can be generated from this routine are: density, specific volume, enthalpy, specific heat capacity, conductivity, viscosity, surface tension and Prandtle number for saturated liquid, saturated vapour, subcooled liquid of superheated vapour. A software for predicting flow regime has also been developed. It determines the flow pattern at specific flow conditions, and provides a correction factor for calculating the CHF during partially stratified horizontal flow. The technical bases for the program and its structure

  13. Microcomputer based program for predicting heat transfer under reactor accident conditions. Volume I

    A microcomputer based program called Heat Transfer Prediction Software (HTPS) has been developed. It calculates the heat transfer for the tube and bundle geometries for steady state and transient conditions. This program is capable of providing the best estimated of the hot pin temperatures during slow transients for 37- and 28-element CANDU type fuel bundles. The program is designed for an IBM-PC AT/XT (or IBM-PC compatible computer) equipped with a Math Co-processor. The following input parameters are required: pressure, mass flux, hydraulic diameter, and quality. For the steady state case, the critical heat flux (CHF), the critical heat flux temperature, the minimum film boiling temperature, and the minimum film boiling heat flux are the primary outputs. With either the surface heat flux or wall temperature specified, the program determines the heat transfer regime and calculates the surface heat flux, wall temperatures and heat transfer coefficient. For the slow transient case, the pressure, mass flux, quality, and volumetric heat generation rate are the time dependent input parameters required to calculate the hot pin sheath temperatures and surface heat fluxes. A simple routine for generating properties has been developed for light water to support the above program. It contains correlations that have been verified for pressures ranging from 0.6kPa to 30 MPa, and temperatures up to 1100 degrees Celcius. The thermodynamic and transport properties that can be generated from this routine are: density, specific volume, enthalpy, specific heat capacity, conductivity, viscosity, surface tension and Prandtl number for saturated liquid, saturated vapour, subcooled liquid for superheated vapour. A software for predicting flow regime has also been developed. It determines the flow pattern at specific flow conditions, and provides a correction factor for calculating the CHF during partially stratified horizontal flow. The technical bases for the program and its

  14. FASTGRASS: A mechanistic model for the prediction of Xe, I, Cs, Te, Ba, and Sr release from nuclear fuel under normal and severe-accident conditions

    The primary physical/chemical models that form the basis of the FASTGRASS mechanistic computer model for calculating fission-product release from nuclear fuel are described. Calculated results are compared with test data and the major mechanisms affecting the transport of fission products during steady-state and accident conditions are identified

  15. Distribution of hydrogen within the HDR-containment under severe accident conditions. OECD standard problem. Final comparison report

    The present report summarizes the results of the International Standard Problem Exercise ISP-29, based on the HDR Hydrogen Distribution Experiment E11.2. Post-test analyses are compared to experimentally measured parameters, well-known to the analysis. This report has been prepared by the Institute for Reactor Dynamics and Reactor Safety of the Technical University Munich under contract with the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) which received funding for this activity from the German Ministry for Research and Technology (BMFT) under the research contract RS 792. The HDR experiment E11.2 has been performed by the Kernforschungszentrum Karlsruhe (KfK) in the frame of the project 'Projekt HDR-Sicherheitsprogramm' sponsored by the BMFT. Ten institutions from eight countries participated in the post-test analysis exercise which was focussing on the long-lasting gas distribution processes expected inside a PWR containment under severe accident conditions. The gas release experiment was coupled to a long-lasting steam release into the containment typical for an unmitigated small break loss-of-coolant accident. In lieu of pure hydrogen a gas mixture consisting of 15% hydrogen and 85% helium has been applied in order to avoid reaching flammability during the experiment. Of central importance are common overlay plots comparing calculated transients with measurements of the global pressure, the local temperature-, steam- and gas concentration distributions throughout the entire HDR containment. The comparisons indicate relatively large margins between most calculations and the experiment. Having in mind that this exercise was specified as an 'open post-test' analysis of well-known measured data the reasons for discrepancies between measurements and simulations were extensively discussed during a final workshop. It was concluded that analytical shortcomings as well as some uncertainties of experimental boundary conditions may be responsible for deviations

  16. Effects of temperature history during cooling process on cladding ductility reduction under lost of coolant accident conditions

    In a Loss-of-Coolant Accident (LOCA) of LWRs, heated fuel rods are quenched by water injected from the Emergency Core Cooling System. Existing analyses indicate that fuel rods are cooled slowly before the quench and the cooling rate at that stage depends on scenario of the accident. Microstructure can be different between fuel claddings quenched directly from high temperatures and those quenched after slow-cooling process. It has been suggested that ductility reduction is enhanced in the latter case. Although reduction in cladding ductility is important for reactor safety in a LOCA, the effects of cooling conditions on the cladding ductility reduction has not been sufficiently investigated and mechanism for the influence of cooling rate has not been clarified. In the present study, samples cut from non-irradiated 17x17-type Zircaloy-4 cladding tubes for PWRs were isothermally oxidized at 1373 and 1473 K, slowly cooled and then quenched, changing temperature at which quenching process is started and rate of the slow cooling. The oxidized and quenched samples were subjected to ring compression test in order to evaluate effects of the cooling conditions on the cladding ductility reduction. In addition, metallographic examination, Vickers hardness test and computer code analysis were conducted to examine the mechanisms. Due to oxidation, ZrO2 layer is formed at the cladding surface, oxygen-stabilized alpha phase layer is formed beneath the oxide layer, and oxygen concentration increases in the central metallic layer. As the metallic layer is cooled slowly from the isothermal oxidation temperatures, oxygen-rich α phase precipitates in the β phase layer. It was found out that reduction in cladding ductility strongly depends on area fraction of α phase region precipitated in the metallic layer. Both size and hardness of α-phase region are increased as the rate of the slow cooing decreases, while the area fraction is nearly constant. Accordingly, the reduction in

  17. Fission product aerosol removal test by containment spray under accident management conditions (3)

    In order to demonstrate the effective FP aerosol removal by containment spray under Japanese AM conditions, two system integral tests and two separate effect tests were carried out using a full-height simulation test facility. In case of PWR LOCA, aerosol concentration in the upper containment vessel decreased even under low spray flow rate. In case of BWR LOCA with water injection into RPV, the aerosol concentration in the entire vessel also decreased rapidly after aerosol supply stopping. In both cases, the removal rate estimated from the NUREG-1465 was coincided with test results. The aerosol washing effect by spray was confirmed to be predominant by conducting suppression chamber isolation test. It turned out that the effect of aerosol solubility and density on aerosol removal by spray was quite small by conducting insoluble aerosol injection test. After the modification of aerosol removal model by the spray and hygroscopic aerosol model in original MELCOR 1.8.4, calculated aerosol concentration transient in the containment vessel agreed well with the test data. (author)

  18. Fission product release in conditions of a spent fuel pool severe accident

    Full text: Depending on the residence time, fuel burnup, and fuel rack configuration, there may be sufficient decay heat for the fuel clad to heat up, swell, and burst in case of a loss of pool water. Initiating event categories can be: loss of offsite power from events initiated by severe weather, internal fire, loss of pool cooling, loss of coolant inventory, seismic event, aircraft impact, tornado, missile attack. The breach in the clad releases the radioactive gases present in the gap between the fuel and clad, what is called 'gap release'. If the fuel continues to heat up, the zirconium clad will reach the point of rapid oxidation in air. This reaction of zirconium and air, or zirconium and steam is exothermic. The energy released from the reaction, combined with the fuel's decay energy, can cause the reaction to become self-sustaining and ignite the zirconium. The increase in heat from the oxidation reaction can also raise the temperature in adjacent fuel assemblies and propagate the oxidation reaction. Simultaneously, the sintered UO2 pellets resulting from pins destroying are oxidized. Due to the self-disintegration of pellets by oxidation, fission gases and low volatile fission products are released. The release rate, the chemical nature and the amount of fission products depend on powder granulation distribution and environmental conditions. The zirconium burning and pellets self-disintegration will result in a significant release of spent fuel fission products that will be dispersed from the reactor site. (author)

  19. Multi-rod burst test under a loss-of coolant accident condition, (4)

    Multi-rod burst test of No.7808 bundle was performed in steam to estimate quantitative coolant flow channel restriction caused by the ballooning of zircaloy claddings in a fuel assembly during a LOCA transient in LWRs. The test was conducted under the condition that the initial internal pressure in each rod was 35kg/cm2 (RT) and the heating rate was 90C/s in steam with flow rate of 0.4g/cm2.min. The following results were obtained; (1) Maximum and burst pressures in rods were in the range 45 to 48kg/cm2 and 41 to 45kg/cm2, respectively. The burst temperature of cladding were estimated to be 850 to 8800C. (2) Axial portions of tubes with greater than 34% strain were observed in the range 0 to 40mm in most rod. The mean length was 19mm in the bundle. (3) The degree of maximum increase in cross-sectional area is 54.2% in the bundle(7 x 7) and 66.9% in the internal rods(5 x 5). (4) Maximum channel area restriction was 40.5% in the bundle(7 x 7) and 51.4% in the internal rods(5 x 5). (author)

  20. Understanding of Iodine Behavior Model Under Severe Accident Conditions in NPPs (I)

    Support from ISAAC code which is developed in Korea will be needed when a regulatory body (KINS) reviews Wolsong specific SAMG submitted by KHNP in late 2009. The SA research area can be divided into thermal hydraulic (T/H) and fission product (FP) areas and international evaluation shows an increasing opinion that past analysis has large uncertainty in the FP area. Using very small manpower (less than 0.5 MY for 3 year period (2007.3-2010.2)), this study reviewed EU results (BIP, ThAI) and ISAAC possibility for experiment simulation. As results, international and domestic technical status until early 2009 was investigated and input model modification based on these investigation was provided. Simulation possibility of EU experiments showed the Benchmark function in MAAP should be activated in ISAAC (by cooperation with MAAP developer, FAI) as a preliminary arrangement and independent model development reflecting those international results could be possible. This documentation will be referred from PHWR analysis for both source term and safety goal and will enhance the internal understanding for the necessity of iodine behavior study under SA conditions. FP (mainly iodine) model development in the next step using this study will contribute to keeping both improved quality and international competitiveness in ISAAC code

  1. [The rehabilitation under alpine conditions of the participants in the cleanup of the accident at the Chernobyl Atomic Electric Power Station who are ill with chronic bronchitis].

    Brimkulov, N N; Abdulina, A A; Davletalieva, N E; Bakirova, A N; Karamuratov, A; Mirrakhimov, M M

    1996-01-01

    24 patients exposed to low-dose radiation after the Chernobyl accident were examined before and after 24-day treatment of chronic bronchitis in the high-altitude rehabilitation center (3200 m above the sea level) in Tien Shan. Sanogenic alpine climate improved the patients' general condition, physical performance and lung ventilation, corrected compromised immunity. After high-altitude adaptation tracheobronchial inflammation alleviated, cytologic composition and surface activity of bronchoalveolar fluid returned to normal. Therefore, high-altitude treatment of Chernobyl accident victims with chronic bronchitis is effective and can be recommended for such patients. PMID:8744100

  2. Khorasan wheat population researching (Triticum turgidum, ssp. Turanicum (McKey in the minimum tillage conditions

    Ikanović Jela

    2014-01-01

    Full Text Available Khorasan wheat occupies a special place in the group of new-old cereals (Triticum turgidum, ssp. Turanicum McKey. It is an ancient species, native to eastern Persia, that is very close to durum wheat by morphological characteristics. Investigations were carried out in agro ecological conditions of the eastern Srem, with two wheat populations with dark and bright awns as objects of study. The following morphological and productive characteristics were investigated: plant height (PH, spike length (SH, number of spikelets per spike (NSS, absolute weight (AW and grain weight per spike (GW, seed germination (G and grains yield (YG. Field micro-experiments were set on the carbonate chernozem soil type on loess plateau in 2011 and 2012. Hand wheat sowing was conducted in early March with drill row spacing of 12 cm. The experiment was established as complete randomized block system with four replications. Tending crops measures were not applied during the growing season. Plants were grown without usage of NPK mineral nutrients. Chemical crop protection measures were not applied, although powdery mildew (Erysiphe graminis was appeared before plants spike formation in a small extent. The results showed that both populations have a genetic yield potential. In general, both populations manifested a satisfactory tolerance on lodging and there was no seed dispersal. Plants from bright awns population were higher, had longer spikes and larger number of spikelet’s per spike. However, plants from dark awns population had higher absolute weight and grains weight per spike, as well as grain yield per plant. Strong correlation connections were identified among the investigated characteristics. The determination of correlations, as well as direct and indirect affects, enabled easier understanding of the mutual relationships and their balancing in order to improve the yield per unit area. [Projekat Ministarstva nauke Republike Srbije, br. TR 31078 i br. TR 31022

  3. Monitoring transport conditions of key comparison travelling standards using a data logger. Experiences from key comparison CCAUV.U-K3.1

    Haller, Julian; Koch, Christian

    2015-12-01

    In the framework of the international key comparison CCAUV.U-K3.1, a data logger was used to monitor temperature, pressure, humidity, and acceleration during transportation of an artefact travelling between participating laboratories. From the recorded data, environmental conditions of different kinds of transportation have been investigated and corresponding recommendations for the safe and proper transfer of artefacts between laboratories could be deduced. Transportation by courier services bears the risk of strong mechanical shocks and exposure to comparably high or low temperatures due to inappropriate handling or storage and is thus only suitable for insensitive or well-packed artefacts. Quite low temperatures (T  ≈  5 °C) have been observed in the cargo area during flights, so that hand-carrying of an artefact with transportation in the passenger cabin during flights is recommended, particularly for temperature-sensitive artefacts. Significant decreases of the pressure (p  ≈  750 mbar) have been recorded both for transportation in the passenger cabin and in the cargo area. Air-tight packing is thus recommended for pressure-sensitive devices. In general, the suitability of a data logger to provide evidence of the transport conditions during a key comparison has been demonstrated and the use of such a device is recommended for all key comparisons. The data logger has also been successfully employed to validate the protection properties of the passively insulating packaging of the artefact against pressure and temperature changes.

  4. Cloud conditions for low atmospheric electricity during disturbed period after the Fukushima nuclear accident

    Yatagai, Akiyo; Yamauchi, Masatoshi; Ishihara, Masahito; Watanabe, Akira; Murata, Ken T.

    2016-04-01

    The vertical (downward) component of the atmospheric electric field, or potential gradient (PG) under cloud generally reflects the electric charge distribution in the cloud. The PG data at Kakioka, 150 km southwest of the Fukushima Dai-ichi Nuclear Power Plant (FNPP1) suggested that this relation can be modified when the radioactive dust was floating in the air, and the exact relation between the weather and this modification could lead to new insight in plasma physics in the wet atmosphere. Unfortunately the detailed weather data was not available above Kakioka (only the precipitation data was available). Therefore, estimation of the cloud condition during March 2011 was strongly needed. We have developed various meteorological information links (http://www.chikyu.ac.jp/akiyo/firis/) and original radar and precipitation data will be released from the page. Here we present various radar images that we have prepared for March 2011. We prepared three-dimensional radar reflectivity of the C-band radar of JMA in every 10 minutes over all Kanto Plain centered at Tokyo and Fukushima prefecture centered at Sendai. We have released images of each altitude (1km interval) for 15th - 16thand 21th March (http://sc-web.nict.go.jp/fukushima/). The vertical structure of the rainfall is almost the same at 4km with the surface and sporadic high precipitation is observed at 6 km height for 15-16th. While, generally precipitation pattern that is similar to the surface is observed at 5km height on 21th. On the other hand, an X-band radar centered at Fukushima university is also used to know more localized raindrop patterns at zenith angle of 4 degree. We prepared 10-minutes/120m mesh precipitation patterns for March 15th, 16th, 17th, 18th, 20th, 21th, 22th and 23th. Quantitative estimate is difficult from this X-band radar, but localized structure, especially for the rain-band along Nakadori (middle valley in Fukushima prefecture), that is considered to determine the highly

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

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

  6. Framework for accident management

    Accident management is an essential element of the Nuclear Regulatory Commission (NRC) Integration Plan for the closure of severe accident issues. This element will consolidate the results from other key elements; such as the Individual Plant Examination (IPE), the Containment Performance Improvement, and the Severe Accident Research Programs, in a form that can be used to enhance the safety programs for nuclear power plants. The NRC is currently conducting an Accident Management Program that is intended to aid in defining the scope and attributes of an accident management program for nuclear power plants. The accident management plan will ensure that a plant specific program is developed and implemented to promote the most effective use of available utility resources (people and hardware) to prevent and mitigate severe accidents. Hardware changes or other plant modifications to reduce the frequency of severe accidents are not a central aim of this program. To accomplish the outlined objectives, the NRC has developed an accident management framework that is comprised of five elements: (1) accident management strategies, (2) training, (3) guidance and computational aids, (4) instrumentation, and (5) delineation of decision making responsibilities. A process for the development of an accident management program has been identified using these NRC framework elements

  7. The radioecological risk of decommissioning of nuclear submarines. Possible accidents and normal conditions

    the expense of the given influence will not exceed E(-6) year-1. Situation with the formation of a constant stain of pollution of sea bottom owing to constant organized discharge of insignificant amounts radionuclides with products of corrosion from temporary storage of reactor compartments of decommissioned nuclear submarines is considered also. Initial data for account of concentration of radionuclide in a ground were the maximum concentration of products of corrosion in a ground 10 Bq/dm2 (6,64 Bq/kg). According to accounts, the total effective dose for critical group of the population at the expense of consumption a fish, caught in region of items long-duration storage, will not exceed E(-8) Sv. The maximum expected capacity of a dose for bentic hydrobionts (weighed in is thicker than marine plankton) will not exceed E(-2) Gy/day. Comparison of sizes of control concentration in water and ground, allowable volumetric activity of drainage waters and settlement sizes of the contents of radionuclides in drainage water disposal have allowed to make a conclusion that the examined conditions of functioning of the temporary storages of reactor compartments cannot resul in excess of radiologic capacity of water area. (author)

  8. Report of a consultants meeting on accidents during shutdown conditions for WWER nuclear power plants. Extrabudgetary programme on the safety of WWER NPPs

    The main objectives of the meeting were to exchange information on the operational occurrences, studies performed and countermeasures taken for the accidents during shutdown for WWERs, and to define the necessity and directions of the further activities which may promote the improvement of WWER safety under shutdown conditions. The consultants have discussed some aspects concerning vulnerability of safety functions during shutdown conditions, several steps required to performed accident analysis and selected operational aspects for shutdown conditions. The discussion was supported by an evaluation of selected operational occurrences. The consultants have agreed that the discussion during the meeting in major parts is relevant to all the WWER designs (i.e. WWER-1000, WWER-440/213 and WWER-440/230). As for the plant conditions, the consultants have agreed to bound the discussion mainly by the cold shutdown and refuelling modes. Refs, figs, tabs

  9. Influence of precipitations, buildings and over increase of radioactive emission in the population dose calculation under accident conditions

    The influence of precipitations is analyzed, as well as of buildings and emission over increase on the dosis produced on the population as a consequence of some postulated accidents in Atucha II nuclear power plant. The following conclusions were achieved: the calculations performed without considering the above mentioned effects are conservative, excluding the case in which the precipitation is very close to the emission source. In this case, the maximal difference observed was 20% for class C and 5% for class D, at 1 km from source, with a decreasing difference according to the distance. The doses calculated without building effect were approximately 25% greater than those calculated considering this effect for class E and 40% for class F, at 1 km from the source. The difference decreases with distance and increases with the stability of atmospheric conditions. This behaviour is also observed with the over increase effect. In this case, the maximal observed differences were of one order of magnitude for class E and three orders for class F, at 1 km from the source. (Author)

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

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

  11. Modelling the release of volatile fission product cesium from CANDU fuel under severe accident conditions using artificial neural networks

    An artificial neural network (ANN) model has been developed to predict the release of volatile fission products from CANDU fuel under severe accident conditions. The model was based on data for the release Of 134Cs measured during three annealing experiments (Hot Cell Experiments 1 and 2, or HCE- 1, HCE-2 and Metallurgical Cell Experiment 1, or MCE- 1) at Chalk River Laboratories. These experiments were comprised of a total of 30 separate tests. The ANN established a correlation among 14 separate input variables and predicted the cumulative fractional release for a set of 386 data points drawn from 29 tests to a normalized error, En, of 0.104 and an average absolute error, Eabs, of 0.064. Predictions for a blind validation set (test HCE2-CM6) had an En of 0.064 and an Eabs of 0.054. A methodology is presented for deploying the ANN model by providing the connection weights. Finally, the performance of an ANN model was compared to a fuel oxidation model developed by Lewis et al. and to the U.S. Nuclear Regulatory Commission's CORSOR-M. (author)

  12. Behavior of irradiated BWR fuel under reactivity-initiated-accident conditions. Results of tests FK-1, -2 and -3

    Boiling water reactor (BWR) fuel rods with burnups of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior during a reactivity initiated accident (RIA) at cold startup. BWR fuel segment rods of 8 x 8BJ (STEP I) type from the Fukushima Daiichi Nuclear Power Station Unit 3 were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g) within about 20 ms. The fuel cladding had enough ductility against the prompt deformation due to pellet cladding mechanical interaction. The plastic hoop strain reached 1.5% at the peak location. The cladding surface temperature locally reached about 600 degC. Recovery of irradiation defects in the cladding due to high temperature during the pulse irradiation was indicated via X-ray diffractometry. The amount of fission gas released during the pulse irradiation was from 3.1% to 8.2% of total inventory, depending on the peak fuel enthalpy and the normal operation conditions. (author)

  13. Role of accident analysis in development of severe accident management guidance for multi-unit CANDU nuclear power plants

    This paper discusses the role of accident analysis in support of the development of Severe Accident Management Guidance for domestic CANDU reactors. In general, analysis can identify what types of challenges can be expected during accident progression but it cannot specify when and to what degree accident phenomena will occur. SAMG overcomes these limitations by monitoring the actual values of key plant indicators that can be used directly or indirectly to infer the condition of the plant and by establishing setpoints beyond which corrective action is required. Analysis can provide a means to correlate observed post-accident plant behavior against predicted behaviour to improve the confidence in and quality of accident mitigation decisions. (author)

  14. Level 2 PSA to evaluate the performance of the DOEL 1 and 2 NPP containment under severe accident conditions

    D`Eer, A.; Boesmans, B.; Auglaire, M.; Wilmart, P. [Tractebel Energy Engineering, Brussels (Belgium); Moeyaert, P. [Electrabel, Brussels (Belgium)

    1997-12-31

    The objective of the Doel 1 and 2 level 2 PSA is to evaluate in probabilistic terms the performance of the containment for core damage scenarios. The progression of the severe accident and its load on the containment is assessed by means of a logical method referred to as the Accident Progression Event Tree (APET). The PSA level 2 analysis shows that the Doel 1 and 2 containment prevents early loss of containment integrity during a severe accident. This is due to the low contribution of containment bypasses and to the extremely low probability of early structural containment failures. The late containment ruptures are dominated by basemat melt-through, whereas the late containment leaks mainly result from static overpressurisation. Sensitivity calculations have been performed to assess the contribution of the different severe accident management (SAM) measures (eg. auto-catalytic hydrogen recombiners) to the reduction of the containment failure probability. (author).

