WorldWideScience

Sample records for accident conditions key

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

    Energy Technology Data Exchange (ETDEWEB)

    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 risk indicators for accident assessment conditioned on pre-crash vehicle trajectory.

    Science.gov (United States)

    Shi, X; Wong, Y D; Li, M Z F; Chai, C

    2018-08-01

    Accident events are generally unexpected and occur rarely. Pre-accident risk assessment by surrogate indicators is an effective way to identify risk levels and thus boost accident prediction. Herein, the concept of Key Risk Indicator (KRI) is proposed, which assesses risk exposures using hybrid indicators. Seven metrics are shortlisted as the basic indicators in KRI, with evaluation in terms of risk behaviour, risk avoidance, and risk margin. A typical real-world chain-collision accident and its antecedent (pre-crash) road traffic movements are retrieved from surveillance video footage, and a grid remapping method is proposed for data extraction and coordinates transformation. To investigate the feasibility of each indicator in risk assessment, a temporal-spatial case-control is designed. By comparison, Time Integrated Time-to-collision (TIT) performs better in identifying pre-accident risk conditions; while Crash Potential Index (CPI) is helpful in further picking out the severest ones (the near-accident). Based on TIT and CPI, the expressions of KRIs are developed, which enable us to evaluate risk severity with three levels, as well as the likelihood. KRI-based risk assessment also reveals predictive insights about a potential accident, including at-risk vehicles, locations and time. Furthermore, straightforward thresholds are defined flexibly in KRIs, since the impact of different threshold values is found not to be very critical. For better validation, another independent real-world accident sample is examined, and the two results are in close agreement. Hierarchical indicators such as KRIs offer new insights about pre-accident risk exposures, which is helpful for accident assessment and prediction. Copyright © 2018 Elsevier Ltd. All rights reserved.

  3. Mechanistic prediction of fission product release under normal and accident conditions: key uncertainties that need better resolution

    International Nuclear Information System (INIS)

    Rest, J.

    1983-09-01

    A theoretical model has been used for predicting the behavior of fission gas and volatile fission products (VFPs) in UO 2 -base fuels during steady-state and transient conditions. This model represents an attempt to develop an efficient predictive capability for the full range of possible reactor operating conditions. Fission products released from the fuel are assumed to reach the fuel surface by successively diffusing (via atomic and gas-bubble mobility) from the grains to grain faces and then to the grain edges, where the fission products are released through a network of interconnected tunnels of fission-gas induced and fabricated porosity. The model provides for a multi-region calculation and uses only one size class to characterize a distribution of fission gas bubbles

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  5. The accidents during shutdown conditions Temelin NPP

    International Nuclear Information System (INIS)

    Sykora, M.; Mlady, O.

    1996-01-01

    Two parallel activities oriented for the accidents during shutdown conditions are performed at Temelin NPP: Development of symptom based emergency operating procedures (EOPs) applicable for the accidents which could occur during operational modes 1 through 4; independent evaluation of plant safety as part of the Temelin Shutdown probabilistic assessment to define the accidents which could occur during mode 5 and 6 for which the EOPs must be extended. Both these activities are in progress now because Temelin plant is still in the construction phase

  6. VICKEY: Mining Conditional Keys on Knowledge Bases

    DEFF Research Database (Denmark)

    Symeonidou, Danai; Prado, Luis Antonio Galarraga Del; Pernelle, Nathalie

    2017-01-01

    A conditional key is a key constraint that is valid in only a part of the data. In this paper, we show how such keys can be mined automatically on large knowledge bases (KBs). For this, we combine techniques from key mining with techniques from rule mining. We show that our method can scale to KBs...

  7. Evaluation of severe accident environmental conditions taking accident management strategy into account for equipment survivability assessments

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Jeong, Ji Hwan; Na, Man Gyun; Kim, Soong Pyung

