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. 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

  9. 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

  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. 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

  17. 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

  18. 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

  19. 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

  20. 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

  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. 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.)

  7. 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)

  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. 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)

  12. 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

  13. 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.

  14. 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)

  15. 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

  16. 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)

  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. 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

  3. 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 ...

  4. 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

  5. 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.

  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. 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)

  10. 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

  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. 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.)

  20. 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

  1. 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

  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. 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

  5. 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

  6. 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.)

  7. 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)

  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. 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

  13. 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

  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. 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)

  16. 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)

  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. 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)

  6. 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)

  7. 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)

  8. 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)

  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. 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

  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. 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…

  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. 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

  2. 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

  3. 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)

  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. 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.)

  9. 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

  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. 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)

  15. 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)

  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. 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)

  19. 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

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. 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.

  5. 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)

  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. 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)

  15. 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.)

  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

    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

  18. 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

  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 (≅ 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

  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 (≅ 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

  3. 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

  4. 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.

  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. 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.)

  11. 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

  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. 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

  20. 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)

  1. 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

  2. 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

  3. 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.)

  4. 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

  5. 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)

  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. 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

  8. 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)

  9. 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.)

  10. 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.)

  11. 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

  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. 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.

  18. 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)

  19. 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

  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. 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)

  6. 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

  7. 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)

  8. 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

  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. 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)

  12. 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

  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. 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

  18. 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

  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.)