  15. Level 2 PSA to evaluate the performance of the DOEL 1 and 2 NPP containment under severe accident conditions

    The objective of the Doel 1 and 2 level 2 PSA is to evaluate in probabilistic terms the performance of the containment for core damage scenarios. The progression of the severe accident and its load on the containment is assessed by means of a logical method referred to as the Accident Progression Event Tree (APET). The PSA level 2 analysis shows that the Doel 1 and 2 containment prevents early loss of containment integrity during a severe accident. This is due to the low contribution of containment bypasses and to the extremely low probability of early structural containment failures. The late containment ruptures are dominated by basemat melt-through, whereas the late containment leaks mainly result from static overpressurisation. Sensitivity calculations have been performed to assess the contribution of the different severe accident management (SAM) measures (eg. auto-catalytic hydrogen recombiners) to the reduction of the containment failure probability. (author)

  16. Accidents at work and living conditions among solid waste segregators in the open dump of Distrito Federal

    Maria da Graça Hoefel; Fernando Ferreira Carneiro; Leonor Maria Pacheco Santos; Muriel Bauerman Gubert; Elisa Maria Amate; Wallace dos Santos

    2013-01-01

    The work of recycling solid waste segregators allows a precarious livelihood, but triggers a disease process that exacerbates their health and well-being. This study aimed to estimate the prevalence of occupational accidents at the open dump in the Federal District and its associated factors. Most segregators have had an accident at work (55.5%), perceived the danger of their working environment (95.0%) and claimed they did not receive personal protective equipment (51.7%). Among other findin...

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

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

  18. Accidents at work and living conditions among solid waste segregators in the open dump of Distrito Federal

    Maria da Graça Hoefel

    2013-09-01

    Full Text Available The work of recycling solid waste segregators allows a precarious livelihood, but triggers a disease process that exacerbates their health and well-being. This study aimed to estimate the prevalence of occupational accidents at the open dump in the Federal District and its associated factors. Most segregators have had an accident at work (55.5%, perceived the danger of their working environment (95.0% and claimed they did not receive personal protective equipment (51.7%. Among other findings, 55.8% ate foods found in the trash, 50.0% experienced food insecurity at home and 44.8% received Bolsa Família. There was a statistically significant relationship between work accidents and perception of dangerous work environment, household food insecurity and the presence of fatigue, stress or sadness (p < 0.05. On the other hand, the fellowship between the segregators was associated with a lower prevalence of accidents (p < 0.006. Women are the majority of the segregators (56.5% and reported more accidents than men (p < 0.025. We conclude that the solid waste segregators constitute a vulnerable community, not only from the perspective of labor, but also from the social and environmental circumstances. To reverse this situation, effective implementation of the National Policy of Solid Wastes is imperative, in association with affirmative policies to grant economic emancipation for this population.

  19. A Comprehensive Test of General Strain Theory: Key Strains, Situational- and Trait-Based Negative Emotions, Conditioning Factors, and Delinquency

    Moon, Byongook; Morash, Merry; McCluskey, Cynthia Perez; Hwang, Hye-Won

    2009-01-01

    Using longitudinal data on South Korean youth, the authors addressed limitations of previous tests of general strain theory (GST), focusing on the relationships among key strains, situational- and trait-based negative emotions, conditioning factors, and delinquency. Eight types of strain previously shown most likely to result in delinquency,…

  20. TASAC a computer program for thermal analysis of severe accident conditions. Version 3/01, Dec 1991. Model description and user`s guide

    Stempniewicz, M.; Marks, P.; Salwa, K.

    1992-06-01

    TASAC (Thermal Analysis of Severe Accident Conditions) is computer code developed in the Institute of Atomic Energy written in FORTRAN 77 for the digital computer analysis of PWR rod bundle behaviour during severe accident conditions. The code has the ability to model an early stage of core degradation including heat transfer inside the rods, convective and radiative heat exchange as well as cladding interactions with coolant and fuel, hydrogen generation, melting, relocations and refreezing of fuel rod materials with dissolution of UO{sub 2} and ZrO{sub 2} in liquid phase. The code was applied for the simulation of International Standard Problem number 28, performed on PHEBUS test facility. This report contains the program physical models description, detailed description of input data requirements and results of code verification. The main directions for future TASAC code development are formulated. (author). 20 refs, 39 figs, 4 tabs.

  1. Neutronics and fuel performance evaluation of accident tolerant FeCrAl cladding under normal operation conditions

    Highlights: • Detailed comparison of monolithic and hybrid (coating + cladding) cladding design. • Cycle length can be matched by optimized FeCrAl cladding design for a PWR assembly. • Detailed fuel performance analysis of FeCrAl cladding under normal operation conditions. - Abstract: Neutronics and fuel performance analysis is done for enhanced accident tolerance fuel (ATF), with the Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON. The purpose is to evaluate the most promising ATF candidate material FeCrAl, which has excellent oxidation resistance, as fuel cladding under normal operational conditions. Due to several major disadvantages of FeCrAl coating, such as difficulty in fabrication, diametrical compression from reactor pressurization, coating spallation and inter diffusion with zirconium, a monolithic FeCrAl cladding design is suggested. To overcome the neutron penalty expected when using FeCrAl as cladding for current oxide fuel, an optimized FeCrAl cladding design from a detailed parametric study in literature is adopted, which suggests reducing the cladding thickness and slightly increasing the fuel enrichment. A neutronics analysis is done that implementing this FeCrAl cladding design in a Pressurized Water Reactor (PWR) single assembly. The results show that the PWR cycle length requirements will be matched, with a slight increase in total plutonium production. Fuel performance analysis with BISON code is carried out to investigate the effects with this FeCrAl cladding design. The results demonstrate that the application of FeCrAl cladding could improve performance. For example, the axial temperature profile is flattened. The gap closure is significantly delayed, which means the pellet cladding mechanical interaction is greatly delayed. The disadvantages for monolithic FeCrAl cladding are that: (1) fission gas release is increased; and (2) fuel temperature is increased, but the increase is less than 50 K even at

  2. Key Concepts for Estimating the Burden of Surgical Conditions and the Unmet Need for Surgical Care

    Bickler, Stephen; Ozgediz, Doruk; Gosselin, Richard; Spiegel, David; Hsia, Renee; Dunbar, Peter; McQueen, Kelly; Jamison, Dean; Weiser, Thomas Geoghegan

    2010-01-01

    Background: Surgical care is emerging as a crucial issue in global public health. Methodology is needed to assess the impact of surgical care from a public health perspective. Methods: A consensus opinion of a group of surgeons, anesthesiologists, and public health experts was established regarding the methodology for estimating the burden of surgical conditions and the unmet need for surgical care. Results: For purposes of analysis, we define surgical conditions as any disease state requirin...

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

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

  4. Analysis of radionuclide behavior in a BWR Mark-II containment under severe accident management condition in low pressure sequence

    In the Level 2 PSA program at INS/NUPEC, MELCOR1.8.3 is extensively applied to analyze radionuclide behavior of dominant sequences. In addition, the revised source terms provided in the NUREG-1465 report have been also discussed to examine the potential of the radionuclides release to the environment in the conventional siting criteria. In the present study, characteristics of source terms to the environment were examined comparing with results by the Hypothetical Accident (LOCA), NUREG-1465 and MELCOR1.8.3. calculation for a typical BWR with a Mark-II containment in order to assure conservatives of the Hypothetical Accident in Japan. Release fractions of iodine to the environment for the Hypothetical Accident and NUREG-1465, which used engineering models for predicting radionuclide behaviors, were about 10-4 and 10-6 of core inventory, respectively, while the best estimate MELCOR1.8.3 code predicted 10-9 of iodine to the environment. The present study showed that the engineering models in the Hypothetical Accident or NUREG-1465 have large conservatives to estimate source term of iodine to the environment. (author)

  5. Research on Key Techniques of Condition Monitoring and Fault Diagnosing Systems of Machine Groups

    WANG Yan-kai; LIAO Ming-fu; WANG Si-ji

    2005-01-01

    This paper describes the development of the condition monitoring and fault diagnosing system of a group of rotating machinery. The data management is performed by means of double redundant data bases stored simultaneously in both the analyzing server and monitoring client. In this way, high reliability of the storage of data is guaranteed. Condensation of trend data releases much space resource of the hard disk. Diagnosing strategies orientated to different typical faults of rotating machinery are developed and incorporated into the system. Experimental verification shows that the system is suitable and effective for condition monitoring and fault diagnosing for a rotating machine group.

  6. Experimental study results on the grounds for the behaviour of high-burnup fuel in pressurized water reactors under conditions of loss of coolant accidents

    The complex of experimental studies is performed with the aim to clarify the behaviour of WWER high-burnup (∼ 60 MW day rg-1 U) fuel elements under conditions of loss of coolant accidents. It is revealed that the gaseous fission products release increases from 15 to 100% with a temperature growth from 1000 to 2500 deg C. The study of oxidation kinetics for irradiated fuel cans of Zr-1% Nb and Eh-635 alloys shows that their oxidation at 1000 deg C is somewhat faster compared to unirradiated cans with a temperatures increase above 1200 deg C this distinction disappears. Short-term mechanical properties of irradiated and unirradiated fuel cans do not differ when the temperature increases up to typical for accident values (> 600 deg C). Brittle fracture in fuel cans is not observed at the temperature below 100 deg C. The temperature dependence of time to fracture is built for the conditions of fuel can loading with excess pressure which is typical for a design basis loss of coolant accident

  7. Analyzing the causation of a railway accident based on a complex network

    In this paper, a new model is constructed for the causation analysis of railway accident based on the complex network theory. In the model, the nodes are defined as various manifest or latent accident causal factors. By employing the complex network theory, especially its statistical indicators, the railway accident as well as its key causations can be analyzed from the overall perspective. As a case, the “7.23” China—Yongwen railway accident is illustrated based on this model. The results show that the inspection of signals and the checking of line conditions before trains run played an important role in this railway accident. In conclusion, the constructed model gives a theoretical clue for railway accident prediction and, hence, greatly reduces the occurrence of railway accidents. (interdisciplinary physics and related areas of science and technology)

  8. Analyzing the causation of a railway accident based on a complex network

    Ma, Xin; Li, Ke-Ping; Luo, Zi-Yan; Zhou, Jin

    2014-02-01

    In this paper, a new model is constructed for the causation analysis of railway accident based on the complex network theory. In the model, the nodes are defined as various manifest or latent accident causal factors. By employing the complex network theory, especially its statistical indicators, the railway accident as well as its key causations can be analyzed from the overall perspective. As a case, the “7.23” China—Yongwen railway accident is illustrated based on this model. The results show that the inspection of signals and the checking of line conditions before trains run played an important role in this railway accident. In conclusion, the constructed model gives a theoretical clue for railway accident prediction and, hence, greatly reduces the occurrence of railway accidents.

  9. Mycobacterium tuberculosis Adaptation to Gradually Oxygen Depletion Condition Is a Key Factor for Latency of Tuberculosis

    M NajafiMosleh

    2007-06-01

    Full Text Available Background: There is ample evidence that the basis for latent tuberculosis infection in humans is persistence of tubercle bacilli for long periods of time even though lifetime. This status is currently defined as non-replicating persistence (NRP. Documented evidence from both macrophage physiology & the nature of TB granuloumas in human lungs suggests that gradually depletion of oxygen, hypoxia and micro aerobic condition is a major factor in inducing NRP state of tubercle bacilli. Methods: 100 clinically isolated tubercle bacilli were examined by the slowly stirred head space ratio method (0.5 HSR, which involve a slow depletion of oxygen within a sealed, slow stirred culture tube. The in-vitro induced hypoxically different stages of NRP was setting up, and the expression of the alpha-crystalline chaperone protein that are expressed when MTB undergoes to NRP srate was detected. Indeed the activity of rifampin, isoniazide, pyrazynamide, ciprofloxacin and meteronidazole were evaluated against two NRP stages of MTB. Results: During oxygen shift-down bacterial physiology changes from active growth to a NRP state. Two characteristic stages of NRP are seen; NRP-1 occurs when the oxygen concentration gradually dropped to microaerophilic condition. The 16 KD α- crystalline protein was expressed at just beginning of NRP-1 stage. The NRP-2 stages occur when the oxygen concentration dropped to anaerobic condition. When the NRP-2 state transferred to an oxygen – reach fresh medium the bacteria consume oxygen and resume growth in a synchronized replication manner from NRP-2 state. Conclusion: slow depletion of O2 appears to permit the occurrence of adaptations that favor long-term non replicating persistence of tubercle bacilli under microaerophilic conditions and also enhance the ability of the bacilli to survive in anaerobic conditions. This versatility could account for long-term latency of tuberculosis in the human host. The model presented here

  10. The effect of key process operational conditions on enhanced biological phosphorus removal from wastewater

    Carvalheira, Mónica Isabel Gonçalves

    2014-01-01

    Enhanced biological phosphorus removal (EBPR) is the most economic and sustainable option used in wastewater treatment plants (WWTPs) for phosphorus removal. In this process it is important to control the competition between polyphosphate accumulating organisms (PAOs) and glycogen accumulating organisms (GAOs), since EBPR deterioration or failure can be related with the proliferation of GAOs over PAOs. This thesis is focused on the effect of operational conditions (volatile fatty acid (VFA) c...

  11. Nitrogen assimilation and transpiration: key processes conditioning responsiveness of wheat to elevated [CO2] and temperature.

    Jauregui, Iván; Aroca, Ricardo; Garnica, María; Zamarreño, Ángel M; García-Mina, José M; Serret, Maria D; Parry, Martin; Irigoyen, Juan J; Aranjuelo, Iker

    2015-11-01

    Although climate scenarios have predicted an increase in [CO(2)] and temperature conditions, to date few experiments have focused on the interaction of [CO(2)] and temperature effects in wheat development. Recent evidence suggests that photosynthetic acclimation is linked to the photorespiration and N assimilation inhibition of plants exposed to elevated CO(2). The main goal of this study was to analyze the effect of interacting [CO(2)] and temperature on leaf photorespiration, C/N metabolism and N transport in wheat plants exposed to elevated [CO(2)] and temperature conditions. For this purpose, wheat plants were exposed to elevated [CO(2)] (400 vs 700 µmol mol(-1)) and temperature (ambient vs ambient + 4°C) in CO(2) gradient greenhouses during the entire life cycle. Although at the agronomic level, elevated temperature had no effect on plant biomass, physiological analyses revealed that combined elevated [CO(2)] and temperature negatively affected photosynthetic performance. The limited energy levels resulting from the reduced respiratory and photorespiration rates of such plants were apparently inadequate to sustain nitrate reductase activity. Inhibited N assimilation was associated with a strong reduction in amino acid content, conditioned leaf soluble protein content and constrained leaf N status. Therefore, the plant response to elevated [CO(2)] and elevated temperature resulted in photosynthetic acclimation. The reduction in transpiration rates induced limitations in nutrient transport in leaves of plants exposed to elevated [CO(2)] and temperature, led to mineral depletion and therefore contributed to the inhibition of photosynthetic activity. PMID:25958969

  12. 核舰船核事故应急救援行动要点研究%Study on Key Points of Nuclear Accident Emergency Rescue Operation for Nuclear Powered Vessel

    许明剑; 余刃; 寿宇强; 杨倩

    2015-01-01

    核动力舰船是移动的核设施,诱发核事故因素多,核应急救援组织指挥难度大。针对核动力舰船核事故特点,提出了迅速展开响应、严密组织指挥、注重技术救援、加强全局统筹等四个方面的行动要点,为组织开展核动力舰船核事故应急救援行动提供了参考。%Nuclear-powered vessels are mobile nuclear facilities. Multiple factors which may induce nuclear accident bring great difficulties to the organization and command of the nuclear emergency rescue. Based on the analysis of the characteristics of nuclear accidents on nuclear-powered vessels, the authors propose four key points of the rescue operation, i.e., rapid response, well-organized command, technical rescue and overall planning, which provide reference for the organization and implementation of the nuclear accident emergency rescue operation.

  13. Reducing conditions are the key for efficient production of active ribonuclease inhibitor in Escherichia coli

    Neubauer Peter

    2011-05-01

    Full Text Available Abstract Background The eukaryotic RNase ribonuclease/angiogenin inhibitors (RI are a protein group distinguished by a unique structure - they are composed of hydrophobic leucine-rich repeat motifs (LRR and contain a high amount of reduced cysteine residues. The members of this group are difficult to produce in E. coli and other recombinant hosts due to their high aggregation tendency. Results In this work dithiothreitol (DTT was successfully applied for improving the yield of correctly folded ribonuclease/angiogenin inhibitor in E. coli K12 periplasmic and cytoplasmic compartments. The feasibility of the in vivo folding concepts for cytoplasmic and periplasmic production were demonstrated at batch and fed-batch cultivation modes in shake flasks and at the bioreactor scale. Firstly, the best secretion conditions of RI in the periplasmic space were evaluated by using a high throughput multifactorial screening approach of a vector library, directly with the Enbase fed-batch production mode in 96-well plates. Secondly, the effect of the redox environment was evaluated in isogenic dsbA+ and dsbA- strains at the various cultivation conditions with reducing agents in the cultivation medium. Despite the fusion to the signal peptide, highest activities were found in the cytoplasmic fraction. Thus by removing the signal peptide the positive effect of the reducing agent DTT was clearly proven also for the cytoplasmic compartment. Finally, optimal periplasmic and cytoplasmic RI fed-batch production processes involving externally added DTT were developed in shake flasks and scaled up to the bioreactor scale. Conclusions DTT highly improved both, periplasmic and cytoplasmic accumulation and activity of RI at low synthesis rate, i.e. in constructs harbouring weak recombinant synthesis rate stipulating genetic elements together with cultivation at low temperature. In a stirred bioreactor environment RI folding was strongly improved by repeated pulse addition

  14. Radiological consequence analyses under severe accident conditions for the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory

    This paper discusses salient aspects of methodology, assumptions, modeling of various features related to radiation exposure, and health consequences from source terms resulting from two conservatively scoped severe accident scenarios. Radiological consequences for a site-suitability scenario based on 10 CFR 100 guidelines also are presented. Consequences arising from severe accidents involving steaming pools and core-concrete interaction (CCI) events combined with several different containment configurations are presented. Results are presented in the form of mean cumulative values for prompt and latent cancer fatality estimates and related cumulative, complementary distribution functions as a function of distance from the reactor site. It is shown that the reactor-site-suitability risk goals are met by a large margin and that overall risk is dominated by early containment failure combined with CCI events

  15. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    Heams, T J [Science Applications International Corp., Albuquerque, NM (United States); Williams, D A; Johns, N A; Mason, A [UKAEA, Winfrith, (England); Bixler, N E; Grimley, A J [Sandia National Labs., Albuquerque, NM (United States); Wheatley, C J [UKAEA, Culcheth (England); Dickson, L W [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Osborn-Lee, I [Oak Ridge National Lab., TN (United States); Domagala, P; Zawadzki, S; Rest, J [Argonne National Lab., IL (United States); Alexander, C A [Battelle, Columbus, OH (United States); Lee, R Y [Nuclear Regulatory Commission, Washington, DC (United States)

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided.

  16. Preclosure radiological safety analysis for accident conditions of the potential Yucca Mountain Repository: Underground facilities; Yucca Mountain Site Characterization Project

    Ma, C.W.; Sit, R.C.; Zavoshy, S.J.; Jardine, L.J. [Bechtel National, Inc., San Francisco, CA (United States); Laub, T.W. [Sandia National Labs., Albuquerque, NM (United States)

    1992-06-01

    This preliminary preclosure radiological safety analysis assesses the scenarios, probabilities, and potential radiological consequences associated with postulated accidents in the underground facility of the potential Yucca Mountain repository. The analysis follows a probabilistic-risk-assessment approach. Twenty-one event trees resulting in 129 accident scenarios are developed. Most of the scenarios have estimated annual probabilities ranging from 10{sup {minus}11}/yr to 10{sup {minus}5}/yr. The study identifies 33 scenarios that could result in offsite doses over 50 mrem and that have annual probabilities greater than 10{sup {minus}9}/yr. The largest offsite dose is calculated to be 220 mrem, which is less than the 500 mrem value used to define items important to safety in 10 CFR 60. The study does not address an estimate of uncertainties, therefore conclusions or decisions made as a result of this report should be made with caution.

  17. Accidents at work and living conditions among solid waste segregators in the open dump of Distrito Federal.

    Hoefel, Maria da Graça; Carneiro, Fernando Ferreira; Santos, Leonor Maria Pacheco; Gubert, Muriel Bauerman; Amate, Elisa Maria; dos Santos, Wallace

    2013-09-01

    The work of recycling solid waste segregators allows a precarious livelihood, but triggers a disease process that exacerbates their health and well-being. This study aimed to estimate the prevalence of occupational accidents at the open dump in the Federal District and its associated factors. Most segregators have had an accident at work (55.5%), perceived the danger of their working environment (95.0%) and claimed they did not receive personal protective equipment (51.7%). Among other findings, 55.8% ate foods found in the trash, 50.0% experienced food insecurity at home and 44.8% received Bolsa Família. There was a statistically significant relationship between work accidents and perception of dangerous work environment, household food insecurity and the presence of fatigue, stress or sadness (p waste segregators constitute a vulnerable community, not only from the perspective of labor, but also from the social and environmental circumstances. To reverse this situation, effective implementation of the National Policy of Solid Wastes is imperative, in association with affirmative policies to grant economic emancipation for this population. PMID:24896289

  18. The long-term cooling of the RU VVER in the conditions of design beyond basis accident

    MCC rupture on outlet, inlet the reactor (with coolant double end leakage); MCC rupture in the bottom part of a loop seal (with coolant double end leakage); Rupture of the connecting pipeline of HA (JNG10-40) of the ECCS (D = 300 mm); Loss of an alternating current sources, failure of safety active systems for more than 24 hours, failure of all diesel-generators; emergency supply from accumulators; Water injection from a fuel storage pool begins after emptying of tanks GE-2 (JNG50-80) with the total mass flow rate of 3.2 kg/s. The RELAP5/ANGAR code was used. The sequence of events and the work of the systems at a guillotine rupture of the MCC is described. The computing analysis of the beyond design accident which includes MCC rupture with loss of all sources of an alternating current, including diesel engines-generators, demonstrated the following: (i) the intervention of the safety systems of the NVNPP-2 project meets all Russian and international requirements to localizing functions by NVNPP-2 containment at beyond design accidents with leaks from the reactor facility; (ii) water injection from ST after the termination of work; (iii) GE-2 provides reliable core cooling for the beyond design basis accident scenarios during 74-88 hours in dependence on the leak size and location. (P.A.)

  19. Radio-contaminant behaviour in the cover-gas space and the containment building of a sodium-cooled fast reactor in accident conditions

    experiments chosen in the literature. The mass of oxide aerosols produced by a sodium spray fire can involve more than 60% of the ejected sodium. Then, we develop the numerical simulation STARK based on the Cooper model that model the physico-chemical transformations of the aerosols. However, this model has never been validated and the literature does not permit to do so. In these conditions, we designed and performed our own experiment to obtain the missing values of parameters that govern the Cooper model. The Cooper model has been improved with the results of this experimental study, ESSTIA, and we present a modified Cooper model that improves the accuracy of Cooper model to investigate the transformation of the sodium aerosols. The last part of the manuscript deals with the interaction between sodium aerosols (hydroxide) and a key fission product (iodine). We use density functional theory numerical simulation (the VASP code) to discover the affinities that can be identified. The results will facilitate simulation of the source term evolution because the sodium aerosols will interact with this FP. All the data and numerical simulations presented here will contribute to implementation of models in the future SFR SA numerical simulation of the IRSN, ASTEC-Na. (author)

  20. Field measurements of key parameters associated with nocturnal OBT formation in vegetables grown under Canadian conditions.