    2003-01-01

    This paper presents a methodology utilizing accident management strategy in order to determine accident environmental conditions in equipment survivability assessments. In case that there is well-established accident management strategy for specific nuclear power plant, an application of this tool can provide a technical rationale on equipment survivability assessment so that plant-specific and time-dependent accident environmental conditions could be practically and realistically defined in accordance with the equipment and instrumentation required for accident management strategy or action appropriately taken. For this work, three different tools are introduced; Probabilistic Safety Assessment (PSA) outcomes, major accident management strategy actions, and Accident Environmental Stages (AESs). In order to quantitatively investigate an applicability of accident management strategy to equipment survivability, the accident simulation for a most likely scenario in Korean Standard Nuclear Power Plants (KSNPs) is performed with MAAP4 code. The Accident Management Guidance (AMG) actions such as the Reactor Control System (RCS) depressurization, water injection into the RCS, the containment pressure and temperature control, and hydrogen concentration control in containment are applied. The effects of these AMG actions on the accident environmental conditions are investigated by comparing with those from previous normal accident simulation, especially focused on equipment survivability assessment. As a result, the AMG-involved case shows the higher accident consequences along the accident environmental stages

  8. VICKEY: Mining Conditional Keys on Knowledge Bases

    OpenAIRE

    Symeonidou , Danai; Galárraga , Luis; Pernelle , Nathalie; Saïs , Fatiha; Suchanek , Fabian

    2017-01-01

    International audience; A conditional key is a key constraint that is valid in only a part of the data. In this paper, we show how such keys can be mined automatically on large knowledge bases (KBs). For this, we combine techniques from key mining with techniques from rule mining. We show that our method can scale to KBs of millions of facts. We also show that the conditional keys we mine can improve the quality of entity linking by up to 47 percentage points.

  9. Primary pump vibration under accident conditions

    International Nuclear Information System (INIS)

    Guthrie, B.M.; Currie, T.C.

    1984-06-01

    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. Network conditioning under conflicting goals: Accident causation

    International Nuclear Information System (INIS)

    Jouse, W.C.

    1992-01-01

    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

  11. Chemical phenomena under severe accident conditions

    International Nuclear Information System (INIS)

    Powers, D.A.

    1988-01-01

    A severe nuclear reactor accident is expected to involve a vast number of chemical processes. The chemical processes of major safety significance begin with the production of hydrogen during steam oxidation of fuel cladding. Physico-chemical changes in the fuel and the vaporization of radionuclides during reactor accidents have captured much of the attention of the safety community in recent years. Protracted chemical interactions of core debris with structural concrete mark the conclusion of dynamic events in a severe accident. An overview of the current understanding of chemical processes in severe reactor accident is provided in this paper. It is shown that most of this understanding has come from application of findings from other fields though a few areas have in the past been subject to in-depth study of a fundamental nature. Challenges in the study of severe accident chemistry are delineated

  12. Interaction of radionuclides in severe accident conditions

    International Nuclear Information System (INIS)

    Nagrale, Dhanesh B.; Bera, Subrata; Deo, Anuj Kumar; Paul, U.K.; Prasad, M.; Gaikwad, A.J.

    2015-01-01

    Nuclear power plants are designed with inherent engineering safety systems and associated operational procedures that provide an in-depth defence against accidents. Radionuclides such as Iodine, Cesium, Tellurium, Barium, Strontium, Rubidium, Molybdenum and many others may get released during a severe accident. Among these, Iodine, one of the fission products, behaviour is significant for the analysis of severe accident consequences because iodine is a chemically more active to the potential components released 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 is governed by aerosol physics, depletion mechanisms gravitational settling, diffusiophoresis and thermophoresis. 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 play significant roles. Water radiolysis products due to presence of dissolved impurities, chloride ions, organic impurities should be considered while calculating iodine release. Containment filtered venting system (CFVS) consists of venturi scrubber and a scrubber tank which is dosed with NaOH and NaS_2O_3 in water where iodine will react with the chemicals and convert into NaI and Na_2SO_4. This paper elaborates the issues with respect to interaction of radionuclides and its consideration in modeling of severe accident. (author)

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

    International Nuclear Information System (INIS)

    Guo, Z.

    1992-01-01

    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

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

    International Nuclear Information System (INIS)

    Brown, G.S.; Cashwell, J.W.; Apple, M.L.

    1993-01-01

    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.)

  15. ACCOUNT OF ROAD CONDITIONS WHILE INVESTIGATING TRAFFIC ACCIDENTS

    Directory of Open Access Journals (Sweden)

    D. D. Selioukov

    2010-01-01

    Full Text Available 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.

  16. Accidents and emergency conditions: Legal aspects

    International Nuclear Information System (INIS)

    Peinsipp, N.