    Kim, S B; Workman, W G; Korolevych, V; Davis, P A

    2012-02-01

    The objective of this study was to provide the parameter values required to model OBT formation in the edible parts of plants following a hypothetical accidental tritium release to the atmosphere at night. The parameters considered were leaf area index, stomatal resistance, photosynthesis rate, the photosynthetic production rate of starch, the nocturnal hydrolysis rate of starch, the fraction of starch produced daily by photosynthesis that appears in the fruits, and the mass of the fruit. Values of these parameters were obtained in the summer of 2002 for lettuce, radishes and tomatoes grown under typical Canadian environmental conditions. Based on the maximum observed photosynthetic rate and growth rate, the fraction of starch translocated to the fruit was calculated to be 17% for tomato fruit and 14% for radish root. PMID:21962480

  1. Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

    A severe accident has inherently significant uncertainties due to the complex phenomena and wide range of conditions. Because of its high temperature and pressure, performing experimental validation and practical application are extremely difficult. With these difficulties, there has been few experimental researches performed and there is no plant-specific experimental data. Instead, computer codes have been developed to simulate the accident and have been used conservative assumptions and margins. This study is an effort to reduce the uncertainty in the probabilistic safety assessment and produce a realistic and physical-based failure probability. The methodology was developed and applied to the OPR1000. The creep rupture failure probabilities of reactor coolant system (RCS) components were evaluated under a station blackout severe accident with all powers lost and no recovery of steam generator auxiliary feed-water. The MELCOR 1.8.6 code was used to obtain the plant-specific pressure and temperature history of each part of the RCS and the creep rupture failure times were calculated by the rate-dependent creep rupture model with the plant-specific data. (author)

  2. Bacterial assisted degradation of chlorpyrifos: The key role of environmental conditions, trace metals and organic solvents.

    Khalid, Saira; Hashmi, Imran; Khan, Sher Jamal

    2016-03-01

    Wastewater from pesticide industries, agricultural or surface runoff containing pesticides and their residues has adverse environmental impacts. Present study demonstrates effect of petrochemicals and trace metals on chlorpyrifos (CP) biotransformation often released in wastewater of agrochemical industry. Biodegradation was investigated using bacterial strain Pseudomonas kilonensis SRK1 isolated from wastewater spiked with CP. Optimal environmental conditions for CP removal were CFU (306 × 10(6)), pH (8); initial CP concentration (150 mg/L) and glucose as additional carbon source. Among various organic solvents (petrochemicals) used in this study toluene has stimulatory effect on CP degradation process using SRK1, contrary to this benzene and phenol negatively inhibited degradation process. Application of metal ions (Cu (II), Fe (II) Zn (II) at low concentration (1 mg/L) took part in biochemical reaction and positively stimulated CP degradation process. Metal ions at high concentrations have inhibitory effect on degradation process. A first order growth model was shown to fit the data. It could be concluded that both type and concentration of metal ions and petrochemicals can affect CP degradation process. PMID:26692411

  3. Plant cell death and cellular alterations induced by ozone: Key studies in Mediterranean conditions

    An account of histo-cytological and ultrastructural studies on ozone effect on crop and forest species in Italy is given, with emphasis on induced cell death and the underlying mechanisms. Cell death phenomena possibly due to ambient O3 were recorded in crop and forest species. In contrast, visible O3 effects on Mediterranean vegetation are often unclear. Microscopy is thus suggested as an effective tool to validate and evaluate O3 injury to Mediterranean vegetation. A DAB-Evans blue staining was proposed to validate O3 symptoms at the microscopic level and for a pre-visual diagnosis of O3 injury. The method has been positively tested in some of the most important crop species, such as wheat, tomato, bean and onion and, with some restriction, in forest species, and it also allows one to gain some very useful insights into the mechanisms at the base of O3 sensitivity or tolerance. - Ozone-induced cell death is a frequent phenomenon in Mediterranean conditions, not only in the most sensitive crops but also in forest species.

  4. Finite element analysis for creep failure of coolant pipe in light water reactor due to local heating under severe accident condition

    During severe accident of a light water reactor (LWR), the piping of the reactor cooling system would be damaged when the piping is subjected to high internal pressure and very high temperature, resulted from high temperature gas generated in a reactor core and decay heat released from the deposit of fission products. It is considered that, under such a condition, short-term creep at high temperatures would cause the piping failure. For the evaluation of piping integrity under a severe accident, a method to predict such high temperature short-term creep deformation should be developed, using a creep constitutive equation considering tertiary creep. In this paper, the creep constitutive equation including tertiary creep was applied to nuclear-grade cold-drawn pipe of 316 stainless steel (SUS316), based on the isotropic damage mechanics proposed by Kachanov and Ravotnov. Tensile creep test data for the material of a SUS316 cold-drawn pipe were used to determine the coefficients of the creep constitutive equation. Using the constitutive equation taking account of creep damage, finite element analyses were performed for the local creep deformation of the coolant piping under two types of conditions; uniform temperature (isothermal condition) and temperature gradient of circumferential direction (non-isothermal condition). The analytical results show that the damage variable integrated into the creep constitutive equation can predict the pipe failure in the test performed by Japan Atomic Energy Research Institute, in which failure occurred from the outside of the pipe wall

  5. An analysis of impact on the environmental pollution under accident conditions due to the disposal of ashes from the Nikola Tesla B thermal power plant

    The ash disposal area of Nikola Tesla B TPP is presenting considerable potential source of environmental pollution. Due to its vast area, that is for an effective plot of about 400 ha, under unfavorable meteorological conditions the emerging of ash cloud spreading over the surrounding area at distances of over 15 km could be produced. This paper deals with accident conditions at ash disposal area which, after stopping of sprinkling operation or in case of an inadequate sequence of ash tapping, when larger portions of disposal area are dried, turn into area of air polluting sources. Computations of particle dispersions under such conditions have been performed for different meteorological conditions and the results are represented as space distribution of particle deposition on the soil and their concentration in the air. (author). 5 figs., 4 refs

  6. Key issues on safety design basis selection and safety assessment

    In current fast reactor design in Japan, four design accident conditions and four design seismic conditions are adopted as the design base classifications. These are classified by the considerations on both likelihood of occurrence and the severeness of the consequences. There are several major problem areas in safety design consideration such as core accident problems which include fuel sodium interaction, fuel failure propagation and residual decay heat removal, and decay heat removal systems problems which is more or less the problem of selection of appropriate system and of assurance of high reliability of the system. In view of licensing, two kinds of accidents are postulated in evaluating the adequacy of a reactor site. The one is the ''major accident'' which is the accident to give most severe radiation hazard to the public from technical point of view. The other is the ''hypothetical accident'', induced public accident of which is severer than that of major accident. While the concept of the former is rather unique to Japanese licensing, the latter is almost equivalent to design base hypothetical accident of the US practice. In this paper, design bases selections, key safety issues and some of the licensing considerations in Japan are described

  7. Influence of aging on the retention of elemental radioiodine by deep bed carbon filters under accident conditions

    No significant difference was found in the retention of I-131 loaded as I2, by various impregnated activated carbons that had been aged in the containment exhaust air of a pressurized water reactor over a period of 12 months. In all the cases, the I-131 passing through deep beds of carbon was in a nonelemental form. It was concluded that a minimum retention of 99.99%, as required by new guidelines for certain accident filters, can be equally well achieved with various carbons in deep beds

  8. Whole core analysis of an open pool research reactor under the most severe loss of coolant accident conditions

    In the present work the accident in which either the outlet or the inlet coolant pipe connected to the bottom of the reactor tank in an open pool research reactor is completely ruptured has been analyzed. The 3-D transient computer code ThEAP-I developed at Democritus NRC has been utilized and applied to the 5 MW Greek Research Reactor (GRR-1). The results show that a partial melting of the reactor core is possible for the GRR-1, the amount of melting being roughly and conservatively estimated to be of the order of 20%. (author)

  9. Report 1. An experiment model. Radiation loading in animals living in conditions of external and internal irradiation within the zone of Chernobyl accident

    Irradiation conditions in which laboratory animals were kept in experimental laboratories of Chernobyl and Kiev after the accident APS are described. The data are presented on the spectral structure and activity of radionuclides in the diet as well as in the organs and tissues of the animals. The radition loads have been estimated with regard to an external gamma-component and the internal one contributed by the incorporated radionuclides. It has been shown that radiation doses received by the animals during their lifetime due to these contributions do not exceed units of cGy

  10. The influence of the environmental and psychological factors of the Chernobyl' accident on the functional condition of regulatory hormone system in man

    The functional condition of the studied hormone systems (hormones in the periphery blood) in young men taken part in the accident liquidation and suffered from neuro-circulatory dystonia has been shown to have a number of peculiarities. In the bar period after leaving the 30-km zone the excess of adrenal cortex activity on the background of the impairment of the mechanism of the negative feedback in the hypothis-adrenal system has been registered. This fact is reflected in the state of the hormonal functions regulating the vessel tonus and hydrocarbon exchange resulting in the dystonia reaction of the hypertonia type. 12 refs.; 1 tab

  11. Important severe accident research issues after Fukushima accident

    After the Fukushima accident several investigation committees issued reports with lessons learned from the accident in Japan. Among those lessons, several recommendations have been made on severe accident research. Similar to the EURSAFE efforts under EU Program, review of specific severe accident research items was started before Fukushima accident in working group of Atomic Energy Society of Japan (AESJ) in terms of significance of consequences, uncertainties of phenomena and maturity of assessment methodology. Re-investigation has been started since the Fukushima accident. Additional effects of Fukushima accident, such as core degradation behaviors, sea water injection, containment failure/leakage and re-criticality have been covered. The review results are categorized in ten major fields; core degradation behavior, core melt coolability/retention in containment vessel, function of containment vessel, source term, hydrogen behavior, fuel-coolant interaction, molten core concrete interaction, direct containment heating, recriticality and instrumentation in severe accident conditions. Based on these activities and also author's personal view, the present paper describes the perspective of important severe accident research issues after Fukushima accident. Those are specifically investigation of damaged core and components, advanced severe accident analysis capabilities and associated experimental investigations, development of reliable passive cooling system for core/containment, analysis of hydrogen behavior and investigation of hydrogen measures, enhancement of removal function of radioactive materials of containment venting, advanced instrumentation for the diagnosis of severe accident and assessment of advanced containment design which excludes long-term evacuation in any severe accident situations. (author)

  12. Analysis of fuel rod behaviour within a rod bundle of a pressurized water reactor under the conditions of a loss of coolant accident (LOCA) using probabilistic methodology

    The assessment of fuel rod behaviour under PWR LOCA conditions aims at the evaluation of the peak cladding temperatures and the (final) maximum circumferential cladding strains. Moreover, the estimation of the amount of possible coolant channel blockages within a rod bundle is of special interest, as large coplanar clad strains of adjacent rods may result in strong local reductions of coolant channel areas. Coolant channel blockages of large radial extent may impair the long-term coolability of the corresponding rods. A model has been developed to describe these accident consequences using probabilistic methodology. This model is applied to study the behaviour of fuel rods under accident conditions following the double-ended pipe rupture between collant pump and pressure vessel in the primary system of a 1300 MW(el)-PWR. Specifically a rod bundle is considered consisting of 236 fuel rods, that is subjected to severe thermal and mechanical loading. The results obtained indicate that plastic clad deformations with circumferential clad strains of more than 30% cannot be excluded for hot rods of the reference bundle. However, coplanar coolant channel blockages of significant extent seem to be probable within that bundle only under certain boundary conditions which are assumed to be pessimistic. (orig./RW)

  13. Modeling of hot tensile and short-term creep strength for LWR piping materials under severe accident conditions

    The analytical study on severe accident shows the possibility of the reactor coolant system (RCS) piping failure before reactor pressure vessel failure under the high primary pressure sequence at pressurized water reactors. The establishment of the high-temperature strength model of the realistic RCS piping materials is important in order to predict precisely the accident progression and to evaluate the piping behavior with small uncertainties. Based on material testing, the 0.2% proof stress and the ultimate tensile strength above 800degC were given by the equations of second degree as a function of the reciprocal absolute temperature considering the strength increase due to fine precipitates for the piping materials. The piping materials include type 316 stainless steel, type 316 stainless steel of nuclear grade, CF8M cast duplex stainless steel and STS410 carbon steel. Also the short-term creep rupture time and the minimum creep rate at high-temperature were given by the modified Norton's Law as a function of stress and temperature considering the effect of the precipitation formation and resolution on the creep strength. The present modified Norton's Law gives better results than the conventional Larson-Miller method. Correlating the creep data (the applied stress versus the minimum creep rate) with the tensile data (the 0.2% proof stress or the ultimate tensile strength versus the strain rate), it was found that the dynamic recrystallization significantly occurred at high-temperature. (author)

  14. Thermal-hydraulic behavior of a PWR under accident conditions complementary test results from UPTF and PKL

    Two complementary test facilities - the Upper Plenum Test Facility (UPTF) and the Primaerkreislauf test facility (PKL) - were constructed to investigate the thermal-hydraulic response of a pressurized water reactor (PWR) during postulated accidents. The UPTF is a geometrical full-scale simulation of the primary system of a 1300-MW PWR. The upper plenum, the downcomer and the four connected loops as well as the pressurizer are represented on a 1:1 scale. The integral test facility PKL also simulates a 1300-MW PWR, whereby the power and volume is reduced by a factor of 1:145 (elevations 1:1). The PKL test facility models the entire primary system, relevant parts of the secondary side and all important engineered safety and auxiliary systems. Whereas the UPTF was mainly designed to perform separate-effect tests focusing on multidimensional thermal-hydraulic phenomena in full-scale simulated components, the main objective of the PKL tests has been the investigation of the thermal-hydraulic system behavior on the primary and secondary side. So far the program objectives represent a reasonable completion and in summary the experimental results from both test facilities provide an essential contribution for a better understanding of assumed accident sequences in a PWR. Test results which demonstrate the complementary character of the UPTF and the PKL test programs as well as the interaction between the two test facilities are presented in this paper. (author)

  15. Considerations of severe accidents in the design of Korean Next Generation Reactor

    The severe accident is one of the key issues in the design of Korean Next Generation Reactor (KNGR) which is an evolutionary type of pressurized water reactor. As IAEA recommends in TECDOC-801, the design objective of KNGR with regard to safety is provide a sound technical basis by which an imminent off-site emergency response to any circumstance could be practically unnecessary. To implement this design objective, probabilistic safety goals were established and design requirements were developed for systems to mitigate severe accidents. The basic approach of KNGR to address severe accidents is firstly prevent severe accidents by reinforcing its capability to cope with the design basis accidents (DBA) and further with some accidents beyond DBAs caused by multiple failures, and secondly mitigate severe accidents to ensure the retention of radioactive materials in the containment by providing mean to maintain the containment integrity. For severe accident mitigation, KNGR principally takes the concept of ex-vessel corium cooling. To implement this concept, KNGR is equipped with a large cavity and cavity flooding system connected to the in-containment refueling water storage tank. Other major systems incorporated in KNGR are hydrogen igniters and safety depressurization systems. In addition, the KNGR containment is designed to withstand the pressure and temperature conditions expected during the course of severe accidents. In this paper, the design features and status of system designs related with severe accidents will be presented. Also, R and D activities related to severe accident mitigation system design will be briefly described

  16. Management of severe accidents

    The definition and the multidimensionality aspects of accident management have been reviewed. The suggested elements in the development of a programme for severe accident management have been identified and discussed. The strategies concentrate on the two tiered approaches. Operative management utilizes the plant's equipment and operators capabilities. The recovery managment concevtrates on preserving the containment, or delaying its failure, inhibiting the release, and on strategies once there has been a release. The inspiration for this paper was an excellent overview report on perspectives on managing severe accidents in commercial nuclear power plants and extending plant operating procedures into the severe accident regime; and by the most recent publication of the International Nuclear Safety Advisory Group (INSAG) considering the question of risk reduction and source term reduction through accident prevention, management and mitigation. The latter document concludes that 'active development of accident management measures by plant personnel can lead to very large reductions in source terms and risk', and goes further in considering and formulating the key issue: 'The most fruitful path to follow in reducing risk even further is through the planning of accident management.' (author)

  17. Contain calculations of debris conditions adjacent to the BWR Mark I drywell shell during the later phases of a severe accident

    Best estimate CONTAIN calculations have recently been performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to predict the primary containment response during the later phases of an unmitigated low-pressure Short Term Station Blackout at the Peach Bottom Atomic Power Station. Debris pour conditions leaving the failed reactor vessel are taken from the results of best estimate BWRSAR analyses that are based upon an assumed metallic debris melting temperature of 2750/degree/F (1783 K) and an oxide debris melting temperature of 4350/degree/F (2672 K). Results of the CONTAIN analysis for the case without sprays indicate failure of the drywell seals due to the extremely hot atmospheric conditions extant in the drywell. The maximum calculated temperature of the debris adjacent to the drywell shell is less than the melting temperature of the shell, yet the sustained temperatures may be sufficient to induce primary containment pressure boundary failure by the mechanism of creep-rupture. It is also predicted that a significant portion of the reactor pedestal wall is ablated during the period of the calculation. Nevertheless, the calculated results are recognized to be influenced by large modeling uncertainties. Several deficiencies in the application of the CORCON module within the CONTAIN code to BWR severe accident sequences are identified and discussed. 5 refs., 9 figs., 4 tabs.,

  18. Analytical functions used for description of the plastic deformation process in Zirconium alloys WWER type fuel rod cladding under designed accident conditions

    The aim of this work was to improve the RAPTA-5 code as applied to the analysis of the thermomechanical behavior of the fuel rod cladding under designed accident conditions. The irreversible process thermodynamics methods were proposed to be used for the description of the plastic deformation process in zirconium alloys under accident conditions. Functions, which describe yielding stress dependence on plastic strain, strain rate and temperature may be successfully used in calculations. On the basis of the experiments made and the existent experimental data the dependence of yielding stress on plastic strain, strain rate, temperature and heating rate for E110 alloy was determined. In future the following research work shall be made: research of dynamic strain ageing in E635 alloy under different strain rates; research of strain rate influence on plastic strain in E635 alloy under test temperature higher than 873 K; research of deformation strengthening of E635 alloy under high temperatures; research of heating rate influence n phase transformation in E110 and E635 alloys

  19. Modelling the chemical behaviour of tellurium species in the reactor pressure vessel and the reactor cooling system under severe accident conditions

    This state of the art report contains information on the behaviour of tellurium and its compounds in the reactor pressure vessel and the reactor coolant system under light water reactor severe accident conditions. To characterise tellurium behaviour, it is necessary the previous knowledge of the species of tellurium released from the core, and simultaneity of its release with that of other materials which can alter the transport, for instance, control rod and structural materials. Release and transport experiments have been reviewed along with the models implemented in the codes which are used in the international community: TRAPMELT, RAFT, VICTORIA and SOPHIE. From the experiments, it can be concluded that other species different to Te2, such as tin telluride and cesium telluride, may be released from the fuel. That is why they must be considered in the transport phenomena. There is also experimental evidence of the strong interaction of Te2 with Inconel 600 and stainless steel of the pipe walls and structures, however this strong interaction is in competition with the interaction of tellurium with aerosols, which under severe accident conditions may represent an area greater than that of the primary system. It is for the absence of significant tellurium species in the transport models, and also for the interaction of tellurium with aerosols, for which some codes show the greatest deficiencies

  20. On fission product retention in the core of the low powered high temperature reactor under accident conditions

    In the core of the high temperature reactor the fuel element and the coated particles contained herein provide the safest enclosure for fission products. The complex process of fission product transport out of the particle kernel, through the particle coating and within the fuel element graphite is described in a simplified form by the Fick's diffusion. The effective diffusion coefficient is used for calculation. Starting from the existing ideas of fission product transport five burn-up and temperature-dependent diffusion coefficients for Cesium in (Th,U)O2-kernels are derived in this study. The results have been gained from several fuel element radiation experiments in recent years, which showed extreme variation in regard to burn-up, temperature cycle, neutron flux and operation time. Cs-137 release measurements from single particle kernels were present from all the experiments. Furthermore, annealing tests of AVR-fuel elements were analyzed. Heat-temperatur and heating-time, the fuel element burn-up in the AVR-reactor, as well as the measured Cs-137 inventory of the fuel elements before and after annealing, are included in the investigation as essential parameters. With the aid of the derived diffusion coeffizients and already present data sets the Cs-137 release of fuel elements into a small reactor core is investigated under unrestricted core heat-up. While the released Cs-137 is derived mainly from defective particles at accident temperatures up to 16000C, the main part diffuses through the particle coating at higher accident temperatures. (orig./HP)

  1. 49 CFR 195.54 - Accident reports.

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Accident reports. 195.54 Section 195.54... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.54 Accident reports. (a) Each operator that experiences an accident that is required to be reported under § 195.50 shall as soon...

  2. Nuclear accidents

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

  3. Radiation accidents

    Radiation accidents may be viewed as unusual exposure event which provide possible high exposure to a few people and, in the case of nuclear plants events, low exposure to large population. A number of radiation accidents have occurred over the past 50 years, involving radiation machines, radioactive materials and uncontrolled nuclear reactors. These accidents have resulted in number of people have been exposed to a range of internal and external radiation doses and those involving radioactive materials have involved multiple routs of exposure. Some of the more important accidents involving significant radiation doses or releases of radioactive materials, including any known health effects involves in it. An analysis of the common characteristics of accidents is useful resolving overarching issues, as has been done following nuclear power, industrial radiography and medical accidents. Success in avoiding accidents and responding when they do occur requires planning in order to have adequately trained and prepared health physics organization; well defined and developed instrument program; close cooperation among radiation protection experts, local and state authorities. Focus is given to the successful avoidance of accidents and response in the events they do occur. Palomares, spain in late 1960, Goiania, Brazil in 1987, Thule, Greenland in 1968, Rocky flats, Colorado in 1957 and 1969, Three mile island, Pennsylvania in 1979, Chernobyl Ukraine in april 1986, Kyshtym, former Soviet Union in 1957, Windscale, UK in Oct. 1957 Tomsk, Russian Federation in 1993, and many others are the important examples of major radiation accidents. (author)

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

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

    2011-04-15

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

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

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

  6. Environmental Conditions Influence Induction of Key ABC-Transporter Genes Affecting Glyphosate Resistance Mechanism in Conyza canadensis

    Tani, Eleni; Chachalis, Demosthenis; Travlos, Ilias S.; Bilalis, Dimitrios

    2016-01-01

    Conyza canadensis has been reported to be the most frequent weed species that evolved resistance to glyphosate in various parts of the world. The objective of the present study was to investigate the effect of environmental conditions (temperature and light) on the expression levels of the EPSPS gene and two major ABC-transporter genes (M10 and M11) on glyphosate susceptible (GS) and glyphosate resistant (GR) horseweed populations, collected from several regions across Greece. Real-time PCR was conducted to determine the expression level of the aforementioned genes when glyphosate was applied at normal (1×; 533 g·a.e.·ha−1) and high rates (4×, 8×), measured at an early one day after treatment (DAT) and a later stage (four DAT) of expression. Plants were exposed to light or dark conditions, at three temperature regimes (8, 25, 35 °C). GR plants were made sensitive when exposed to 8 °C with light; those sensitized plants behaved biochemically (shikimate accumulation) and molecularly (expression of EPSPS and ABC-genes) like the GS plants. Results from the current study show the direct link between the environmental conditions and the induction level of the above key genes that likely affect the efficiency of the proposed mechanism of glyphosate resistance. PMID:27104532

  7. Environmental Conditions Influence Induction of Key ABC-Transporter Genes Affecting Glyphosate Resistance Mechanism in Conyza canadensis.

    Tani, Eleni; Chachalis, Demosthenis; Travlos, Ilias S; Bilalis, Dimitrios

    2016-01-01

    Conyza canadensis has been reported to be the most frequent weed species that evolved resistance to glyphosate in various parts of the world. The objective of the present study was to investigate the effect of environmental conditions (temperature and light) on the expression levels of the EPSPS gene and two major ABC-transporter genes (M10 and M11) on glyphosate susceptible (GS) and glyphosate resistant (GR) horseweed populations, collected from several regions across Greece. Real-time PCR was conducted to determine the expression level of the aforementioned genes when glyphosate was applied at normal (1×; 533 g·a.e.·ha(-1)) and high rates (4×, 8×), measured at an early one day after treatment (DAT) and a later stage (four DAT) of expression. Plants were exposed to light or dark conditions, at three temperature regimes (8, 25, 35 °C). GR plants were made sensitive when exposed to 8 °C with light; those sensitized plants behaved biochemically (shikimate accumulation) and molecularly (expression of EPSPS and ABC-genes) like the GS plants. Results from the current study show the direct link between the environmental conditions and the induction level of the above key genes that likely affect the efficiency of the proposed mechanism of glyphosate resistance. PMID:27104532

  8. Stakeholder involvement in the rehabilitation of living conditions in contaminated territories affected by the Chernobyl accident. The ETHOS Project in Belarus

    The management of the Chernobyl post-accident situation is a complex process comprising not only radiological protection but also psychological, social, economic, political and ethical dimensions. Involving in this process local communities who are directly concerned by the consequences of the accident is a strong lever in improving their living conditions as well as restoring their confidence in experts and the authorities. This paper presents the experience of the involvement of a group of mothers from a village in the Republic of Belarus in activities to improve the radiological protection of their children. This experience took place in the framework of the ETHOS Project supported by the radiation protection research programme of the European Commission with the objective of implementing an alternative approach to the rehabilitation strategies adopted so far in the contaminated territories of the CIS. The first part of the paper presents briefly the main features of the methodological and practical approach of the ETHOS Project. The second part describes in more detail how the mothers voluntarily got involved in a working group set up within the framework of the Project, the characterization of the radiological situation they carried out, the concrete approach they developed to regain control of the situation, the way the health care system has been involved in the process and finally, the results they achieved in reducing the internal contamination of their children. (author)

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

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

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

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

    2005-01-01

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

  11. An international standard problem: analysis of 1:4-scale prestressed concrete containment vessel model under severe accident conditions

    In June 2002, The OECD-NEA Committee on the Safety of Nuclear Installations (CSNI), with the encouragement of the US NRC, initiated an International Standard Problem on containment integrity (ISP 48) based on the NRC/NUPEC/Sandia test. The objectives of the ISP are to extend the understanding of capacities of actual containment structures based on results of the recent PCCV Model test and other previous research. From 1997 through 2001 Sandia National Laboratories (SNL) conducted a Cooperative Containment Integrity Program under the joint sponsorship of the Nuclear Power Engineering Corporation (NUPEC) of Japan, and the NRC Office of Nuclear Regulatory Research. The purpose of the program was to investigate the response of representative models of nuclear containment structures to pressure loading beyond the design basis accident and to compare analytical predictions to measured behavior. A uniform 1:4-scale model of a prestressed concrete containment vessel (PCCV) was constructed and tested at SNL. This model was representative of the containment structure of an actual pressurized-water reactor plant in Japan. The ISP consists of four phases over a period of 2 years: Phase 1: Data Collection and Identification Phase 2: Calculation of the Limit State Test (LST), i.e. static pressure loading Phase 3: Calculation of response to both Thermal and Mechanical Loadings Phase 4: Reporting Workshop Eleven organizations (or teams) from nine OECD member countries accepted the invitation to participate in the ISP and perform calculations to predict the structural response of the PCCV model to static and transient pressure and thermal loading. Each participating organization was provided with the model and loading data and was asked to perform independent analyses to simulate the response of the PCCV model. The results of each team's calculations were compiled and the results presented at a final workshop in April 2005. These results and the conclusions and insights gained from

  12. Analysis of a simulated small break in the semiscale system under loss-of-coolant accident conditions

    The Semiscale Mod-1 experimental program conducted by EG and G Idaho, Inc., is part of the overall U.S. Nuclear Regulatory Commission (NRC) and Department of Energy (DOE) sponsored research and development program to investigate the behavior of the pressurized water reactor (PWR) system during an hypothesized loss-of-coolant accident (LOCA). The Semiscale Mod-1 program is intended to provide transient thermal-hydraulic data from a simulated LOCA using a small-scale experimental nonnuclear system. The Semiscale Mod-1 program is a major contributor of experimental data that provide a means of evaluating the adequacy of overall system analytical models as well as the models of the individual system components. Selected experimental data produced by this program will also be used to aid other DOE and NRC sponsored experimental programs, such as the Loss-of-Fluid Test (LOFT) program in optimizing test series, selecting test parameters, and evaluating test results. The Semiscale Mod-1 tests are performed with an experimental system which simulates the principal features of a nuclear plant but which is smaller in volume. Nuclear heating is simulated in the tests by a core composed of an array of electrically heated rods. The core is contained in a pressure vessel which also includes a downcomer, lower plenum, and upper plenum. The Semiscale system piping is arranged such that the intact loop represents three loops of a four-loop nuclear plant, and the broken loop represents the fourth loop. In the present configuration the intact loop contains an active steam generator and pump, and the broken loop contains passive simulators for the steam generator and pump

  13. Licensing topical report: the measurement and modelling of time-dependent fission product release from failed HTGR fuel particles under accident conditions

    The release of fission products from failed fuel particles was measured under simulated accident (core heatup) conditions. A generic model and specific model parameters that describe delayed fission product release from the kernels of failed HTGR fuel particles were developed from the experimental results. The release of fission products was measured from laser-failed BISO ThO2 and highly enriched (HEU) TRISO UC2 particles that had been irradiated to a range of kernel burnups. The burnups were 0.25, 1.4, and 15.7% FIMA for ThO2 particles and 23.5 and 74% FIMA for UC2 particles. The fission products measured were nuclides of xenon, iodine, krypton, tellurium, and cesium

  14. Operational conditions of storages for products of decontamination of the territories of Belarus after the accident at Chernobyl NPP and evaluation of their radioecological state

    As a result of carrying out the measures on decontamination of the territory of Belarus after the Chernobyl accident, the storages for radioactive products of decontamination have been arranged in the Republic. Up to now, 69 storage sites for radioactive products of decontamination have been examined and registered. Six of them, the most typical, with the highest activity, are under constant control with the help of the network of the hydrogeological observation holes. The analysis of the field conditions of the storage sites at the territory of Belarus has shown that there is the violation of requirements for safe storage practically for all storages. The evaluation of protection of the ground water against radioactive contamination has shown, that in 10--100 years, the contamination of the ground water with caesium-137 is possible in concentrations lower than the Republican permissible levels and with strontium in concentrations significantly exceeding the specified values

  15. A kinetic model for fission-product release and fuel oxidation behaviour for zircaloy-clad fuel elements under reactor accident conditions

    During a severe reactor accident fission products will be released from the degraded fuel in the reactor core. In addition, hydrogen will be generated at high-temperature by the steam oxidation of the core materials. This oxidation process will also influence the rate of fission-product release. Separate-effects tests performed out-of-pile at the Chalk River Laboratories (CRL) have provided a better understanding of the processes of fission product release during severe accident conditions. The annealing experiments were conducted in steam at temperatures ranging from 1200 to 1700 deg C with irradiated fuel specimens of uranium dioxide in the form of bare fuel fragments and a short-length Zircaloy-clad fuel element. The fission product release was monitored by online gamma ray spectrometry. The oxygen partial pressure was also measured with solid-state oxygen sensors, providing a calculation of the rate of oxygen consumption and hydrogen production in the fuel specimens. Based on the CRL tests, an analytical model has been developed to describe the kinetic release behaviour of the volatile fission-product species (cesium) during high-temperature accident conditions. The physically-based model accounts for the kinetics of fuel oxidation as a rate-determining reaction at the fuel/steam interface. A more general framework is therefore provided to detail the influence of the atmosphere (i.e. oxygen potential) on the behaviour of the fission product release. Solid state diffusion in the fuel matrix is shown to be the rate-controlling mechanism in the early stages of release. The enhanced diffusivity of fission products in the hyperstoichiometric fuel is modelled with the assumption that diffusion takes place on vacant cation lattice sites. When the fuel reaches a state of oxidation of x ∼ 0.07 for the UO2+x phase, a more rapid release process occurs in accordance with first-order rate kinetics. The retarding influence of the hydrogen production on the fuel oxidation

  16. Numerical simulation of the insulation material transport to a PWR core under loss of coolant accident conditions

    Höhne, Thomas, E-mail: T.Hoehne@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Fluid Dynamics, P.O. Box 510119, D-01328 Dresden (Germany); Grahn, Alexander; Kliem, Sören; Rohde, Ulrich; Weiss, Frank-Peter [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Fluid Dynamics, P.O. Box 510119, D-01328 Dresden (Germany)

    2013-05-15

    Highlights: ► Detailed results of a numerical simulation of the insulation material transport to a PWR core are shown. ► The spacer grid is modeled as a strainer which completely retains the insulation material carried by coolant. ► The CFD calculations showed that the fibers at the upper spacer grid plane are not uniformly distributed. ► Furthermore the pressure loss does not exceed a critical limit. ► The PWR core coolablity can be guaranteed all the time during the transient. -- Abstract: In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modeled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the

  17. Numerical simulation of the insulation material transport to a PWR core under loss of coolant accident conditions

    Highlights: ► Detailed results of a numerical simulation of the insulation material transport to a PWR core are shown. ► The spacer grid is modeled as a strainer which completely retains the insulation material carried by coolant. ► The CFD calculations showed that the fibers at the upper spacer grid plane are not uniformly distributed. ► Furthermore the pressure loss does not exceed a critical limit. ► The PWR core coolablity can be guaranteed all the time during the transient. -- Abstract: In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modeled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the

  18. The IDA cognitive model for the analysis of nuclear power plant operator response under accident conditions. Part I: problem solving and decision making model

    This paper is the first of a series of papers describing IDA which is a cognitive model for analysing the behaviour of nuclear power plant operators under accident conditions. The domain of applicability of the model is a relatively constrained environment where behaviour is significantly influenced by high levels of training and explicit requirement to follow written procedures. IDA consists of a model for individual operator behaviour and a model for control room operating crew expanded from the individual model. The model and its derivatives such as an error taxonomy and data collection approach has been designed with ultimate objective of becoming a quantitative method for human reliability analysis (HRA) in probabilistic risk assessment (PRA). The present paper gives a description of the main components of IDA such as memory structure, goals, and problem solving and decision making strategies. It also identifies factors that are at the origin of transitions between goals or between strategies. These factors cover the effects of external conditions and psychological state of the operator. The description is generic at first and then made specific to the nuclear power plant environment and more precisely to abnormal conditions

  19. Hydrogen radiolytic production in light and heavy water mixtures under conditions similar to LOCA (loss of coolant accidents)

    H2, HD and D2 radiolytic yield in heavy and light water mixtures has been determined to supply the necessary data which will allow to make a realistic estimation of the solution of such gas under LOCA conditions as a function of time. (Author)

  20. Accident Statistics

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

  1. Accident management information needs

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

  2. Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation.

    Tentner, A. M.; Parma, E.; Wei, T.; Wigeland, R.; Nuclear Engineering Division; SNL; INL

    2010-03-01

    An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

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

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

  4. SUBSTANTIATION OF THE CONCEPT OF TRANSFER TO CONDITIONS OF NORMAL POPULATION ACTIVITY OF THE SETTLEMENTS CONSIDERED TO BE ZONES OF RADIOACTIVE CONTAMINATION AFTER THE CHERNOBYL NPP ACCIDENT

    I. K. Romanovich

    2016-01-01

    Full Text Available The article contains substantiation of criteria of return of territories with radioactive pollution caused by Chernobyl NPP accident to conditions of normal population activity. It is established that in 12 entities of the Russian Federation (except Bryansk and Kaluga regions all agricultural food produce, including that from the personal part-time farms, corresponds to hygienic specifications. Non- corresponding to the standard SanPiN 2.3.2.1078-01 on 137Cs are part of the milk samples produced at personal part-time farms of the Bryansk region and most of natural foodstuff samples (berries, mushrooms, fish and wild animals meat in Bryansk and Kaluga regions. The content of 137Cs both in agricultural and in wild-growing foodstuff produced at radioactively contaminated territories depends not only on the density of radioactive pollution, but also on the types of soil. The average settlement annual effective dose of population irradiation (AAED90 in the 3700 among 4413 settlements as of 2014 was below 0.3 mSv/year. Only in 713 settlements of Bryansk, Kaluga, Oryol and Tula regions the AAED90 exceeds 0.3 mSv/year. In the Bryansk region, once subject to the greatest radioactive contamination, in 276 settlements AAED90 exceeds 1 mSv/year, and in 8 of them - 5 mSv/year.The legislation of the Russian Federation defines only criteria and requirements for consideration of the suffered territories as zones of radioactive contamination. Requirements on transfer of territories polluted by radiation accidents and their population to normal life activity conditions (regarding the radiological factor are not developed.Radiological criteria are suggested for transfer of the settlements considered to be the zone of radioactive pollution to conditions of normal life activity: average irradiation dose of critical population group: 1.0 mSv per year and lower (AAED crit; decrease of radionuclide soil contamination density to the level enabling to use the territory

  5. Iodine releases from reactor accidents

    The airborne releases of iodine from water reactor accidents are small fractions of the available iodine and occur only slowly. However, in reactor accidents in which water is absent, the release of iodine to the environment can be large and rapid. These differences in release fraction and rate are related to the chemical states attained by iodine under the accident conditions. It is clear that neither rapid issue of blocking KI nor rapid evacuation of the surrounding population is required to protect the public from the radioiodine released in the event of a major water reactor accident

  6. [Psychogenesis of accidents].

    Giannattasio, E; Nencini, R; Nicolosi, N

    1988-01-01

    After having carried out a historical review of industrial psychology with specific attention to the evolution of the concept of causality in accidents, the Authors formulate their work hypothesis from that research which take into highest consideration the executives' attitudes in the genesis of the accidents. As dogmatism appears to be one of the most negative of executives' attitudes, the Authors administered Rockeach's Scale to 130 intermediate executives from 6 industries in Latium and observed the frequency index for accidents and the morbidity index (absenteeism) of the 2149 workhand. The Authors assumed that to high degree of dogmatism on the executives' side should correspond o a higher level of accidents and absenteeism among the staff. The data processing revealed that, due to the type of machinery employed, three of the industries examined should be considered as High Risk Industrie (HRI), while the remaining three could be considered as Low Risk Industries (LRI): in fact, due to the different working conditions, a significant lower number of accidents occurred in last the three. A statistically significant correlation between the executives' dogmatism and the number of accidents among their workhand in the HRI has been noticed, while this has not been observed in the LRI. This confirms, as had already been pointed out by Gemelli in 1944, that some "objective conditions" are requested so that the accident may actually take place. On the other hand the morbidity index has not shown any difference related to the different kind of industries (HRI, LRI): in both cases statistically significant correlations were obtained between the executives' dogmatism and the staff's absenteeism. absenteeism.(ABSTRACT TRUNCATED AT 250 WORDS) PMID:3154344

  7. Modelling of behaviour of 37 fuel rod assembly with Zr1%Nb-alloy simulators cladding under loss-of-coolant accident conditions on PARAMETR-M facility

    The experiment described in this report involves the implementation of conditions complying with the second stage of LOCA accident, for representative group of WWER-1000 fuel rods with relative heat generation rate in the range of 1.2-1.4 from the average one: maximal cladding temperature up to 9000C. The testing of experimental fuel rod assembly consisting of 37 fuel elements with Zr1%Nb-alloyed claddings has been made for representative group of heat-stressed fuel rods of the WWER-1000 type reactor on the electro heated PARAMETR-M facility under LOCA simulating conditions. The cladding rupture of fuel rods took place at the heating-up stage within the stated temperature interval 800-9000C. There were identified the basic cladding deformation and rupture parameters: temperature, pressure, axial distribution of hoop strain, and azimuthal distributions of radial deformation in rupture section. The experimental and calculated value of cross section blockage in the assembly under testing was 38%. The calculated values of cladding deformation and rupture parameters determined using RAPTA-5 Code agree well with experimental ones

  8. German (GRS) approach to accident analysis (part I). German licensing basis for accident analyses. Applicants accident analyses in second part license for Konvoi-plants. Appendix 1. Assessor accident analyses in second part license for Konvoi-plants. Appendix 2. Reference list of DBA to be considered in the safety status analysis of a PSR. Appendix 3a. Reference list of special very rare and BDB plant conditions to be considered in the safety status analysis of a PSE. Appendix 3b

    LOCA analyses.Appendix 3a: Concerns Level 3, accidents and events to be considered for transients, accidents and events to be considered for losses of coolant (LOCA), accidents and radiologically representative events, as well as PWR-specific spreading impacts to be considered as possible initiating events for transients and SBLOCA. These events are defined for PWR and BWR. Appendix 3b: Level 4, PWR- and BWR-specific special, very rare events,(ATWS and site-specific external civil impacts (certain emergencies)), beyond-design-basis plant conditions. These event and accident as in previous Appendix is defined again for PWR and BWR reactor types

  9. Computer system for the assessment of radiation situation in the cases of radiological accidents and extreme weather conditions in the Chernobyl exclusion zone

    Talerko, M.; Garger, E.; Kuzmenko, A. [Institute for Safety Problems of Nuclear Power Plants (Ukraine)

    2014-07-01

    Radiation situation within the Chernobyl Exclusion Zone (ChEZ) is determined by high radionuclides contamination of the land surface formed after the 1986 accident, as well as the presence of a number of potentially hazardous objects (the 'Shelter' object, the Interim Spent Nuclear Fuel Dry Storage Facility ISF-1, radioactive waste disposal sites, radioactive waste temporary localization sites etc.). The air concentration of radionuclides over the ChEZ territory and radiation exposure of personnel are influenced by natural and anthropogenic factors: variable weather conditions, forest fires, construction and excavation activity etc. The comprehensive radiation monitoring and early warning system in the ChEZ was established under financial support of European Commission in 2011. It involves the computer system developed for assessment and prediction of radiological emergencies consequences in the ChEZ ensuring the protection of personnel and the population living near its borders. The system assesses radiation situation under both normal conditions in the ChEZ and radiological emergencies which result in considerable radionuclides emission into the air (accidents at radiation hazardous objects, extreme weather conditions). Three different types of radionuclides release sources can be considered in the software package. So it is based on a set of different models of emission, atmospheric transport and deposition of radionuclides: 1) mesoscale model of radionuclide atmospheric transport LEDI for calculations of the radionuclides emission from stacks and buildings; 2) model of atmospheric transport and deposition of radionuclides due to anthropogenic resuspension from contaminated area (area surface source model) as a result of construction and excavation activity, heavy traffic etc.; 3) model of resuspension, atmospheric transport and deposition of radionuclides during grassland and forest fires in the ChEZ. The system calculates the volume and surface

  10. Computer system for the assessment of radiation situation in the cases of radiological accidents and extreme weather conditions in the Chernobyl exclusion zone

    Radiation situation within the Chernobyl Exclusion Zone (ChEZ) is determined by high radionuclides contamination of the land surface formed after the 1986 accident, as well as the presence of a number of potentially hazardous objects (the 'Shelter' object, the Interim Spent Nuclear Fuel Dry Storage Facility ISF-1, radioactive waste disposal sites, radioactive waste temporary localization sites etc.). The air concentration of radionuclides over the ChEZ territory and radiation exposure of personnel are influenced by natural and anthropogenic factors: variable weather conditions, forest fires, construction and excavation activity etc. The comprehensive radiation monitoring and early warning system in the ChEZ was established under financial support of European Commission in 2011. It involves the computer system developed for assessment and prediction of radiological emergencies consequences in the ChEZ ensuring the protection of personnel and the population living near its borders. The system assesses radiation situation under both normal conditions in the ChEZ and radiological emergencies which result in considerable radionuclides emission into the air (accidents at radiation hazardous objects, extreme weather conditions). Three different types of radionuclides release sources can be considered in the software package. So it is based on a set of different models of emission, atmospheric transport and deposition of radionuclides: 1) mesoscale model of radionuclide atmospheric transport LEDI for calculations of the radionuclides emission from stacks and buildings; 2) model of atmospheric transport and deposition of radionuclides due to anthropogenic resuspension from contaminated area (area surface source model) as a result of construction and excavation activity, heavy traffic etc.; 3) model of resuspension, atmospheric transport and deposition of radionuclides during grassland and forest fires in the ChEZ. The system calculates the volume and surface specific

  11. Irradiated fuel behavior under accident heating conditions and correlation with fission gas release and swelling model (Chicago)

    We analyse the mixed oxide fast fuel response to off normal conditions obtained by means of an out-of-pile transient simulation apparatus designed to provide direct observations (temperature, pressure, fuel motion) of fuel fission gas phenomena that might occur during the transients. The results are concerning fast transient tests (0,1 to 1 second) obtained with high gas concentration irradiated fuel (4 to 7 at % burn up, 0,4 cm3Xe + Kr /g.UPuO2). The kinetics of fission gas release during the transients have been directly measured and then compared with the calculated results issued of the Chicago model. This model agrees, quite well, with other experiments done in the silene prompt reactor. Other gases than xenon and krypton (such as hydrogen and carbon monoxide) do not play any role in fuel behavior, since they have been completely ruled out

  12. Experimental investigation of the focusing effect of the metallic layer heat transfer in a severe accident condition

    Focusing effect of the metallic layer was investigated experimentally for Rayleigh numbers ranging 8.49×107∼5.49×109 and aspect ratios 0.135∼0.540 respectively. The height of the side wall was varied. High Rayleigh numbers were achieved using mass transfer experiments based on the heat and mass transfer analogy. Piecewise electrodes are adopted to measure the local average mass transfer. An electrical resistance was attached to the top wall so as to mimic top hotter wall condition. The measured results and existing heat transfer correlations were in good agreements. As the height reduces, the focusing effect becomes severe, especially at the corner near the bottom. (author)

  13. A preliminary study for the implementation of general accident management strategies

    Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of each strategy are also investigated. 11 refs., 3 figs., 3 tabs. (Author)

  14. Stress in accident and post-accident management at Chernobyl

    The effects of the Chernobyl nuclear accident on the psychology of the affected population have been much discussed. The psychological dimension has been advanced as a factor explaining the emergence, from 1990 onwards, of a post-accident crisis in the main CIS countries affected. This article presents the conclusions of a series of European studies, which focused on the consequences of the Chernobyl accident. These studies show that the psychological and social effects associated with the post-accident situation arise from the interdependency of a number of complex factors exerting a deleterious effect on the population. We shall first attempt to characterise the stress phenomena observed among the population affected by the accident. Secondly, we will be presenting an anlysis of the various factors that have contributed to the emerging psychological and social features of population reaction to the accident and in post-accident phases, while not neglecting the effects of the pre-accident situation on the target population. Thirdly, we shall devote some initial consideration to the conditions that might be conducive to better management of post-accident stress. In conclusion, we shall emphasise the need to restore confidence among the population generally. (Author)

  15. The Role of Trace Evidence in the Identification of Air-Conditioning Leakage Accident%微量物证在空调漏电事故鉴定中的作用

    沈胜凡; 张旭; 孙芳荣; 陈鹏浩; 高伟

    2014-01-01

    通过对事故现场勘验、检测,找出了空调漏电的原因。表明空调故障检修时必须按章操作,同时强调空调安装系统接地是必须的。%We invest the accident scene,detect the air conditioning,and find the reason for the air conditioning leakage accidents.Through the process of the investing,we know that we must go by the operating system when the air conditioning maintenance.While it is necessary for the air conditioning system gets ground.