    1985-01-01

    The currently valid versions of the Federal German Atomic Energy Act, the Radiation Protection Ordinance, and the X-Ray Ordinance show differences with regard to the use of the terms 'accident' or 'emergency', respectively, preferring either one or the other term, or using both terms. The author comments on this lack of harmonization in terminology and goes into details on aspects such as legislation and application of law. (DG) [de

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

    International Nuclear Information System (INIS)

    Nakoski, John

    2013-01-01

    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

  18. [Keys to preventing accidents in children in the school context].

    Science.gov (United States)

    Gabari Gambarte, M Inés; Sáenz Mendía, Raquel

    2016-11-02

    To learn about children's perception of the causes and prevention strategies involved in school accidents. The sample included 584 school children aged 8-9 years from Navarra. A mixed design was chosen by questionnaire with three open-response questions and one multiple-choice assessment. Analysis was performed in two phases: 1) qualitative development of categories and dimensions of the responses of narrative content, and 2) quantitative variables for recoding correlational analysis. 22 categories emerged, which make up three perceptual dimensions: 1) attribution of causality (5), 2) identification of mechanisms of avoidance (11), and 3) development of coping strategies (6). The correlation intra-variables portray varying degrees: on the one hand, moderate positive numbers (r>0.5) in allocating and identifying causality avoidance mechanisms and, on the other hand, high positive correlation values (r>0.7) referred to developing coping strategies. Children are able to identify accidents as a health problem. They question the multiplicity of elements involved and relate the origin and kind of accident to prevention and support mechanisms. Copyright © 2016 SESPAS. Publicado por Elsevier España, S.L.U. All rights reserved.

  19. Graphite Oxidation Simulation in HTR Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed

    2012-10-19

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

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

    International Nuclear Information System (INIS)

    Lombardo, N.J.; Lanning, D.D.; Panisko, F.E.

    1992-12-01

    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

  1. NPP Krsko containment environmental conditions during postulated accident

    International Nuclear Information System (INIS)

    Kozaric, M.; Cavlina, N.; Spalj, S.

    1989-01-01

    This paper presents NPP Krsko containment pressure and temperature increase during Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB). Containment environmental condition calculation was performed by CONTEMPT4/MOD4 computer code. Design accident calculations were performed by RELAP4/MOD6 and RELAP5/MOD1 computer codes. Calculational abilities and application methodology of these codes are presented. The CONTEMPT code is described in more detail. The containment pressure and temperature time distribution are presented as well. (author)

  2. Analysis of the behaviour of the Kozloduy NPP Unit 3 under severe accident conditions

    International Nuclear Information System (INIS)

    Velev, V.; Saraeva, V.

    2004-01-01

    The objective of the analysis is to study the behaviour of the Kozloduy NPP Unit 3 under severe accident conditions. The analysis is performed using computer code MELCOR 1.8.4. This report includes a brief description of Unit 3 active core as well as description and comparison of the key events

  3. Behaviour of molten reactor fuels under accident conditions

    International Nuclear Information System (INIS)

    Xavier Swamikannu, A.; Mathews, C.K.

    1980-01-01

    The behaviour of molten reactor fuels under accident conditions has received considerable importance in recent times. The chemical processes that occur in the molten state among the fuel, the clad components and the concrete of the containment building under the conditions of a core melt down accident in oxide fuelled reactors have been reviewed with the purpose of identifying areas of developmental work required to be performed to assess and minimize the consequences of such an accident. This includes the computation and estimation of vapour pressure of various gaseous species over the fuel, the clad and the coolant, providing of sacrificial materials in the concrete in order to protect the containment building in order to prevent release of radioactive gases into the atmosphere and understanding the distribution and chemical state of fission products in the molten fuel in order to provide for the effective removal of their decay heats. (auth.)

  4. Response of HEPA filters to simulated-accident conditions

    International Nuclear Information System (INIS)

    Gregory, W.S.; Martin, R.A.; Smith, P.R.; Fenton, D.E.

    1982-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

    Majumdar, S.

    1996-09-01

    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 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

  7. Hydrogen formation and control under postulated LMFBR accident conditions

    International Nuclear Information System (INIS)

    Armstrong, G.R.; Wierman, R.W.