  16. Models and criteria for prediction of Deflagration-to-Detonation Transition (DDT) in hydrogen-air-steam systems under severe accident conditions. Final report

    The European Commission in Brussels supported a joint project on Deflagration-to-Detonation Transition (DDT) studies for hydrogen safety within the framework programme on nuclear fission safety. The project was initiated by the Forschungszentrum Juelich based on the results of a pilot project. The following main project was coordinated by the Freie Universitaet Berlin involving seven european partners. The partners came from universities, research centers and industry, as follows: FU-Berlin, RWTH-Aachen, CNRS-Marseille, IPSN-Saclay, FZ-Juelich, FZ-Karlsruhe, and NNC-Knutsford, which worked closely together. The working period was two years (1997-1998). The aim of the project was to develop models and criteria for prediction of deflagration-to-detonation transition (DDT) in hydrogen-air-steam systems under severe accident conditions. The results obtained are documented in this final report, which was finished in 1999. The report consists of seven chapters, concerning: - Introduction - Experimental Investigations - Modelling and Numerics - Validation - Mitigation - Further Deliverables - Summary and Conclusion. The final report presents special experimental, theoretical, and computational aspects of the complex DDT phenomena for hydrogen safety studies, and it should be a solid basis for end user applications and further developments. (orig.)

  17. CAMDYN: a new model to describe the axial motion of molten fuel inside the pin of a fast breeder reactor during accident conditions

    The new in-pin fuel motion model CAMDYN (Cavity Material Dynamics) describes the axial motion of both partially and fully molten fuel inside the pin of a fast breeder reactor during accident conditions. The motion of the two types of molten fuel and the imbedded fission gas bubbles is treated both before and after cladding failure. The basic modelling approach consists of the treatment of two one-dimensional flows which are coupled by interaction terms. Each of these flows is treated compressively and with axially variable flow cross sections. The mass and energy equations of both fields are solved explicitly using upwind differencing on a fixed Eulerian grid. The two momentum equations are solved simultaneously, using the convective momentum fluxes of the previous timestep. Both partially and fully molten fuel can move axially into a central hole extending to the plenum in the case of certain hollow pellet designs. The fuel temperature calculation includes the determination of a radial temperature profile. A simple conduction freezing model is included. After cladding failure, ejection into the coolant channel is modeled

  18. Lizimetric investigation of vertical transportation of Cs-137 in the different soil types for the different raining conditions: a new solution proposal and application for the Chernobyl accident

    The aim of this Ph.D. thesis is to trace the vertical transportation of water in soil by using the Cs-137 as radiotracer. We have studied 3 different soil types clay, loamy and sandy and 3 different raining conditions. High, normal and low raining regimes that are typical especially for the Black Sea Region were selected for experimental parameters. These experiments have been realized in lizimetric conditions. As a result, the type of applications for the combinations of soil type and raining regime contain experimental originality. Humidity observation technique was also used. The result of the experiments show that in every raining regime the activities of different types of soils can be classified as sand, loam and clay in descending activity order. When the results of the experiments were evaluated according to the raining regime it can be seen that the relative activity for every type of soil is always towards higher to lower raining conditions. These results are what we already have expected and are related to the water permeability of the soil. The surface soil activities have also been measured. The surface soil activities were arranged as clayey, loamy and sandy in descending activity order for every laining regime. The surface soil activities were found to be higher in the soil types in which water could not penetrate deeper. When the relative activity values were compared with the results of the surface soil activities it was recognized that the order of soils changes upside down. This is an evident of harmony for the experimental results. For the evaluation of surface activity it can be seen that the relative activity values for every kind of soil type is always in the same order from the highest to the lowest raining regimes. As a result, it can be said that the all results have significant and have harmonies with each other. Experimental results were comparatively examined by using the semi infinite convection dispersion model. All the conclusions

  19. Severe Accident Management Measures Introduced in Belgian NPP's

    In response to the Belgian Safety Authorities' request to address the severe accident issue within a decennial safety review, Tractebel, on behalf of the Belgian Utility, Electrabel, examined in detail specific severe accident topics and provided the Utility with several measures that could be implemented to reduce the risk associated with beyond-design accidents. The present paper summarizes the key elements of the approach applied in Belgium: - Presentation of plant-specific studies related to severe accident issues; - Use of PSA results; - Inputs of international R and D projects; - Selection and justification of severe accident measures; - Comparative study between possible mitigative measures; - Definition and justification of implemented severe accident management strategies. The vulnerability to severe accidents as well as the potential causes of containment failures have been identified leading to the study of possible countermeasures taking into account the combination of conservative design and post-TMI measures already implemented . A section of the paper will also be devoted to the specific study made for the selection, the sizing and the implementation of hydrogen control means. After the description of the selected measures implemented, the paper also describes the content of the 'Severe Accident Management Guidelines' developed by Tractebel for the Tihange NPPs and for the Doel NPPs. This project aimed at providing the operators with procedures or guidelines enabling to deal with complex situations not formally considered in the standard Emergency Response Guidelines, including accidents in which a significant portion of the core melts. The objective of these SAMG's programs is to indicate actions that must bring the plant to a controlled stable state and, above all, mitigate any challenges to the fission product barriers. The plant personnel must use the available plant information to determine the best severe accident management measures. Obviously

  20. Key; key ring

    Key and/or key ring is provided with illuminating means which do not require a source of electrical power. The key shown comprises a shank and a head, the head being moulded from a luminous plastic material which glows in the dark. To improve keying together of shank and head there is a pair of holes in the head of the shank. Alternatively the key may be coated in illuminating material or provided with sheath or cover of illuminating material. The key may be provided with company logo. The key ring has attached thereto a hollow plastics container whose interior is coated with a phosphor and which is filled with tritium gas. (author)

  1. Modeling alternative clad behavior for accident tolerant systems

    The US Department of Energy Fuel Cycle Research and Development program has a key goal of helping develop accident tolerant fuels (ATF) through investigating fuel and clad forms. In the current work thermochemical modeling and experiment are being used to assess fuel and clad alternatives. Cladding alternatives that have promise to improve fuel performance under accident conditions include the FeCrAl family of alloys and SiC-based composites. These are high strength and radiation resistant alloys and ceramics that have increased resistance to oxidation as compared to zirconium alloys. Accident modeling codes have indicated substantially increased time to failure and resulting effects. In the current work the thermochemical behavior of these materials are being assessed and the work reported here. (author)

  2. Chernobyl accident. Exposures and effects

    The Chernobyl accident that occurred in Ukraine in April 1986 happened during an experimental test of the electrical control system as the reactor was being shut down for routine maintenance. The operators, in violation of safety regulations, had switched off important control systems and allowed the reactor to reach unstable, low-power conditions. A sudden power surge caused a steam explosion that ruptured the reactor vessel and allowed further violent fuel-steam interactions that destroyed the reactor and the reactor building. The Chernobyl accident was the most serious to have ever occurred in the nuclear power industry. The accident caused the early death of 30 power plant employees and fire fighters and resulted in widespread radioactive contamination in areas of Belarus, the Russian Federation, and Ukraine inhabited by several million people. Radionuclides released from the reactor that caused exposure of individuals were mainly iodine-131, caesium-134 and caesium-137. Iodine-131 has a short radioactive half-life (8 days), but it can be transferred relatively rapidly through milk and leafy vegetables to humans. Iodine becomes localized in the thyroid gland. For reasons of intake of these foods, size of thyroid gland and metabolism, the thyroid doses are usually greater to infants and children than to adults. The isotopes of caesium have relatively long half-lives (caesium-134: 2 years; caesium-137: 30 years). These radionuclides cause long-term exposures through the ingestion pathway and from external exposure to these radionuclides deposited on the ground. In addition to radiation exposure, the accident caused long-term changes in the lives of people living in the contaminated regions, since measures intended to limit radiation doses included resettlements, changes in food supplies, and restrictions in activities of individuals and families. These changes were accompanied by major economic, social and political changes in the affected countries resulting

  3. Proceedings of the CSNI workshop on International Standard Problem 48 - Analysis of 1:4-scale prestressed concrete containment vessel model under severe accident conditions

    At the CSNI meeting in June 2002, the proposal for an International Standard Problem on containment integrity (ISP 48) based on the NRC/NUPEC/Sandia test was approved. Objectives were to extend the understanding of capacities of actual containment structures based on results of the recent PCCV Model test and other previous research. The ISP was sponsored by the USNRC, and results had been made available thanks to NUPEC and to the USNRC. Sandia National Laboratory was contracted to manage the technical aspects of the ISP. At the end of the ISP48, a workshop was organized in Lyon, France on April 6-7, 2005 hosted by Electricite de France. Its overall objective was to present results obtained by participants in the ISP 48 and to assess the current practices and the state of the art with respect to the calculation of concrete structures under severe accident conditions. Experience from other areas in civil engineering related to the modelling of complex structures was greatly beneficial to all. Information obtained as a result of this assessment were utilized to develop a consensus on these calculations and identify issues or 'gaps' in the present knowledge for the primary purpose of formulating and prioritizing research needs on this topic. The ISP48 exercise was published in the report referenced NEA/CSNI/R(2005)5 in 3 volumes. Volume 1 contains the synthesis of the exercise; Volumes 2 and 3 contain individual contributions of participating organizations. The CSNI Working Group on the Integrity and Ageing and in particular its sub-group on the behaviour of concrete structures has produced extensive material over the last few years. The complete list of references is given in this document. These proceedings gather the papers and presentations given by the participants at the Lyon workshop

  4. Tchernobyl accident

    First, R.M.B.K type reactors are described. Then, safety problems are dealt with reactor control, behavior during transients, normal loss of power and behavior of the reactor in case of leak. A possible scenario of the accident of Tchernobyl is proposed: events before the explosion, possible initiators, possible scenario and events subsequent to the core meltdown (corium-concrete interaction, interaction with the groundwater table). An estimation of the source term is proposed first from the installation characteristics and the supposed scenario of the accident, and from the measurements in Europe; radiological consequences are also estimated. Radioactivity measurements (Europe, Scandinavia, Western Europe, France) are given in tables (meteorological maps and fallouts in Europe). Finally, a description of the site is given

  5. Accident: Reminder

    2003-01-01

    There is no left turn to Point 1 from the customs, direction CERN. A terrible accident happened last week on the Route de Meyrin just outside Entrance B because traffic regulations were not respected. You are reminded that when travelling from the customs, direction CERN, turning left to Point 1 is forbidden. Access to Point 1 from the customs is only via entering CERN, going down to the roundabout and coming back up to the traffic lights at Entrance B

  6. Study of the impact on PSA success criteria of the variability of the initial liquid level in case of the loss of the RHR system accident scenario under mid-loop operating conditions

    Probabilistic safety assessment (PSA) is recognized nowadays as an important tool to support risk-informed decision-making aimed at providing both operational flexibility and plant safety [1]. Experience of current PSA studies shows the importance of some risky scenarios with the plant at low power and shutdown conditions as compared to the accident scenarios with the plant operating at full power. In particular, current low power and shutdown PSA (LPSA) studies shows that the loss of the Residual Heat Removal System (RHRS) transient is one of the most risk-significant events under low power conditions [2]. This accident type is supposed to occur for various plant operating states, of which mid-loop operation represents one of the main contributors [3]. LPSA has widely used methods for thermal-hydraulic analysis that play an important role in determining success criteria of safety-related functions involved to mitigate the severity of accident scenarios with the plant operating in such conditions. Various best estimate thermal-hydraulic analysis codes have been used to analyze the loss of the RHRS during low power and shutdown conditions [4, 5]. It is known that RELAP code can give good results as derived after a number of benchmark exercises using results from experiments at research facilities (e.g. ROSA-IV, BETHSY, PKL). [6] Previous research has shown how thermal-hydraulic phenomena after the loss of the RHRS, e.g. peak reactor coolant system pressure, are sensitive to the initial liquid level at the time of loss of the RHRS [2]. This paper presents the results of the study of the thermalhydraulic analysis of the accident scenarios after the loss of the RHRS under mid-loop conditions paying particular attention to the analysis of the effect of the variability of the initial liquid level on the success criteria of the safety-related functions considered in a typical LPSA [3]. (author)

  7. Expert software for accident identification

    Each type of an accident in a Nuclear Power Plant (NPP) causes immediately after the start of the accident variations of physical parameters that are typical for that type of the accident thus enabling its identification. Examples of these parameter are: decrease of reactor coolant system pressure, increase of radiation level in the containment, increase of pressure in the containment. An expert software enabling a fast preliminary identification of the type of the accident in Krsko NPP has been developed. As input data selected typical parameters from Emergency Response Data System (ERDS) of the Krsko NPP are used. Based on these parameters the expert software identifies the type of the accident and also provides the user with appropriate references (past analyses and other documentation of such an accident). The expert software is to be used as a support tool by an expert team that forms in case of an emergency at Slovenian Nuclear Safety Administration (SNSA) with the task to determine the cause of the accident, its most probable scenario and the source term. The expert software should provide initial identification of the event, while the final one is still to be made after appropriate assessment of the event by the expert group considering possibility of non-typical events, multiple causes, initial conditions, influences of operators' actions etc. The expert software can be also used as an educational/training tool and even as a simple database of available accident analyses. (author)

  8. Methodological basis for qualification of the main steam line isolation valve in the beyond design basis accident conditions of expiration of steam-water mixture

    The work presents the computational and empirical methods and performance criteria for qualification of the main steam-isolation valve (MSIV) in accident scenarios with the expiration of steam-water mixture from the steam generator. Expiration of steam mixture through MSIV is possible in the beyond design basis accidents with inter loop leaks with full failure of functions of pressure control and/or isolation for emergency steam generator feed water. As a result of the preliminary analysis by the proposed method, the qualification criteria were determined for the modes of steam-water mixture expiration through MSIV at stable processes and surge

  9. Guidelines for calculating radiation doses to the public from a release of airborne radioactive material under hypothetical accident conditions in nuclear reactors

    This Standard provides guidelines and a methodology for calculating effective doses and thyroid doses to people (either individually or collectively) in the path of airborne radioactive material released from a nuclear facility following a hypothetical accident. The specific radionuclides considered in the Standard are those associated with substances having the greatest potential for becoming airborne in reactor accidents (eg, tritium (HTO), noble gases and their daughters (Kr-Rb, Xe-Cs), and radioiodines (I)); and certain radioactive particulates (eg, Cs, Ru, Sr, Te) that may become airborne under exceptional circumstances

  10. Transuranics and fission products release from PWR fuels in severe accident conditions. Lessons learnt from VERCORS RT3 and RT4 tests

    Over the last decades, several experimental programs devoted to the source term of fission products (FP) and actinides released from PWR fuel samples in severe accident (SA) conditions have been initiated throughout the world. In France, in this context, the Institute for Radiological Protection and Safety (IRSN) and Electricite de France (EDF) have supported the analytical VERCORS program which was performed by the Commissariat a l'Energie Atomique (CEA). The VERCORS facility at the LAMA-laboratory (CEA-Grenoble, France) was designed to heat up an irradiated fuel sample - taken from EDF's nuclear power reactors - to fuel relocation, and to capture the fission products released from the fuel and deposited downstream on a series of specific filters (impactors, bead-bed filter). On-line gamma detectors aimed at the fuel position, filters and gas capacity monitored the progress of FP release from the fuel, FP deposition on the filters and the fission gases emitted by the fuel (xenon and krypton). Before and after the test, a longitudinal gamma-scan of the fuel was conducted to measure the initial and final FP inventory in order to evaluate the quantitative fractions of FP emitted by the fuel during the test. All the components of the loop were then gamma-scanned to measure and locate the FPs released during the test and to draw up a mass balance of these FP. 25 annealing tests were performed between 1983 and 2002 on irradiated PWR fuels under various conditions of temperature and atmospheres (oxidising or reducing conditions). The influence of the nature of the fuel (UO2 versus MOX, burn up) and the fuel morphology (initially intact or fragmented fuel) have also been investigated. This led to an extended data base allowing on the one hand to study mechanisms which promote FP release in SA conditions, and on the other hand to enhance models implemented in SA codes. Because gamma spectrometry is well suited to FP measurement and not to actinides (except neptunium

  11. Transportation accidents

    Predicting the possible consequences of transportation accidents provides a severe challenge to an analyst who must make a judgment of the likely consequences of a release event at an unpredictable time and place. Since it is impractical to try to obtain detailed knowledge of the meteorology and terrain for every potential accident location on a route or to obtain accurate descriptions of population distributions or sensitive property to be protected (data which are more likely to be more readily available when one deals with fixed-site problems), he is constrained to make conservative assumptions in response to a demanding public audience. These conservative assumptions are frequently offset by very small source terms (relative to a fixed site) created when a transport vehicle is involved in an accident. For radioactive materials, which are the principal interest of the authors, only the most elementary models have been used for assessing the consequences of release of these materials in the transportation setting. Risk analysis and environmental impact statements frequently have used the Pasquill-Gifford/gaussian techniques for releases of short duration, which are both simple and easy to apply and require a minimum amount of detailed information. However, after deciding to use such a model, the problem of selecting what specific parameters to use in specific transportation situations still presents itself. Additional complications arise because source terms are not well characterized, release rates can be variable over short and long time periods, and mechanisms by which source aerosols become entrained in air are not always obvious. Some approaches that have been used to address these problems will be reviewed with emphasis on guidelines to avoid the Worst-Case Scenario Syndrome

  12. Investigation on accident management measures for VVER-1000 reactors

    A consequence of a total loss of AC power supply (station blackout) leading to unavailability of major active safety systems which could not perform their safety functions is that the safety criteria ensuring a secure operation of the nuclear power plant would be violated and a consequent core heat-up with possible core degradation would occur. Currently, a study which examines the thermal-hydraulic behaviour of the plant during the early phase of the scenario is being performed. This paper focuses on the possibilities for delay or mitigation of the accident sequence to progress into a severe one by applying Accident Management Measures (AMM). The strategy 'Primary circuit depressurization' as a basic strategy, which is realized in the management of severe accidents is being investigated. By reducing the load over the vessel under severe accident conditions, prerequisites for maintaining the integrity of the primary circuit are being created. The time-margins for operators' intervention as key issues are being also assessed. The task is accomplished by applying the GRS thermal-hydraulic system code ATHLET. In addition, a comparative analysis of the accident progression for a station blackout event for both a reference German PWR and a reference VVER-1000, taking into account the plant specifics, is being performed. (authors)

  13. Technical Requirements and Principles for the Standards Development of the Key Parts for Rotor Air-conditioning Compressors

    Sun Min; Wen Yun; Fan Zhangzeng

    2011-01-01

    ntroductionSince 2000,air-conditioning sales continues to grow,and the development of air-conditioning market makes a booming market of compressor.At the present time,compressor production rising all the way,and the sales steps up the new steps constantly.Tendency chart is shown in figure 1.Rotor compressor with its simple structure,small volume,light weight,easy processed mechanical parts,reliable operation and other excellent characteristics occupied the dominant position in the market.Compared with reciprocating compressor on the same application situation,decreased in the size by 40%~50%,weight was reduced by 40%~50%.But there were also disadvantages,mainly large friction loss,friction power consumption was about 10%of compressor's total power input.

  14. The role of radiation protection in the rehabilitation of living conditions in contaminated territories after a nuclear accident or a radiological event

    The implementation of radiation protection programmes in long -term contaminated territories is raising difficult technical, methodological, but also ethical questions, which commit the responsibility of the radiation protection professionals and the way they are exercising their expertise in such situation. The presence of contamination in the environment is a new reality which is profoundly affecting the living conditions of the population in all aspects: health, environment but also the psychological, social and economical cultural dimensions of daily life. In such a context, the classical radiation protection approach is inoperative to restore the normality of the ante accident situation On the contrary, the scientific and technological expertise and the prescriptive administrative arrangements, including the radiation protection ones, which are heavily mobilized to respond to the many challenges facing the society induce a progressive de -qualification of the environment and goods and a loss of control of the situation by the population which turn into a social depression. The paper illustrates how radiation protection programmes for managing long term contaminated territories have an ambivalent impact: on the one hand they ensure the protection of the population against the potential health impacts of radiations, on the other hand they re -enforce the loss of control, de-qualification and exclusion processes which results in a weakening of the resilience capacity of the population and the territories. The recent experience of co-expertise and co-operation implemented in the contaminated territories of Belarus have demonstrated that the direct engagement of local professionals and the population in the daily management of the radiological situation is not only feasible but also necessary to stop this vicious circle of loss of control and exclusion. To be effective and sustainable this engagement must rely on the diffusion among the population of a 'practical

  15. Reactor accident in Chernobyl

    The bibliography contains 1568 descriptions of papers devoted to Chernobylsk accident and recorded in ''INIS Atomindex'' to 30 June 1990. The descriptions were taken from ''INIS Atomindex'' and are presented in accordance with volumes of this journal (chronology of recording). Therefore all descriptions have numbers showing first the number of volume and then the number of record. The bibliography has at the end the detailed subject index consisting of 465 main headings and a lot of qualifiers. Some of them are descriptors taken from ''INIS Atomindex'' and some are key words taken from natural language. The index is in English as descriptions in the bibliography. (author)

  16. Business incubation in a university as a key condition for the formation of innovational micro entrepreneurship in a region

    Anatoliy Viktorovich Grebenkin

    2012-09-01

    Full Text Available This paper substantiates the hypothesis of the special role of universities in creating an environment of innovational micro entrepreneurship in a region. The role of business incubators is allocated; the algorithm for selecting projects is described. The results of a three-year organizational and economic experiment (with the changing conditions on the functioning of the student business incubator in the Ural State University are shown. Various models of the selection of ideas and projects for different cycles of incubation areimplemented. A decision on theestablishment of the Entrepreneurship Center in theInstitute of Management and Entrepreneurship is made. The Center’s main task is to form a series of events to support continuous generation of students’ business ideas, finding resonant response with the University experts and representatives of business environment in the region. A student in the business incubation system plays a new role for a Russian university — a role of a catalyst, i.e., directly acts as an element of positive feedback in the innovational system. It is shown that the catalytic path of the establishment and development of small high-tech business — Science to Business (StB — leads to the phenomenon of resonance, i.e., sustainable innovation flow generated by the business incubator of the University. The poll of the USU students in 2009-2011 (a sample from 660 to 854 respondents confirmed their positive attitude towards entrepreneurship and allowed to estimate the structure of the factors that hamper to increase student participation in the innovational business. Three blocks of factors were identified: the reluctance to take risks, inaccessibility of material and financial resources and the turbulence of the environment. A system of monitoring students' attitudes towards entrepreneurship, which allows adjusting the curriculum and creating institutional conditions for activation of innovative entrepreneurship

  17. Summary of major accidents with radiation sources and lessons learned

    The paper reviews some of the major radiological accidents that have occurred around the world and identifies key lessons to be learned from them. It emphasizes the value of feedback from the reporting of accidents, the need for effective reporting mechanisms and, most important, the importance of acting on the lessons learned to ensure accident prevention. (author)

  18. Guidelines for calculating radiation doses to the public from a release of airborne radioactive material under hypothetical accident conditions in nuclear reactors

    This standard provides guidelines and a methodology for calculating effective doses and thyroid doses to people (either individually or collectively) in the path of airborne radioactive material released from a nuclear facility following a hypothetical accident. The radionuclides considered are those associated with substances having the greatest potential for becoming airborne in reactor accidents: tritium (HTO), noble gases and their daughters, radioiodines, and certain radioactive particulates (Cs, Ru, Sr, Te). The standard focuses on the calculation of radiation doses for external exposures from radioactive material in the cloud; internal exposures for inhalation of radioactive material in the cloud and skin penetration of tritium; and external exposures from radionuclides deposited on the ground. It uses as modified Gaussian plume model to evaluate the time-integrated concentration downwind. (52 refs., 12 tabs., 21 figs.)

  19. Kiche; A Simulation tool for kinetics of iodine chemistry in the containment of light water reactors under severe accident conditions (Contract research)

    森山 清史; 丸山 結; 中村 秀夫

    2011-01-01

    An iodine chemistry simulation tool, Kiche, was developed for analyses of chemical kinetics relevant to iodine volatilization in the containment vessel of light water reactors (LWRs) during a severe accident. It consists of a Fortran code to solve chemical kinetics models, reaction databases written in plain text format, and peripheral tools to convert the reaction databases into Fortran codes. Potential advantages of Kiche are the text format reaction database separated from the code that pr...

  20. Systematics of Reconstructed Process Facility Criticality Accidents

    Pruvost, N.L.; McLaughlin, T.P.; Monahan, S.P.

    1999-09-19

    The systematics of the characteristics of twenty-one criticality accidents occurring in nuclear processing facilities of the Russian Federation, the United States, and the United Kingdom are examined. By systematics the authors mean the degree of consistency or agreement between the factual parameters reported for the accidents and the experimentally known conditions for criticality. The twenty-one reported process criticality accidents are not sufficiently well described to justify attempting detailed neutronic modeling. However, results of classic hand calculations confirm the credibility of the reported accident conditions.

  1. JAERI's activities in JCO accident

    The Japan Atomic Energy Research Institute (JAERI) was actively involved in a variety of technical supports and cooperative activities, such as advice on terminating the criticality condition, contamination checks of the residents and consultation services for the residents, as emergency response actions to the criticality accident at the uranium processing facility operated by the JCO Co. Ltd., which occurred on September 30, 1999. These activities were carried out in collaborative ways by the JAERI staff from the Tokai Research Establishment, Naka Fusion Research Establishment, Oarai Research Establishment, and Headquarter Office in Tokyo. As well, the JAERI was engaged in the post-accident activities such as identification of accident causes, analyses of the criticality accident, and dose assessment of exposed residents, to support the Headquarter for Accident Countermeasures of the Science and Technology Agency (STA), the Accident Investigation Committee and the Health Control Committee of the Nuclear Safety Commission of Japan (NSC). This report compiles the activities, that the JAERI has conducted to date, including the discussions on measures for terminating the criticality condition, evaluation of the fission number, radiation monitoring in the environment, dose assessment, analyses of criticality dynamics. (author)

  2. Drilling the Mediterranean Messinian Evaporites to Answer Key Questions Related to Massive Microbial Dolomite Formation under Hypersaline Alkaline Conditions

    McKenzie, Judith A.; Bontognali, Tomaso R. R.; Vasconcelos, Crisogono

    2014-05-01

    Deep-sea drilling in the Mediterranean during DSSP Leg 13 in 1970 revealed the basin-wide occurrence of a Messinian evaporite formation. This spectacular discovery was pursued further during a subsequent drilling program, DSDP Leg 42A, in 1975, which was designed, in part, to obtain continuous cores to study the evolution of the salinity crisis itself (Hsü, Montadert, et al., 1978). Specifically, drilling at a water depth of 4,088 m in the Ionian Sea, DSDP Site 374: Messina Abyssal Plain, penetrated about 80 m into the uppermost part of the Messinian upper evaporite formation. The sedimentary sequence comprises dolomitic mudstone overlying dolomitic mudstone/gypsum cycles, which in turn overlie anhydrite and halite. The non-fossiliferous dolomitic mudstone is generally rich in organic carbon, with TOC values ranging from 0.9% to 5.3%, of possible marine origin with a good source rock potential. Commonly laminated dolomitic mudstones contain preserved filamentous cyanobacterial remains suggesting that conditions were conducive for microbial mat growth. The Ca-dolomite, composed of fine-grained anhedral crystals in the size range of 2-4 μm, is probably a primary precipitate. The unusual interstitial brines of the dolomitic mudstone units have very high alkalinities with a low pH of 5 to 6. The Mg concentration (2250 mmoles/l) is extremely elevated, whereas the Ca concentration is nearly zero. Finally, the drilled evaporite sedimentary sequence was interpreted as being deposited in an alkaline lake/sea ("Lago Mare"), which covered the area during the latest Messinian. Projecting forward 40 years since the DSDP Leg 42A drilling campaign, research into the factors controlling dolomite precipitation under Earth surface conditions has led to the development of new models involving the metabolism of microorganisms and associated biofilms to overcome the kinetic inhibitions associated with primary dolomite precipitation. Together with laboratory experiments, microbial

  3. Opaque minerals as keys for distinguishing oxidising and reducing diagenetic conditions in the Lower Triassic Bunter Sandstone, North German Basin

    Weibel, R.; Friis, H.

    2004-07-01

    The diagenetic evolution in red and white/drab parts of the Bunter Sandstone from outcrops and wells in the North German Basin was unraveled by a detailed petrographical, mineralogical, and chemical study. The diagenetic sequences for the red and white/drab parts differ in the degree of alteration of iron-bearing detrital minerals and in the authigenic opaque minerals. In the white/drab parts, one typically sees reduction of red mica, dissolution of hematite and magnetite, and alteration of ilmenite and titanomagnetite to leucoxene. In contrast, the red parts commonly display oxidation of glauconite; red zonation in ooids; hematisation of magnetite, biotite, and magnetite host in titanomagnetite; and leucoxene replacement of the ilmenite lamellae in titanomagnetite. The red colour of the red bed host is due to red coatings of goethite needles and/or hematite needles. Goethite needles precipitated in the prevailing oxidising conditions, present shortly after deposition, whereas pseudomorphous replacement by hematite occurred during the burial process. The white/drab colour of the reduced areas is the original colour of the sediment. The black core in reduction spots is due to mineralization. Uranium and vanadium minerals (coffinite, montroseite, and V-illite) in reduction spots are located close to carbonate fragments, which probably contained organic matter. Isolated reduction spots may have formed immediately after deposition; inside these were clasts eroded from algal mats formed in the sabkha environment. The copper minerals typically occur in areas with high porosity, reflecting the importance of access to chloride-rich brines, which transported dissolved copper. When these brines reached reducing environments, chalcocite and chalcopyrite were precipitated, possibly by replacement of pyrite. In the red areas, these saline brines may also have promoted the second-stage alteration of ilmenite, in which leucoxene, the first-stage alteration product, was

  4. Strategy generation in accident management support

    An increased interest for research in the field of Accident Management can be noted. Several international programmes have been started in order to be able to understand the basic physical and chemical phenomena in accident conditions. A feasibility study has shown that it would be possible to design and develop a computerized support system for plant staff in accident situations. To achieve this goal the Halden Project has initiated a research programme on Computerized Accident Management Support (CAMS project). The aim is to utilize the capabilities of computerized tools to support the plant staff during the various accident stages. The system will include identification of the accident state, assessment of the future development of the accident and planning of accident mitigation strategies. A prototype is developed to support operators and the Technical Support Centre in decision making during serious accident in nuclear power plants. A rule based system has been built to take care of the strategy generation. This system assists plant personnel in planning control proposals and mitigation strategies from normal operation to severe accident conditions. The ideal of a safety objective tree and knowledge from the emergency procedures have been used. Future prediction requires good state identification of the plant status and some knowledge about the history of some critical variables. The information needs to be validated as well. Accurate calculations in simulators and a large database including all important information form the plant will help the strategy planning. (author). 12 refs, 2 figs

  5. Criticality Accident

    At a meeting of electric utility presidents in October, 1999, the Federation Power Companies (FEPCO) officially decided to establish a Japanese version of WANO, following the JCO criticality accident. The Japanese WANO is expected to be launched by the end of the year: initially, with some 30 private sector companies concerned with nuclear fuel. It is said that the private sector had to make efforts to ensure that safety was the most important value in management policy throughout the industry, and that comprehensive inspections would be implemented. In anything related to nuclear energy, sufficient safety checks are required even for the most seemingly trivial matters. Therefore, the All-Japan Council of Local Governments with Atomic Power Stations has already proposed to the Japanese government that it should enact the special law for nuclear emergency, providing that the unified responsibility for nuclear disaster prevention should be shifted to the national government, since the nuclear disaster was quite special from the viewpoint of its safety regulation and technical aspects. (G.K.)

  6. Experimental investigation of capacity for work of protective outer skin of microfuel particles in reference of heavy accident conditions of lightwater reactors

    One investigated into the service ability of the silicon carbide coated particle fuels as applied to the light-water reactor loss-of-coolant severe accidents. According to the tests performed in the synthetic steam-and-gas medium representing the products of propane combustion in oxygen within 730-1670 deg C range, the particle fuel coating under up to 1590 deg C temperature showed high corrosion resistance and integrity. The particle fuel essential vulnerability to damage up to 25% was observed under 1670 deg C. The particle fuels satisfied advantageously the tests with up to 800-1600 deg C heating in air followed by the abrupt cooling in water

  7. ESFR Severe Accident Analyses with SIMMER-III

    The Collaborative Project on European Sodium Fast Reactor, CP-ESFR, combines European efforts advancing fast reactor technology towards economics, safety and nuclear waste reduction. A key issue of development is the promise of a higher and improved safety level. Both on the prevention and mitigation side significant efforts are invested to fulfill the high safety goals. Research in severe accident phenomenology and safety analyses help to develop means for better prevention and mitigation. Within this framework accident initiators are investigated leading to an unprotected loss-of-flow (ULOF) and a total instantaneous blockage (TIB) scenario. Simulations focusing on the energetics behavior apply SIMMER-III, an advanced accident code coupled with space- and energy-dependent neutronics. For the ULOF especially the transition phase with its recriticality potential has been of interest, while for the TIB the issue of melt propagation has been a key focus. In addition it has been investigated whether the available core material removal paths are sufficiently effective to prevent recriticality scenarios. The ULOF conditions for SIMMER have been provided by a SAS-SFR simulation of the ULOF initiation phase. For the TIB the SIMMER simulations started from steady state core conditions. (author)

  8. Simulations of the design basis accident at conditions of power increase and the o transient of MSIV at overpressure conditions of the Laguna Verde Power Station; Simulaciones del accidente base de diseno a condiciones de aumento de potencia y del transitorio de cierre de MSIV a condiciones de sobrepresion de la Central Laguna Verde

    Araiza M, E.; Nunez C, A. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    2001-07-01

    This document presents the analysis of the simulation of the loss of coolant accident at uprate power conditions, that is 2027 MWt (105% of the current rated power of 1931MWt). This power was reached allowing an increase in the turbine steam flow rate without changing the steam dome pressure value at its rated conditions (1020 psiaJ. There are also presented the results of the simulation of the main steam isolation va/ve transient at overpressure conditions 1065 psia and 1067 MWt), for Laguna Verde Nuclear Power Station. Both simulations were performed with the best estimate computer code TRA C BF1. The results obtained in the loss of coolant accident show that the emergency core coolant systems can recover the water level in the core before fuel temperature increases excessively, and that the peak pressure reached in the drywell is always below its design pressure. Therefore it is concluded that the integrity of the containment is not challenged during a loss of coolant accident at uprate power conditions.The analysis of the main steam isolation valve transients at overpressure conditions, and the analysis of the particular cases of the failure of one to six safety relief valves to open, show that the vessel peak pressures are below the design pressure and have no significant effect on vessel integrity. (Author)

  9. Persistence on airline accidents.

    L. A. GIL-ALANA; Barros, C.P. (Carlos P.); J.R. Faria

    2009-01-01

    This paper analyses airline accidents data from 1927-2006. The fractional integration methodology is adopted. It is shown that airline accidents are persistent and (fractionally) cointegrated with airline traffic. Thus, there exists an equilibrium relation between air accidents and airline traffic, with the effect of the shocks to that relationship disappearing in the long run. Policy implications are derived for countering accidents events.

  10. Persistence in Airline Accidents

    Carlos Pestana Barros; João Ricardo Faria; Luis A. Gil-Alana

    2008-01-01

    This paper analyses airline accident data from 1927-2006, through fractional integration. It is shown that airline accidents are persistent and (fractionally) cointegrated with airline traffic. There exists a negative relation between air accidents and airline traffic, with the effect of the shocks to that relationship disappearing in the long run. Policy implications are derived for countering accident events.

  11. Estimation of molecular carriers of electrons in mitochondria of adrenal cortex under conditions of long-term permanent action of low-intensive ionizing radiation after the accident at the Chernobyl NPP

    By EPR at 77K, we study molecular carriers of electrons in mitochondria of adrenal cortex in animals under conditions of the chronic action of low-dose ionizing irradiation from incorporated radionuclides related to the accident at the Chernobyl NPP and have found a significant decrease in the content of adrenodoxine which is a molecular carrier of electrons in the system of steroid hydroxylation. The last can play an important role in the mechanism of growth of neoplasms in adrenal glands, ovarial, and mammary glands

  12. Occupational Radiation Protection in Severe Accident Management

    protection job coverage during severe accident response. The IAEA defines a 'Severe Accident' as a beyond design basis accident comprising of accident conditions more severe than a design basis accident, involving significant core degradation. Preparation of the report The expert group met several times to share their experience and develop an interim (preliminary) report by the end of 2013. The content of the report is thus based on current reflections and action plans undertaken by the ISOE participating utilities and regulatory authorities to improve the emergency response plans in the event of a severe nuclear accident from the point of view of occupational radiation protection. A specific attention has been given to the analysis of past nuclear accidents (TMI-2, USA-1979; Chernobyl, USSR-1986 and Fukushima Daiichi, Japan-2011) and to the integration of the occupational radiation protection (ORP) lessons learned from these accidents into the various chapters of the report (See synthesis of these lessons learned in Appendix-1). To finalize the report, an international workshop was organized in 2014 to present and discuss the content of the interim version and share national experiences on best occupational RP management practices and protocols for optimum RP job coverage during severe accident, initial response and recovery efforts (see Appendix-2). The workshop notably allowed to improve and complete the report which has then be submitted to the ISOE Management Board for approval. This report comprises five main chapters. Chapter 2 provides essential information on radiation protection management and organisation. Chapter 3 establishes the goal of radiation protection training and exercises related to severe accident management. Chapter 4 discusses facility characteristics that must be considered when planning actions in response to a severe accident. Chapter 5 introduces an overall approach for the protection of workers / responders with its interpretation and

  13. Accident knowledge and emergency management

    Rasmussen, B.; Groenberg, C.D.

    1997-03-01

    The report contains an overall frame for transformation of knowledge and experience from risk analysis to emergency education. An accident model has been developed to describe the emergency situation. A key concept of this model is uncontrolled flow of energy (UFOE), essential elements are the state, location and movement of the energy (and mass). A UFOE can be considered as the driving force of an accident, e.g., an explosion, a fire, a release of heavy gases. As long as the energy is confined, i.e. the location and movement of the energy are under control, the situation is safe, but loss of confinement will create a hazardous situation that may develop into an accident. A domain model has been developed for representing accident and emergency scenarios occurring in society. The domain model uses three main categories: status, context and objectives. A domain is a group of activities with allied goals and elements and ten specific domains have been investigated: process plant, storage, nuclear power plant, energy distribution, marine transport of goods, marine transport of people, aviation, transport by road, transport by rail and natural disasters. Totally 25 accident cases were consulted and information was extracted for filling into the schematic representations with two to four cases pr. specific domain. (au) 41 tabs., 8 ills.; 79 refs.

  14. Accident knowledge and emergency management

    The report contains an overall frame for transformation of knowledge and experience from risk analysis to emergency education. An accident model has been developed to describe the emergency situation. A key concept of this model is uncontrolled flow of energy (UFOE), essential elements are the state, location and movement of the energy (and mass). A UFOE can be considered as the driving force of an accident, e.g., an explosion, a fire, a release of heavy gases. As long as the energy is confined, i.e. the location and movement of the energy are under control, the situation is safe, but loss of confinement will create a hazardous situation that may develop into an accident. A domain model has been developed for representing accident and emergency scenarios occurring in society. The domain model uses three main categories: status, context and objectives. A domain is a group of activities with allied goals and elements and ten specific domains have been investigated: process plant, storage, nuclear power plant, energy distribution, marine transport of goods, marine transport of people, aviation, transport by road, transport by rail and natural disasters. Totally 25 accident cases were consulted and information was extracted for filling into the schematic representations with two to four cases pr. specific domain. (au) 41 tabs., 8 ills.; 79 refs

  15. Severe accident phenomena

    Severe accidents are nuclear reactor accidents in which the reactor core is substantially damaged. The report describes severe reactor accident phenomena and their significance for the safety of nuclear power plants. A comprehensive set of phenomena ranging from accident initiation to containment behaviour and containment integrity questions are covered. The report is based on expertise gained in the severe accident assessment projects conducted at the Technical Research Centre of Finland (VTT). (49 refs., 32 figs., 12 tabs.)

  16. Air cleaning in accident situations

    The Organization for Economic Co-Operation and Development (OECD) through its subsidiaries the Nuclear Energy Agency (NEA) and the Committee on the Safety of Nuclear Installations (CSNI) established in 1979 a Group of Experts or Air Cleaning in Accident Situations. This group met seven times to establish a draft report based on its Terms of Reference which were to: 1) review the performance of off-gas cleaning systems in accident conditions; 2) collect information about operating experience with these systems; 3) seek to establish common principles for the design of off-gas systems; 4) review methods used in the different countries for testing filters from the standpoint of accident conditions; and 5) suggest specific mechanisms for improving cooperation, with regard, for example, to filter testing. The conclusions and recommendations of the Group are summarized

  17. A study on the use of neural network for severe accident management

    Based on the consensus that the course and consequence of a severe core damage accident can be greatly influenced by the operators' action, there have been extensive efforts to establish severe accident management program. A severe accident management process is essentially a sequence of decision making with a wide variety of available information under the highly uncertain condition, aimed at successful termination of accident progression or consequence minimization. For operators to take correct and timely accident management actions, they should be informed of the accident progression. Some key events, such as onset of core uncovery, core-melt initiation, reactor vessel lower head failure, containment failure, etc., act as landmarks for operators to make decisions in severe accident management process. Thus it is of critical importance to identify the timing at which such events occur in accident management. Unfortunately, it is difficult task partly due to phenomenological complexity and partly due to the lack of instrumentation reliability in severe accident environment, making the traditional procedural or rule-based approach inappropriate to be adopted to this end. Instead a technique, called artificial neural network, has been successfully applied to the similar problem domain out of various disciplines including nuclear industry. This paper presents a study on the application of a special kind of artificial neural network having the capability of recognizing time-varying patterns, called spatiotemporal network (STN), to the event timing prediction which is an important sub function of integrated computer supporting system for severe accident management. As the first trial, concentration was put on the identification of reactor vessel lower head failure which is considered the most critical events discriminating between so called in-vessel and ex-vessel accident management phases. Several sets of seven parameter signals from MAAP-based severe accident

  18. Nuclear accident dosimetry intercomparison studies.

    Sims, C S

    1989-09-01

    Twenty-two nuclear accident dosimetry intercomparison studies utilizing the fast-pulse Health Physics Research Reactor at the Oak Ridge National Laboratory have been conducted since 1965. These studies have provided a total of 62 different organizations a forum for discussion of criticality accident dosimetry, an opportunity to test their neutron and gamma-ray dosimetry systems under a variety of simulated criticality accident conditions, and the experience of comparing results with reference dose values as well as with the measured results obtained by others making measurements under identical conditions. Sixty-nine nuclear accidents (27 with unmoderated neutron energy spectra and 42 with eight different shielded spectra) have been simulated in the studies. Neutron doses were in the 0.2-8.5 Gy range and gamma doses in the 0.1-2.0 Gy range. A total of 2,289 dose measurements (1,311 neutron, 978 gamma) were made during the intercomparisons. The primary methods of neutron dosimetry were activation foils, thermoluminescent dosimeters, and blood sodium activation. The main methods of gamma dose measurement were thermoluminescent dosimeters, radiophotoluminescent glass, and film. About 68% of the neutron measurements met the accuracy guidelines (+/- 25%) and about 52% of the gamma measurements met the accuracy criterion (+/- 20%) for accident dosimetry. PMID:2777549

  19. Accidents and human factors

    When the TMI accident occurred it was 4 a.m., an hour when the error potential of the operators would have been very high. The frequency of car and train accidents in Japan is also highest between 4 a.m. and 6 a.m. The error potential may be classified into five phases corresponding to the electroencephalogramic pattern (EEG). At phase 0, when the delta wave appears, a person is unconscious and in deep sleep; at phase I, when the theta wave appears, he is very tired, sleepy and subnormal; at phase II, when the alpha wave appears, he is normal, relaxed and passive; at phase III, when the beta wave appears, he is normal, clear-minded and active; at phase IV, when the strong beta or epileptic wave appears, he is hypernormal, excited and incapable of normal judgement. Should an accident occur at phase II, the brain condition may jump to phase IV. At this phase the error or accident potential is maximum. The response of the human brain to different types of noises and signals may vary somewhat for different individuals and for different groups of people. Therefore, the possibility that such differences in brain functions may influence the mental structure would be worthy of consideration in human factors and in the design of man-machine systems. Human reliability and performance would be affected by many factors: medical, physiological and psychological, etc. The uncertainty involved in human factors may not necessarily be probabilistic, but fuzzy. Therefore, it would be important to develop a theory by which both non-probabilistic uncertainties, or fuzziness, of human factors and the probabilistic properties of machines can be treated consistently. From the mathematical point of view, probabilistic measure is considered a special case of fuzzy measure. Therefore, fuzzy set theory seems to be an effective tool for analysing man-machine systems. To minimize human error and the possibility of accidents, new safety systems should not only back up man and make up for his

  20. A new NEA expert group on accident-tolerant fuels

    collaboration in the development of core materials and designs that provide an improved tolerance to accidents. The start-up meeting for the new expert group was held at the NEA on 28-29 April 2014 and was attended by over 30 delegates from countries representing major LWR operators. It was agreed at this meeting that the programme of work would focus on the following key areas: system assessments, cladding and core materials, and fuel concept development. As part of the systems assessment programme, analyses will be made using state-of-the-art modelling and simulation methods to establish the most important parameters affecting accident tolerance and to rank the effectiveness of proposed concepts in the form of a performance metric. System performance under normal and accident reactor conditions, the impact on spent fuel management operations and the economic viability of new fuel designs are likely to be among the key components of this metric. As high-ranked candidate materials and fuel concepts emerge from this process, evaluations will be undertaken on the status of related technical readiness levels, including the availability of the experimental information needed to qualify performance and safety analyses. (author)

  1. Severe accident analysis using dynamic accident progression event trees

    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

  2. Short-term selective alleviation of glucotoxicity and lipotoxicity ameliorates the suppressed expression of key β-cell factors under diabetic conditions.