    1976-09-01

    The objective of this study is to experimentally investigate the potential for autoignition and combustion of hydrogen-sodium mixtures which may be produced in LMFBR accidents. The purpose and ultimate usefulness of this work is to provide data that will establish the validity and acceptability of mechanisms inherent to the LMFBR that could either prevent or delay the accumulation of hydrogen gas to less than 4 percent (V) in the Reactor Containment Building (RCB) under accident conditions. The results to date indicate that sodium and sodium-hydrogen mixtures such as may be expected during LMFBR postulated accidents will ignite upon entering an air atmosphere and that the hydrogen present will be essentially all consumed until such time that the oxygen concentration is depleted

  8. Researches of WWER fuel rods behaviour under RIA accident conditions

    International Nuclear Information System (INIS)

    Nechaeva, O.; Medvedev, A.; Novikov, V.; Salatov, A.

    2003-01-01

    Unirradiated fuel rod and refabricated fuel rod tests in the BIGR as well as acceptance criteria proving absence of fragmentation and the settlement modeling of refabricated fuel rods thermomechanical behavior in the BIGR-tests using RAPTA-5 code are discussed in this paper. The behaviour of WWER type simulators with E110 and E635 cladding was researched at the BIGR reactor under power pulse conditions simulating reactivity initiated accident. The results of the tests in four variants of experimental conditions are submitted. The behaviour of 12 WWER type refabricated fuel rods was researched in the BIGR reactor under power pulse conditions simulating reactivity initiated accident: burnup 48 and 60 MWd/kgU, pulse width 3 ms, peak fuel enthalpy 115-190 cal/g. The program of future tests in the research reactor MIR with high burnup fuel rod (up to 70 MWd/kgU) under conditions simulating design RIA in WWER-1000 is presented

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

    International Nuclear Information System (INIS)

    Hilliard, R.K.; Postma, A.K.; Jeppson, D.W.

    1983-03-01

    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

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

    International Nuclear Information System (INIS)

    Giordano, P.; Micaelli, J.C.; Haste, T.; Herranz, L.

    2005-01-01

    About fifty European organisations integrate in SARNET (Network of Excellence of the EU 6 th 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

  11. Status of USNRC research on fuel behavior under accident conditions

    International Nuclear Information System (INIS)

    Johnston, W.V.

    1976-01-01

    The program of the Fuel Behaviour Research is directed at providing a detailed understanding of the response of nuclear fuel assemblies to off-normal or accident conditions. This understanding is expressed in physical and analytical correlations which are incorporated into computer codes. The results of these experiments and the resulting codes are available to the licensing authorities for use in evaluating utility submissions. (orig.) [de

  12. Hydrogen generation, distribution and combustion under severe LWR accident conditions: a state-of-technology report

    International Nuclear Information System (INIS)

    Postma, A.K.; Hilliard, R.K.

    1983-03-01

    This report reviews the current state of technology regarding hydrogen safety issues in light water reactor plants. Topics considered in this report include hydrogen generation, distribution in containment, and combustion characteristics. A companion report addresses hydrogen control. 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

  13. Radiological protection and the Fukushima Daiichi accident. Responses of the key international organisations

    Energy Technology Data Exchange (ETDEWEB)

    Clement, Christopher

    2017-10-01

    The Fukushima Daiichi nuclear power plant accident in March 2011 shook the radiological protection world. All major organisations in the radiological protection field turned their eyes to Japan. Their actions, driven by their mandates, are reflected in their respective landmark reports on the accident. Reports of the International Commission on Radiological Protection, World Health Organisation, United Nations Scientific Committee on the Effects of Atomic Radiation, and International Atomic Energy Agency are summarised. Collaboration between key international organisations is strong, based in part on informal interactions which need to be backed up with formal relations to ensure solid long-term collaboration.

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

    International Nuclear Information System (INIS)

    Ruedinger, V.; Wilhelm, J.G.

    1985-03-01

    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.) [de

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

    International Nuclear Information System (INIS)

    Hill, Axel; Stiepani, Cristoph; Drechsler, Michael

    2015-01-01

    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)

  16. Investigation of air cleaning system response to accident conditions

    International Nuclear Information System (INIS)

    Andrae, R.W.; Bolstad, J.W.; Foster, R.D.; Gregory, W.S.; Horak, H.L.; Idar, E.S.; Martin, R.A.; Ricketts, C.I.; Smith, P.R.; Tang, P.K.