    Shimo, Naoki; Matsuoka, Taka-aki; Miyatsuka, Takeshi; Takebe, Satomi; Tochino, Yoshihiro; Takahara, Mitsuyoshi; Kaneto, Hideaki; Shimomura, Iichiro

    2015-11-27

    Alleviation of hyperglycaemia and hyperlipidemia improves pancreatic β-cell function in type 2 diabetes. However, the underlying molecular mechanisms are still not well clarified. In this study, we aimed to elucidate how the expression alterations of key β-cell factors are altered by the short-term selective alleviation of glucotoxicity or lipotoxicity. We treated db/db mice for one week with empagliflozin and/or bezafibrate to alleviate glucotoxicity and/or liptotoxicity, respectively. The gene expression levels of Pdx1 and Mafa, and their potential targets, insulin 1, Slc2a2, and Glp1r, were higher in the islets of empagliflozin-treated mice, and levels of insulin 2 were higher in mice treated with both reagents, than in untreated mice. Moreover, compared to the pretreatment levels, Mafa and insulin 1 expression increased in empagliflozin-treated mice, and Slc2a2 increased in combination-treated mice. In addition, empagliflozin treatment enhanced β-cell proliferation assessed by Ki-67 immunostaining. Our date clearly demonstrated that the one-week selective alleviation of glucotoxicity led to the better expression levels of the key β-cell factors critical for β-cell function over pretreatment levels, and that the alleviation of lipotoxicity along with glucotoxicity augmented the favorable effects under diabetic conditions. PMID:26471305

  3. REFORMATION OF THE GRAIN EXCHANGE OF RUSSIA, AS A KEY ELEMENT OF ADAPTATION OF GRAIN MULTINATIONAL CORPORATIONS IN THE CONDITIONS OF THE RUSSIAN ECONOMY

    Dudinova E. E.

    2014-04-01

    Full Text Available Reformation of the grain exchange of Russia becomes a key aspect of strengthening of the reached line items, and also an entry into the new regional markets due to the need of development of an export potential of Russia within the international grain market. In this work basic elements of upgrade of the existing grain mechanism are considered (on basis of JSC National Mercantile Exchange, examples are given from practical operating activities of CME Group Inc (Group of the Chicago board of trade. The conclusion that consolidation of the Russian Federation, Ukraine and the Republic of Kazakhstan under the auspices of a single grain platform becomes a priority in development of export trade in the Black Sea pool, especially in the conditions of the fixed competition for the markets

  4. Nuclear accident impact on the ecological environment

    This article reviewed the eco-environmental behavior of radionuclides released into the environment by nuclear explosion and nuclear accidents, especially of several key radionuclides with biological significance, including 137Cs, 95Zr, 90Sr, 131I, 3H and 14C, in order to correctly understand the case of nuclear accidents and its pollution, maintain the social stable, and provide suitable measures for environmental protection and safety. (author)

  5. 10 CFR 50.67 - Accident source term.

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Accident source term. 50.67 Section 50.67 Energy NUCLEAR... Conditions of Licenses and Construction Permits § 50.67 Accident source term. (a) Applicability. The... to January 10, 1997, who seek to revise the current accident source term used in their design...

  6. Porosity effects during a severe accident

    Cazares R, R. I. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Posgrado en Energia y Medio Ambiente, San Rafael Atlixco 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Espinosa P, G.; Vazquez R, A., E-mail: ricardo-cazares@hotmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, San Rafael Atlixco 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2015-09-15

    The aim of this work is to study the behaviour of porosity effects on the temporal evolution of the distributions of hydrogen concentration and temperature profiles in a fuel assembly where a stream of steam is flowing. The analysis considers the fuel element without mitigation effects. The mass transfer phenomenon considers that the hydrogen generated diffuses in the steam by convection and diffusion. Oxidation of the cladding, rods and other components in the core constructed in zirconium base alloy by steam is a critical issue in LWR accident producing severe core damage. The oxygen consumed by the zirconium is supplied by the up flow of steam from the water pool below the uncovered core, supplemented in the case of PWR by gas recirculation from the cooler outer regions of the core to hotter zones. Fuel rod cladding oxidation is then one of the key phenomena influencing the core behavior under high-temperature accident conditions. The chemical reaction of oxidation is highly exothermic, which determines the hydrogen rate generation and the cladding brittleness and degradation. The heat transfer process in the fuel assembly is considered with a reduced order model. The Boussinesq approximation was applied in the momentum equations for multicomponent flow analysis that considers natural convection due to buoyancy forces, which is related with thermal and hydrogen concentration effects. The numerical simulation was carried out in an averaging channel that represents a core reactor with the fuel rod with its gap and cladding and cooling steam of a BWR. (Author)

  7. Monitoring severe accidents using AI techniques

    After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

  8. Accident selection methodology for TA-55 FSAR

    In the past, the selection of representative accidents for refined analysis from the numerous scenarios identified in hazards analyses (HAs) has involved significant judgment and has been difficult to defend. As part of upgrading the Final Safety Analysis Report (FSAR) for the TA-55 plutonium facility at the Los Alamos National Laboratory, an accident selection process was developed that is mostly mechanical and reproducible in nature and fulfills the requirements of the Department of Energy (DOE) Standard 3009 and DOE Order 5480.23. Among the objectives specified by this guidance are the requirements that accident screening (1) consider accidents during normal and abnormal operating conditions, (2) consider both design basis and beyond design basis accidents, (3) characterize accidents by category (operational, natural phenomena, etc.) and by type (spill, explosion, fire, etc.), and (4) identify accidents that bound all foreseeable accident types. The accident selection process described here in the context of the TA-55 FSAR is applicable to all types of DOE facilities

  9. Fukushima accident study using MELCOR

    Randall O Gauntt

    2013-01-01

    The accidents at the Fukushima Daiichi nuclear power station stunned the world as the sequences played out over severals days and videos of hydrogen explosions were televised as they took place.The accidents all resulted in severe damage to the reactor cores and releases of radioactivity to the environment despite heroic measures had taken by the operating personnel.The following paper provides some background into the development of these accidents and their root causes,chief among them,the prolonged station blackout conditions that isolated the reactors from their ultimate heat sink — the ocean.The interpretations given in this paper are summarized from a recently completed report funded by the United States Department of Energy (USDOE).

  10. Experimental Program QUENCH at KIT on Core Degradation During Reflooding Under LOCA Conditions and in the Early Phase of a Severe Accident

    The most important accident management measure to terminate a severe accident transient in LWR is the injection of water to cool the uncovered degraded core. In order to detailed investigation of the reflood effect on bundle degradation the QUENCH program was initiated in 1996 followed-up the CORA bundle tests and is still on-going. So far, 17 integral bundle QUENCH experiments with 21–31 electrically heated fuel rod simulators of 2.5 m length using zirconia as fuel substitute have been conducted. Influence of following parameters on bundle degradation were investigated: degree of pre-oxidation, temperature at reflood initiation, flooding rate, effect of neutron absorber materials (B4C, Ag-In-Cd), air ingress, influence of the type of cladding alloy, formation of a debris bed in the core. Integral bundle experiments are supported by separate-effects tests. The program provides experimental data for the development of quench-related models and for the validation of SFD code systems. In seven tests, reflooding of the bundle led to a temporary temperature excursion. Considerable formation, relocation, and oxidation of melt were observed in all tests with escalation. The temperature boundary between rapid cool down and temperature escalation was typically 2100–2200 K in tests without absorber. Tests with absorber led to temperature escalations at lower temperatures. Although separate-effects tests have shown some differences in oxidation kinetics of advanced cladding materials, the influence of the various cladding alloys on the integral bundle behaviour during oxidation and reflooding was only limited. The two bundle tests with air ingress phase confirmed the strong effect of air on core degradation especially when pre-oxidation in steam is limited and oxygen starvation occurs during the air ingress phase. Oxidation in a nitrogen-containing atmosphere accelerates the kinetics by the temporary formation of zirconium nitride and causes strongly degraded and non

  11. Compilation and response key points of emergency plan for dangerous chemical poisoning accident attention and response points%危险化学品中毒事故应急预案编制与响应关注要点

    邢娟娟

    2011-01-01

    大工业的发展导致各类职业中毒事件的发生日益严重,除人员伤亡等影响外,也会对于社会公众安全带来影响.应急预案编制中强调应急准备的重要性和针对性.职业危害事故识别和分类分级处理以及对于社会公共安全影响的危机处理能力是应急工作中应该关注的重要内容.应急响应中强调第一目击者的处理能力的培养、现场的实时监测、事故报告与事态演变的持续预警、有效应急资源应对危机处理的能力.%The development of great industry lead to the occurrence of all kinds of occupational poisoning event, in addition to the increasingly serious casualties, the social public security will also be affected. Emergency plancom-pilation stressed in the the importance and relevance of emergency preparedness. Occupational hazard accidents I-dentification, classification and grading disposal, and the ability of handling crisis for public safety influence were the important content that should be focused on in emergency work. Emergency response stressed in the training of first witnesses processing ability that should be focused on in emergency work, on-site real-time monitoring, the accident report and the early warning of evolution, and the ability to effectively deal with crisis disposal of emergency resources.

  12. SEVERE ACCIDENT MANAGEMENT TRAINING

    The purpose of this paper is (a) to define the International Atomic Energy Agency's role in the area of severe accident management training, (b) to briefly describe the status of representative severe accident analysis tools designed to support development and validation of accident management guidelines, and more recently, simulate the accident with sufficient accuracy to support the training of technical support and reactor operator staff, and (c) provide an overview of representative design-specific accident management guidelines and training. Since accident management and the development of accident management validation and training software is a rapidly evolving area, this paper is also intended to evolve as accident management guidelines and training programs are developed to meet different reactor design requirements and individual national requirements

  13. On closure strategy for 1-D thermohydraulics models and closure relationships of two-phase flows in simple and subchannel geometry for NPP accident conditions

    One-dimensional mathematical models are extensively used in thermohydraulics assessment of Nuclear Power Plant (NPP) transients and accidents, because specifically 1-D system of the conservation laws allows to reduce computing time and required memory, especially in ''best estimate'' code calculations. This work is generalization of the well-known Zuber-Findley and Hancox-Nicoll methods for two-phase flow distribution parameters Cs taking into account the non-monotonous void fraction distribution in the transverse direction in terms of two superimposed monotonous profiles. The method is very useful in evaluating the saddle-shape void fraction profile effects. In this work two-phase flow distribution parameters Cs were developed for simple circular and rectangular pipes, and subchannel geometry in a rod bundle. Basic assumptions were power-mode approximations for describing the profiles of local volume flux density, phase velocity and temperature. The general analytical (quadrature) relationships for Cs were obtained and their 3-D illustrations are proposed. Also, we propose generalized formulation and simple approach to construct friction factor, heat and mass transfer coefficients within the gradient hypothesis and boundary layer assumptions. The contribution of momentum, heat and mass transfer as well as their sources and sinks in the channel cross-section are taken into account. In the same way, the friction factor, heat and mass transfer coefficients with the transversal and azimuthal variations being taken into account are proposed for subchannel geometry as well. (author)

  14. Influence of boron vapor on transport behavior of deposited CsI during heating test simulating a BWR severe accident condition

    Sato, Isamu, E-mail: sato.isamu@jaea.go.jp; Onishi, Takashi; Tanaka, Kosuke; Iwasaki, Maho; Koyama, Shin-ichi

    2015-06-15

    In order to evaluate influence of B on the release and transport of Cs and I during severe accidents, basic experiments have been performed on the interaction between deposited Cs/I compounds and vapor/aerosol B compounds. CsI and B{sub 2}O{sub 3} were utilized as a Cs/I compound and a B compound, respectively. Deposited CsI on the thermal gradient tube (TGT) at temperatures ranging from 423 K to 1023 K was reacted with vapor/aerosol B{sub 2}O{sub 3}, and then observed how it changed Cs/I deposition profiles. As a result, vapor/aerosol B{sub 2}O{sub 3} stripped a portion of deposited CsI within a temperature range from 830 K to 920 K to make gaseous CsBO{sub 2} and I{sub 2}. In addition, gaseous I{sub 2} was re-deposited at a temperature range from 530 K to 740 K, while CsBO{sub 2} travelled through the sampling tubes and filters without deposition. It is evident that B enables Cs compounds such as CsBO{sub 2} to transport Cs to the colder regions.

  15. Traffic Congestion and Accidents

    Schrage, Andrea

    2006-01-01

    Obstructions caused by accidents can trigger or exacerbate traffic congestion. This paper derives the efficient traffic pattern for a rush hour with congestion and accidents and the corresponding road toll. Compared to the model without accidents, where the toll equals external costs imposed on drivers using the road at the same time, a new insight arises: An optimal toll also internalizes the expected increase in future congestion costs. Since accidents affect more drivers if traffic volumes...

  16. Psychology of nuclear accidents

    Tysoe, M.

    1983-03-31

    Incidents involving nuclear weapons are described, as well as the accident to the Three Mile Island-2 reactor. Methods of assessment of risks are discussed, with particular reference to subjective judgements and the possible role of human error in civil nuclear accidents. Accidents or misunderstandings in communication or human actions which might lead to nuclear war are also discussed.

  17. Consistency in accident analyses in DOE safety, environmental, and emergency planning documents

    A consistency review of accident analyses in US Department of Energy (DOE) safety, environmental, and emergency planning documents is presented. The range of and key differences in driving assumptions used in accident definition and frequency assessment, radiological source term generation, and atmospheric transport and fate modeling across recent environmental impact statements (EISs) and emergency planning documents and the effects of these differences on results are summarized. Considerable variation in both the assumptions and the underlying level of conservatism is shown to exist. Recommendations are made for source term generation and assumed meteorological conditions to reduce inconsistencies without being overly prescriptive. Recommendations also are made to improve consistency in assessing the frequencies of various generic accident sequences traditionally analyzed in EIS and emergency planning documents. All recommendations are shown to be consistent with currently applicable DOE guidance

  18. Insights from Severe Accident Analyses for Verification of VVER SAMG

    The severe accident analyses of simultaneous rupture of all four steam lines (case-a), simultaneous occurrence of LOCA with SBO (case-b) and Station blackout (case-c) were performed with the computer code ASTEC V2r2 for a typical VVER-1000. The results obtained will be used for verification of sever accident provisions and Severe Accident Management Guidelines (SAMG). Auxiliary feed water and emergency core cooling systems are modelled as boundary conditions. The ICARE module is used to simulate the reactor core, which is divided into five radial regions by grouping similarly powered fuel assemblies together. Initially, CESAR module computes thermal hydraulics in primary and secondary circuits. As soon as core uncovery begins, the ICARE module is actuated based on certain parameters, and after this, ICARE module computes the thermal hydraulics in the core, bypass, downcomer and the lower plenum. CESAR handles the remaining components in the primary and secondary loops. CPA module is used to simulate the containment and to predict the thermal-hydraulic and hydrogen behaviour in the containment. The accident sequences were selected in such a way that they cover low/high pressure and slow/fast core damage progression events. Events simulated included slow progression events with high pressure and fast accident progression with low primary pressure. Analysis was also carried out for the case of SBO with the opening of the PORVs when core exit temperature exceeds certain value as part of SAMG. Time step sensitivity study was carried out for LOCA with SBO. In general the trends and magnitude of the parameters are as expected. The key results of the above analyses are presented in this paper. (author)

  19. Supervisor's accident investigation handbook

    This pamphlet was prepared by the Environmental Health and Safety Department (EH and S) of Lawrence Berkeley Laboratory (LBL) to provide LBL supervisors with a handy reference to LBL's accident investigation program. The publication supplements the Accident and Emergencies section of LBL's Regulations and Procedures Manual, Pub. 201. The present guide discusses only accidents that are to be investigated by the supervisor. These accidents are classified as Type C by the Department of Energy (DOE) and include most occupational injuries and illnesses, government motor-vehicle accidents, and property damages of less than $50,000

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

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

  1. General Aspects of the JCO Criticality Accident

    A criticality accident occurred on September 30, 1999, at a uranium processing plant of JCO Company in Tokaimura. Delayed criticality continued for approximately 20 hours after the first few prompt critical peaks. Two employees subsequently died. Nearby residents were evacuated or told to remain indoors. This accident was at Level 4 on the International Nuclear Event Scale. A table of radiation exposures resulting from the accident is given. Besides dealing with health physics, the investigation committee's final report covered technical observations and the nature of the accident. The direct causes of the accident were found to be violation of rules and technical specifications and deviation from licensing conditions; some of these were permitted by the company itself, and fatal mistakes were made by employees on the job without consulting with authorized persons. Many recommendations to revise government regulations on licensing of nuclear fuel handling were discussed in the report

  2. Accidents, probabilities and consequences

    Following brief discussion of the safety of wind-driven power plants and solar power plants, some aspects of the safety of fast breeder and thermonuclear power plants are presented. It is pointed out that no safety evaluation of breeders comparable to the Rasmussen investigation has been carried out and that discussion of the safety aspects of thermonuclear power is only just begun. Finally, as an illustration of the varying interpretations of risk and safety analyses, four examples are given of predicted probabilities and consequences in Copenhagen of the maximum credible accident at the Barsebaeck plant, under the most unfavourable meterological conditions. These are made by the Environment Commission, Risoe Research Establishment, REO (a pro-nuclear group) and OOA (an anti-nuclear group), and vary by a factor of over 1000. (JIW)

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

    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).

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

    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).

  5. Assessment of the dispersion of fission products in the atmosphere following a reactor accident under meteorological conditions of low wind speed

    The aim of the study is the assessment of the dispersion in a low speed situation and the validation of the computer code ICAIR3 by means of SF6 tracing experiments carried out on the CADARACHE site under different stability conditions. The results show clearly some characteristic features of the dispersion. In particular, high concentrations are found in the experimental field several hours after the end of the release. Large differences of the plume width are observed depending on the atmospheric stability. The flow seems well organized under stable conditions, probably in relation with a topographic effect (CADARACHE is situated in a valley), while there is a much larger spread out of the plume in neutral or unstable conditions. A reasonable agreement with the values predicted by the calculation code is found for the maximum concentration

  6. Framework for accident management

    A program is being conducted to establish those attributes of a severe accident management plan which are necessary to assure effective response to all credible severe accidents and to develop guidance for their incorporation in a plant's Accident Management Plan. This program is one part of the Accident Management Research Program being conducted by the U. S. Nuclear Regulatory Commission (NRC). The approach used in establishing attributes and developing guidance includes three steps. In the first step the general attributes of an accident management plan were identified based on: (1) the objectives established for the NRC accident management program, (2) the elements of an accident management framework identified by the NRC, and (3) a review of the processes used in developing the currently used approach for classifying and analyzing accidents. For the second step, a process was defined that uses the general attributes identified from the first step to develop an accident management plan. The third step applied the process defined in the second step at a nuclear power plant to refine and develop it into a benchmark accident management plan. Step one is completed, step two is underway and step three has not yet begun

  7. Visualization of Traffic Accidents

    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.

  8. Determination of accident related release data

    For accident safety analyses, for the assessment of potential radiological consequences, for the review of current requirements of the Transport Regulations and for their possible further development as well as for the demonstration that radioactive materials such as LDM candidate material fulfil the regulatory requirements reliable release data following mechanical impact are required. This is definitely one of the demanding issues in the field of transport safety of radioactive materials. In this context special attention has to be paid to radioactive wastes immobilised in brittle materials, e.g. cement/concrete, glass, ceramics or other brittle materials such as fresh and spent fuel. In this presentation we report on a long-term experimental program aiming at improving the general physical understanding of the release process as well as the quantity and the quality of release data. By combining laboratory experiments using small scale test specimens with a few key scaling experiments with large scale test objects significant progress was achieved to meet this objective. The laboratory equipment enables the in-situ determination of the amount and aerodynamic size distribution of the airborne particles generated upon impact of the test specimen on a hard target. Impact energies cover the range experienced in transport accidents including aircraft accidents. The well defined experimental boundary conditions and the good reproducibility of the experimental procedure allowed for systematic studies to exactly measure the amount and aerodynamic size distribution of the airborne release and to quantify its dependence on relevant parameters such as energy input, material properties, and specimen geometry. The experimental program was performed within the scope of various national and international (e.g. EU-funded) projects. The small scale experiments with brittle materials revealed a pronounced universality of the airborne release in view of the material properties and

  9. Deepwater Horizon Accident Investigation Report

    NONE

    2010-09-15

    On the evening of April 20, 2010, a well control event allowed hydrocarbons to escape from the Macondo well onto Transocean's Deepwater Horizon, resulting in explosions and fire on the rig. Eleven people lost their lives, and 17 others were injured. The fire, which was fed by hydrocarbons from the well, continued for 36 hours until the rig sank. Hydrocarbons continued to flow from the reservoir through the wellbore and the blowout preventer (BOP) for 87 days, causing a spill of national significance. BP Exploration and Production Inc. was the lease operator of Mississippi Canyon Block 252, which contains the Macondo well. BP formed an investigation team that was charged with gathering the facts surrounding the accident, analyzing available information to identify possible causes and making recommendations to enable prevention of similar accidents in the future. The BP investigation team began its work immediately in the aftermath of the accident, working independently from other BP spill response activities and organizations. The ability to gather information was limited by a scarcity of physical evidence and restricted access to potentially relevant witnesses. The team had access to partial real-time data from the rig, documents from various aspects of the Macondo well's development and construction, witness interviews and testimony from public hearings. The team used the information that was made available by other companies, including Transocean, Halliburton and Cameron. Over the course of the investigation, the team involved over 50 internal and external specialists from a variety of fields: safety, operations, subsea, drilling, well control, cementing, well flow dynamic modeling, BOP systems and process hazard analysis. This report presents an analysis of the events leading up to the accident, eight key findings related to the causal chain of events and recommendations to enable the prevention of a similar accident. The investigation team worked

  10. Deepwater Horizon Accident Investigation Report

    On the evening of April 20, 2010, a well control event allowed hydrocarbons to escape from the Macondo well onto Transocean's Deepwater Horizon, resulting in explosions and fire on the rig. Eleven people lost their lives, and 17 others were injured. The fire, which was fed by hydrocarbons from the well, continued for 36 hours until the rig sank. Hydrocarbons continued to flow from the reservoir through the wellbore and the blowout preventer (BOP) for 87 days, causing a spill of national significance. BP Exploration and Production Inc. was the lease operator of Mississippi Canyon Block 252, which contains the Macondo well. BP formed an investigation team that was charged with gathering the facts surrounding the accident, analyzing available information to identify possible causes and making recommendations to enable prevention of similar accidents in the future. The BP investigation team began its work immediately in the aftermath of the accident, working independently from other BP spill response activities and organizations. The ability to gather information was limited by a scarcity of physical evidence and restricted access to potentially relevant witnesses. The team had access to partial real-time data from the rig, documents from various aspects of the Macondo well's development and construction, witness interviews and testimony from public hearings. The team used the information that was made available by other companies, including Transocean, Halliburton and Cameron. Over the course of the investigation, the team involved over 50 internal and external specialists from a variety of fields: safety, operations, subsea, drilling, well control, cementing, well flow dynamic modeling, BOP systems and process hazard analysis. This report presents an analysis of the events leading up to the accident, eight key findings related to the causal chain of events and recommendations to enable the prevention of a similar accident. The investigation team worked separately

  11. Standby after the Chernobyl accident

    The report is an investigation concerning strandby and actions by SKI (Swedish Nuclear Power Inspectorate) and SSI (National Institute of Radiation Protection) due to the Chernobyl reactor accident. It consists of a final report and two appendices. The final report is divided into two parts: 'I: Facts' and 'II: Analyzes'. 'Facts': The Swedish model for information: radio, press. Basic knowledge about ionizing radiation in the society. Resources for information. Need for information. Message forms for information. Announcements from the authorities in TV, radio, press, meeting, advertisements. Statements concerning the reactor accident and its consequences in Swedish mass media. How did the public recieve the information? 'Analyzis': Information responsibilities and policies. SSI information activities concerning radiologic accidents, conditions, methods and resources. Ditto for SKI, Swedish National Food Administration and the National Board of Agriculture. Appendix I: Information from authorities in the press three weeks after the Chernobyl accident: The material and the methods. The acute phase, the adoptation phase, the extension of the persective. What is said about the authorities in connection with Chernobyl? Appendix II: The fallout from Chernobyl, the authorities and the media coverage: The nationwide, regional and local coverage from radio and television. Ditto from the press. Topic and problem areas in reporting. Instructions from the authorities in media. Contribution in the media from people representing the authorities. Fallout in a chronologic perspective. (L.F.)

  12. Development of criticality accident analysis code AGNES

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

  13. Assessment of the MSLB accident safety margins for NPP Krsko

    This paper presents the comparison between the 3-D neutronics (coupled code RELAP5/QUABOX/CUBBOX) and point kinetics (RELAP5/mod3.2.2) calculations of MSLB accident for NPP Krsko (NEK) after steam generator (SG) replacement and at uprated conditions (6% power increase). The main steam line break (MSLB) accident in pressurized water reactors (PWR) is overcooling accident characterized by large variations in primary coolant conditions, asymmetric core inlet and outlet conditions and strongly localized reactivity disturbance due to assumed stuck control rod. The characteristics and phenomena related to the MSLB accident for best-estimate calculations can be accurately analyzed through the use of coupled code. (author)

  14. Laser accidents: Being Prepared

    Barat, K

    2003-01-24

    The goal of the Laser Safety Officer and any laser safety program is to prevent a laser accident from occurring, in particular an injury to a person's eyes. Most laser safety courses talk about laser accidents, causes, and types of injury. The purpose of this presentation is to present a plan for safety offices and users to follow in case of accident or injury from laser radiation.

  15. The Chernobyl accident consequences

    Five teen years later, Tchernobyl remains the symbol of the greater industrial nuclear accident. To take stock on this accident, this paper proposes a chronology of the events and presents the opinion of many international and national organizations. It provides also web sites references concerning the environmental and sanitary consequences of the Tchernobyl accident, the economic actions and propositions for the nuclear safety improvement in the East Europe. (A.L.B.)

  16. Communication and industrial accidents

    As, Sicco van

    2001-01-01

    This paper deals with the influence of organizational communication on safety. Accidents are actually caused by individual mistakes. However the underlying causes of accidents are often organizational. As a link between these two levels - the organizational failures and mistakes - I suggest the concept of role distance, which emphasizes the organizational characteristics. The general hypothesis is that communication failures are a main cause of role distance and accident-proneness within orga...

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

    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

  18. Comparison of HRA methods based on WWER-1000 NPP real and simulated accident scenarios

    Full text: Adequate treatment of human interactions in probabilistic safety analysis (PSA) studies is a key to the understanding of accident sequences and their relative importance in overall risk. Human interactions with machines have long been recognized as important contributors to the safe operation of nuclear power plants (NPP). Human interactions affect the ordering of dominant accident sequences and hence have a significant effect on the risk of NPP. By virtue of the ability to combine the treatment of both human and hardware reliability in real accidents, NPP fullscope, multifunctional and computer-based simulators provide a unique way of developing an understanding of the importance of specific human actions for overall plant safety. Context dependent human reliability assessment (HRA) models, such as the holistic decision tree (HDT) and performance evaluation of teamwork (PET) methods, are the so-called second generation HRA techniques. The HDT model has been used for a number of PSA studies. The PET method reflects promising prospects for dealing with dynamic aspects of human performance. The paper presents a comparison of the two HRA techniques for calculation of post-accident human error probability in the PSA. The real and simulated event training scenario 'turbine's stop after loss of feedwater' based on standard PSA model assumptions is designed for WWER-1000 computer simulator and their detailed boundary conditions are described and analyzed. The error probability of post-accident individual actions will be calculated by means of each investigated technique based on student's computer simulator training archives

  19. Unconventional sources of plant information for accident management

    value and rate of change of key plant parameters can be directly measured or inferred from plant instrumentation. Even when the actual measured values for a particular instrument are less accurate than normally required, valuable information may still be obtained. Trends observed in the time-dependent behavior of that instrument can provide valuable information in how the accident is progressing. The failure of instruments themselves (especially those for which defined service condition requirements exist ) can yield information about plant status. Selected plant-specific examples of unconventional uses of plant information sources will be discussed in detail. Monitoring of secondary side cooling water system temperature can be used to infer core status and the ability to successfully remove heat to the ultimate heat sink. Containment/reactor vessel temperature can be used to infer core status and heat removal requirements. Tank pressure indication (e.g. pressurizer relief tank, waste gas surge tank) can be used to infer containment status. Tank level indication can be used to infer loss of inventory, system or containment over-pressurization. Water temperature increases, pump discharge pressures and pressure drops across pumps can be used to infer containment status. Use of process or area radiation monitors can be used to infer core status and fission product releases. Considerations regarding the use of unconventional information sources will be discussed. Plant- specific operator actions and procedural requirements may have significant impact on assessment of plant status. Instrument range capabilities may be significantly larger than the required/indicated range. Instrument accuracy requirements during an accident may be less stringent, allowing the use of marginally functional instruments. Information from locally indicated instrumentation may be unavailable due to harsh environments. (However, the presence of such environments would represent additional valuable

  20. Nuclear accidents and epidemiology

    A consultation on epidemiology related to the Chernobyl accident was held in Copenhagen in May 1987 as a basis for concerted action. This was followed by a joint IAEA/WHO workshop in Vienna, which reviewed appropriate methodologies for possible long-term effects of radiation following nuclear accidents. The reports of these two meetings are included in this volume, and cover the subjects: 1) Epidemiology related to the Chernobyl nuclear accident. 2) Appropriate methodologies for studying possible long-term effects of radiation on individuals exposed in a nuclear accident. Figs and tabs

  1. Perspective on post-Fukushima severe accident research

    After the Fukushima Daiichi accident in March 2011 several investigation committees issued reports with lessons learned from the accident, in which some recommendations on severe accident research are included. The review of specific severe accident research items had already started before Fukushima accident in working group of Atomic Energy Society of Japan (AESJ) in terms of significance of consequences, uncertainties of phenomena and maturity of assessment methodology. Re-investigation started after the Fukushima accident in this working group to cover additional effects of Fukushima accident, such as core degradation behaviors, sea water injection, containment failure/leakage and re-criticality. The review results are categorized in nine major fields; core degradation behavior, core melt coolability/retention in containment vessel, function of containment vessel, source term, hydrogen behavior, fuel-coolant interaction, molten core concrete interaction, recriticality and instrumentation in severe accident conditions. In January 2012, in collaboration with this working group, Research Expert Committee on Evaluation of Severe Accident was established in AESJ in order to investigate severe accident related issues for future LWR development. Based on these activities and also author's personal view, the present paper describes the seven important severe accident research issues after Fukushima accident. They are (1) investigation of damaged core and components, (2) advanced severe accident analysis capabilities and associated experimental investigations, (3) development of reliable passive cooling system for core/containment, (4) analysis of hydrogen behavior and investigation of hydrogen measures, (5) enhancement of removal function of radioactive materials of containment venting, (6) advanced instrumentation for the diagnosis of severe accident and (7) assessment of advanced containment design which exchides long-term evacuation in any severe accident situations

  2. Accident analysis in nuclear power plants

    The way the philosophy of Safety in Depth can be verified through the analysis of simulated accidents is shown. This can be achieved by verifying that the integrity of the protection barriers against the release of radioactivity to the environment is preserved even during accident conditions. The simulation of LOCA is focalized as an example, including a study about the associated environmental radiological consequences. (Author)

  3. Vehicle accidents related to sleep: a review

    Horne, J.; Reyner, L.

    1999-01-01

    Falling asleep while driving accounts for a considerable proportion of vehicle accidents under monotonous driving conditions. Many of these accidents are related to work--for example, drivers of lorries, goods vehicles, and company cars. Time of day (circadian) effects are profound, with sleepiness being particularly evident during night shift work, and driving home afterwards. Circadian factors are as important in determining driver sleepiness as is the duration of the drive, but only ...

  4. The Fukushima Daiichi Accident. Technical Volume 2/5. Safety Assessment

    Technical Volume 1 of this report has described what happened during the accident at the Fukushima Daiichi nuclear power plant (NPP). This volume begins (Section 2.1) with a review of how the design basis of the site for external events was assessed initially and then reassessed over the life of the NPP. The section also describes the physical changes that were made to the units as a result. The remainder of the volume describes the treatment of beyond design basis events in the safety assessment of the site, the accident management provisions, the effectiveness of regulatory programmes, human and organizational factors and the safety culture, and the role of operating experience. Further background information is contained in three annexes included on the CD-ROM of this Technical Volume which describe analytical investigations of the accident along with information on topics such as system performance, defence in depth and severe accident phenomena. Section 2.2 provides an assessment of the systems that failed, resulting in a failure to maintain the fundamental safety functions in Units 1–3, which were in operation at the time of the tsunami and in which the reactor pressure vessels (RPV) and containment vessels failed. The section also describes Units 4-6, which were shut down at the time of the tsunami, and the site’s central spent fuel storage facility. Section 2.3 discusses the probabilistic and deterministic safety assessments of beyond design basis accidents (BDBAs) that had been performed for the plant and the insights from these assessments that had led to changes in the plant’s design. The section pays particular attention to the assessment of extreme natural hazards, such as the one which led to the total loss of AC power supply on the site. The additional loss of DC power supply in Units 1 and 2 played a key role in the progression of the accident because it impeded the diagnosis of plant conditions and made the operators unaware of the status of

  5. The nuclear accident risk: a territorial approach

    How many people live in the vicinity of French nuclear power stations? Recent events - notably in Japan, but also in France - highlight the urgent need to be able to predict the possible effects of a nuclear accident on surrounding territories. Here, Ambroise Pascal identifies two key criteria for such an estimation: residential density and land use. (author)

  6. Severe accident risk minimization studies for the Advanced Neutron Source (ANS) reactor plant at the Oak Ridge National Laboratory

    This paper discusses salient aspects of severe accident related phenomenological considerations, scoping studies, and mitigative design features being studied for incorporation into a high-power research reactor plant. Key results of scoping studies on steam explosions, recriticality, core-concrete interactions, and containment transport are highlighted. Evolving design features of the containment are described. Containment response calculations for a site-suitability basis transient are presented that demonstrate acceptable source term values and superior containment performance. Oak Ridge National Laboratory's (ORNL) Advanced Neutron Source (ANS) will be a new user facility for all kinds of neutron research, centered around a research reactor of unprecedented neutron beam flux. A defense-in-depth philosophy has been adopted. In response to this commitment, ANS Project management initiated severe accident analysis and related technology development early-on in the design phase itself. This was done to aid in designing a sufficiently robust containment for retention and controlled release of radionuclides in the event of such an accident. It also provides a means for satisfying on- and off-site regulatory requirements, accident-related dose exposures, and containment response and source-term best-estimate analyses for level-2 and -3 Probabilistic Risk Analysis (PRAs) that will be produced. Moreover, it will provide the best possible understanding of the ANS under severe accident conditions and consequently provide insights for the development of strategies and design philosophies for accident mitigation, management, and emergency preparedness efforts

  7. The radioecology of the Chernobyl fallout. Interim report of a scientific analysis of conditions in an area far away from the reactor accident

    Knowledge about the degree of Cs contamination of all animal fodder of the Mariensee 1986 harvest in combination with the transfer factors determined under field conditions: feed - milk or meat of cattle and pork - made it possible to develop a strategy for further measures. The taking out of cattle to the pasture was delayed until the end of May 1986 as sufficient winter fodder from the 1985 harvest was still available, in order to reduce the medium Cs content in milk. Besides, the effects of the coming winter feeding 86/87 on the milk and meat of the Mariensee production were estimated to be less serious and recommendations for the partially considerably loaded singular areas of South Germany which became known in the meantime, could be published. (orig./DG)

  8. The size-dependent charge fraction of sub-3-nm particles as a key diagnostic of competitive nucleation mechanisms under atmospheric conditions

    Yu, F.; Turco, R. P.

    2011-09-01

    A clear physical understanding of atmospheric particle nucleation mechanisms is critical in assessing the influences of aerosols on climate and climate variability. Currently, several mechanisms have been proposed and are being employed to interpret field observations of nucleation events. Roughly speaking, the two most likely candidates are neutral cluster nucleation (NCN) and ion-mediated nucleation (IMN). Detailed nucleation event data has been obtained in boreal forests. In one set of analyses of these measurements, NCN was suggested as the dominant formation mode, while in another, it was IMN. Information on the electrical charge distribution carried by the nucleating clusters is one key for identifying the relative contributions of neutral and ion-mediated processes under various conditions. Fortunately, ground-breaking measurements of the charged states or fractions of ambient nanometer-sized particles soon after undergoing nucleation are now available to help resolve the main pathways. In the present study, the size-dependent "apparent" formation rates and fractions of charged and neutral particles in a boreal forest setting are simulated with a detailed kinetic model. We show that the predicted "apparent" formation rates of charged and neutral particles at 2 nm for eight representative case study days agree well with the corresponding values based on observations. In the simulations, the "apparent" contribution of ion-based nucleation increases by up to ~one order of magnitude as the size of "sampled" particles is decreased from 2 nm to ~1.5 nm. These results suggest that most of the neutral particles sampled in the field at sizes around 2 nm are in reality initially formed on ionic cores that are neutralized before the particles grow to this size. Thus, although the apparent rate of formation of neutral 2-nm particles might seem to be dominated by a neutral clustering process, in fact those particles may be largely the result of an ion-induced nucleation

  9. Severe accident testing of electrical penetration assemblies

    This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs

  10. Severe accident testing of electrical penetration assemblies

    Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))

    1989-11-01

    This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs.

  11. Criticality accident in Argentina

    A recent criticality type accident, ocurred in Argetina, is commented. Considerations about the nature of the facility where this accident took place, its genesis, type of operation carried out on the day of the event, and the medical aspects involved are done. (Author)

  12. Chernobyl accident and Danmark

    The report describes the Chernobyl accident and its consequences for Denmark in particular. It was commissioned by the Secretary of State for the Environment. Volume 1 contains copies of original documents issued by Danish authorities during the first accident phase and afterwards. Evaluations, monitoring data, press releases, legislation acts etc. are included. (author)

  13. Radiation accidents in hospitals

    Some of the radiation accidents that have occurred in Indian hospitals and causes that led to them are reviewed. Proper organization of radiation safety minimizes such accidents. It has been pointed out that there must be technical competence and mental preparedness to tackle emergencies when they do infrequently occur. (M.G.B.)

  14. Communication and industrial accidents

    As, Sicco van

    2001-01-01

    This paper deals with the influence of organizational communication on safety. Accidents are actually caused by individual mistakes. However the underlying causes of accidents are often organizational. As a link between these two levels - the organizational failures and mistakes - I suggest the conc

  15. Chernobyl accident and Denmark

    The report describes the Chernobyl accident and its consequences for Denmark in particular. It was commissioned by The Secretary of State for the Environment. Volume 2 contains copies of original documents issued by Danish authorities during the first accident phase and afterwards. Evaluations, monitoring data, press releases, legislation acts etc. are included. (author)

  16. Proceedings of the Second OECD/NEA Organisation Meeting on Increased Accident Tolerance of Fuels for LWRs

    Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the 2. Meeting on Increased Accident Tolerance of Fuels for LWRs. Content: 1 - Overview of the exchanges after the December-2012 Workshop through the discussion forum established at the OECD-NEA (S. Massara, NEA); 2 - Metrics Development for Enhanced Accident Tolerant LWR Fuels (S. Bragg-Sitton, INL); 3 - Candidate ATF Clad Technologies and Key Feasibility Issues (L. Snead, ORNL); 4 - CEA studies on nuclear fuel claddings for LWRs enhanced accident tolerant fuel. Some recent results, pending issues and prospects (J.C. Brachet, CEA); 5 - Current status on the accident tolerant fuel development in the Republic of Korea (J.Y. Park, J.H. Chang, KAERI); 6 - The current status of fuel R and D in the P.R. of China (T. Liu, CGN). Session 2: Key elements for a work programme on ATF: 7 - Beneficial characteristics of ATF (metrics) (L. Hallstadius, Westinghouse); 8 - Reactor types of interest (applicability) (L. Ott, ORNL); 9 - Impact on normal operations (N

  17. Accidents with orphan sources

    The International Atomic Energy Agency has specifically defined statutory functions relating to the development of standards of safety and the provision for their application. It also has responsibilities placed on it by virtue of a number of Conventions, two of which are relevant to nuclear accidents or radiological emergencies - the Convention on Early Notification of a Nuclear Accident and the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency. An overview of the way in which these functions are being applied to prevent and respond to radiological accidents, particularly those involving orphan sources, is described in this paper. Summaries of a number of such accidents and of the Agency's Action Plan relating to the safety and security of radiation sources are given. (orig.)

  18. Estimation of cost per severe accident for improvement of accident protection and consequence mitigation strategies

    To assess the complex situations regarding the severe accidents such as what observed in Fukushima Accident, not only radiation protection aspects but also relevant aspects: health, environmental, economic and societal aspects; must be all included into the consequence assessment. In this study, the authors introduce the “cost per severe accident” as an index to analyze the consequences of severe accidents comprehensively. The cost per severe accident consists of various costs and consequences converted into monetary values. For the purpose of improvement of the accident protection and consequence mitigation strategies, the costs needed to introduce the protective actions, and health and psychological consequences are included in the present study. The evaluations of these costs and consequences were made based on the systematic consequence analysis using level 2 and 3 probabilistic safety assessment (PSA) codes. The accident sequences used in this analysis were taken from the results of level 2 seismic PSA of a virtual 1,100 MWe BWR-5. The doses to the public and the number of people affected were calculated using the level 3 PSA code OSCAAR of Japan Atomic Energy Agency (JAEA). The calculations have been made for 248 meteorological sequences, and the outputs are given as expectation values for various meteorological conditions. Using these outputs, the cost per severe accident is calculated based on the open documents on the Fukushima Accident regarding the cost of protective actions and compensations for psychological harms. Finally, optimized accident protection and consequence mitigation strategies are recommended taking into account the various aspects comprehensively using the cost per severe accident. The authors must emphasize that the aim is not to estimate the accident cost itself but to extend the scope of “risk-informed decision making” for continuous safety improvements of nuclear energy. (author)

  19. Calculation of relative tube/tube support plate displacements in steam generators under accident condition loads using non-linear dynamic analysis methodologies

    A non-linear analysis has been performed to determine relative motions between tubes and tube support plates (TSP) during a steam line break (SLB) event for steam generators. The SLB event results in blowdown of steam and water out of the steam generator. The fluid blowdown generates pressure drops across the TSPS, resulting in out-of-plane motion. The SLB induced pressure loads are calculated with a computer program that uses a drift-flux modeling of the two-phase flow. In order to determine the relative tube/TSP motions, a nonlinear dynamic time-history analysis is performed using a structural model that considers all of the significant component members relative to the tube support system. The dynamic response of the structure to the pressure loads is calculated using a special purpose computer program. This program links the various substructures at common degrees of freedom into a combined mass and stiffness matrix. The program accounts for structural non-linearities, including potential tube and TSP interaction at any given tube position. The program also accounts for structural damping as part of the dynamic response. Incorporating all of the above effects, the equations of motion are solved to give TSP displacements at the reduced set of DOF. Using the displacement results from the dynamic analysis, plate stresses are then calculated using the detailed component models. Displacements form the dynamic analysis are imposed as boundary conditions at the DOF locations, and the finite element program then solves for the overall distorted geometry. Calculations are also performed to assure that assumptions regarding elastic response of the various structural members and support points are valid

  20. Behaviour of fission products under severe PWR accident conditions. The VERCORS experimental programme-Part 3: Release of low-volatile fission products and actinides

    The VERCORS analytical programme consisted of a series of tests carried out on irradiated PWR fuel samples. The tests - funded jointly by EDF and IRSN - were carried out by the Commissariat a l'Energie Atomique (CEA) at their Grenoble site. They were performed in a hot cell belonging to the Active Materials Analysis Laboratory (LAMA). The general outline of the programme was set out in a first article (of a series of 3), which described the different levels of fission products (FP) volatility and their characteristics. This led to a classification into five main categories of volatility and/or behaviour: (1) Volatile FP including fission gases, iodine, caesium, antimony, tellurium, cadmium, rubidium and silver; (2) Semi-volatile FP, a category made up of molybdenum, rhodium, barium, palladium and technetium; (3) Low-volatile FP comprising ruthenium, cerium, strontium, yttrium, europium, niobium and lanthanum with generally low but significant release; (4) Non-volatile FP including zirconium, neodymium and praseodymium; and lastly (5) Actinides which group together uranium, plutonium, neptunium, americium and curium. The specific behaviour of fission gases and volatile FP is dealt with in the second article, which also includes the specific characteristics of volatile FP regarding transport. The main variables (i.e. temperature, which is the main variable at least until loss of sample geometry, oxidising-reducing conditions, burn-up, interactions with the cladding and/or the structural components, the nature of the fuel, and finally the state of the fuel) affecting the kinetics and/or the released fraction of these same FP could also be identified. This final article represents the Third Part of the series. It concerns the release of actinides and less volatile FP, in keeping with the classification by categories previously identified, which are as follows: (1) semi-volatile FP, comprising of Mo, Ba, Rh, Pd, Tc, (2) low-volatile FP, comprising of Sr, Y, Nb, Ru, La