    1980-01-01

    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

  17. Behavior of LWR fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Albrecht, H.; Bocek, M.; Erbacher, F.; Fiege, A.; Fischer, M.; Hagen, S.; Hofmann, P.; Holleck, H.; Karb, E.; Leistikow, S.; Melang, S.; Ondracek, G.; Thuemmler, F.; Wiehr, K.

    1977-01-01

    In the frame of the German reactor safety research program, the Kernforschungszentrum Karlsruhe is carrying out a comprehensive program on the behavior of LWR fuel elements under a variety of power cooling mismatch conditions in particular during loss-of-coolant accidents. The major objectives are to establish a detailed quantitative understanding of fuel rod failures mechanisms and their thresholds, to evaluate the safety margins of power reactor cores under accident conditions and to investigate the feedback of fuel rod failures on the efficiency of emergency core cooling systems. This detailed quantitative understanding is achieved through extensive basic and integral experiments and is incorporated in a fuel behavior code. On the basis of these results the design of power reactor fuel elements and of safety devices can be further improved. The results of investigations on the inelastic deformation (ballooning) behavior of Zircaloy 4 cladding at LOCA temperatures in oxidizing atmosphere are presented. Depending upon strain rate and temperature superplastic deformation behavior was observed. In the equation of state of Zry 4 the strain rate sensitivity index depends strongly upon strain and in the superplastic region upon sample anisotropy. Oxidation kinetics experiments with Zry-tubes at 900-1300 0 C showed that the Baker-Just correlation describes the reality quite conservative. Therefore a reduction of the amount of Zry oxidation can be assumed in the course of a LOCA. The external oxidation of Zry-cladding by steam as well as internal oxidation by the oxygen in oxide fuel and fission products (Cs, I, Te) have an influence on the strain and rupture behavior of Zry-cladding at LOCA temperatures. In out-of-pile and inpile experiments the mechanical and thermal behavior of fuel rods during the blowdown, the heatup and the reflood phases of a LOCA are investigated under representative and controlled thermohydraulic conditions. The task of the inpile experiments is

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

    International Nuclear Information System (INIS)

    Hill, Axel

    2014-01-01

    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

  19. Numerical Study of Severe Accidents on Containment Venting Conditions

    Energy Technology Data Exchange (ETDEWEB)

    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

  20. Numerical Study of Severe Accidents on Containment Venting Conditions

    International Nuclear Information System (INIS)

    Lee, Na Rae; Bang, Young Suk; Park, Tong Kyu; Lee, Doo Yong; Choi, Yu Jung; Lee, Sang Won; Kim, Hyeong Taek

    2014-01-01

    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

  1. Behaviour of gas cooled reactor fuel under accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    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

  2. Most likely failure location during severe accident conditions

    International Nuclear Information System (INIS)

    Rempe, J.L.; Allison, C.M.

    1991-01-01

    This paper describes preliminary results from which finite element calculation results are used in conjunction with analytical calculation results to predict failure in different LWR vessel designs during a severe accident. Detailed analyses are being performed to investigate the relative likelihood of a BWR vessel and drain line penetration to fail during a wide range of severe accident conditions. Analytically developed failure maps, which were developed in terms of dimensionless groups, are applied to consider geometries and materials occurring in other LWR vessel designs. Preliminary numerical analysis results indicate that if ceramic debris relocates within the BWR drain line to a distance below the lower head, the drain line will reach failure temperatures before the vessel fails. Application of failure maps for these debris conditions to other LWR geometries indicate that in-vessel tube melting will occur in either BWR or PWR vessel designs. Furthermore, if this melt is assumed to fill the entire penetration flow area, the melt is predicted to travel well below the lower head in any of the reference LWR penetrations. However, failure maps suggest the result that ex-vessel tube temperatures exceed the penetration's ultimate strength is specific to the BWR drain line because of its material composition and relatively large effective diameter for melt flow

  3. Structural Evaluation on HIC Transport Packaging under Accident Conditions

    International Nuclear Information System (INIS)

    Chung, Sung Hwan; Kim, Duck Hoi; Jung, Jin Se; Yang, Ke Hyung; Lee, Heung Young

    2005-01-01

    HIC transport packaging to transport a high integrity container(HIC) containing dry spent resin generated from nuclear power plants is to comply with the regulatory requirements of Korea and IAEA for Type B packaging due to the high radioactivity of the content, and to maintain the structural integrity under normal and accident conditions. It must withstand 9 m free drop impact onto an unyielding surface and 1 m drop impact onto a mild steel bar in a position causing maximum damage. For the conceptual design of a cylindrical HIC transport package, three dimensional dynamic structural analysis to ensure that the integrity of the package is maintained under all credible loads for 9 m free drop and 1 m puncture conditions were carried out using ABAQUS code.

  4. Behaviour of organic iodides under pwr accident conditions

    International Nuclear Information System (INIS)

    Alm, M.

    1982-01-01

    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 CH 3 sup(127/131)I-containing steam-air mixture within the expansion vessel was analysed at 80 0 C 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)

  5. Failure of fretted steam generator tubes under accident conditions

    International Nuclear Information System (INIS)

    Forrest, C.F.

    1996-10-01

    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

  6. Computation of reactor control rod drop time under accident conditions

    International Nuclear Information System (INIS)

    Dou Yikang; Yao Weida; Yang Renan; Jiang Nanyan

    1998-01-01

    The computational method of reactor control rod drop time under accident conditions lies mainly in establishing forced vibration equations for the components under action of outside forces on control rod driven line and motion equation for the control rod moving in vertical direction. The above two kinds of equations are connected by considering the impact effects between control rod and its outside components. Finite difference method is adopted to make discretization of the vibration equations and Wilson-θ method is applied to deal with the time history problem. The non-linearity caused by impact is iteratively treated with modified Newton method. Some experimental results are used to validate the validity and reliability of the computational method. Theoretical and experimental testing problems show that the computer program based on the computational method is applicable and reliable. The program can act as an effective tool of design by analysis and safety analysis for the relevant components

  7. Computer code calculations of the TMI-2 accident: initial and boundary conditions

    International Nuclear Information System (INIS)

    Behling, S.R.

    1985-05-01

    Initial and boundary conditions during the Three Mile Island Unit 2 (TMI-2) accident are described and detailed. A brief description of the TMI-2 plant configuration is given. Important contributions to the progression of the accident in the reactor coolant system are discussed. Sufficient information is provided to allow calculation of the TMI-2 accident with computer codes

  8. Ruthenium transport experiments in air ingress accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Teemu, Karkele; Ulrika, Backman; Ari, Auvinen; Unto, Tapper; Jorma, Jokiniemi [VTT Technical Research Centre of Finland, Fine Particles (Finland); Riitta, Zilliacus; Maija, Lipponen; Tommi, Kekki [VTT Technical Research Centre of Finland, Accident Management (Finland); Jorma, Jokiniemi [Kuopio Univ., Dept. of Environmental Sciences, Fine Particle and Aerosol Technology Lab. (Finland)

    2007-07-01

    In this study the release, transport and speciation of ruthenium in conditions simulating an air ingress accident was studied. Ruthenium dioxide was exposed to oxidising environment at high temperature (1100-1700 K) in a tubular flow furnace. At these conditions volatile ruthenium species were formed. A large fraction of the released ruthenium was deposited in the tube as RuO{sub 2}. Depending on the experimental conditions 1-26 wt% of the released ruthenium was trapped in the outlet filter as RuO{sub 2} particles. In stainless steel tube 0-8.8 wt% of the released ruthenium reached the trapping bottle as gaseous RuO{sub 4}. A few experiments were carried out, in which revaporization of ruthenium deposited on the tube walls was studied. In these experiments, oxidation of RuO{sub 2} took place at a lower temperature. During revaporization experiments 35-65 % of ruthenium was transported as gaseous RuO{sub 4}. In order to close mass balance and achieve better time resolution 4 experiments were carried out using a radioactive tracer. In these experiments ruthenium profiles were measured. These experiments showed that the most important retention mechanism was decomposition of gaseous RuO{sub 3} into RuO{sub 2} as the temperature of the furnace was decreasing. In these experiments the transport rate of gaseous ruthenium was decreasing while the release rate was constant.

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

    International Nuclear Information System (INIS)

    Ang, M.L.; Grindon, E.; Dutton, L.M.C.; Garcia-Sedano, P.; Santamaria, C.S.; Centner, B.; Auglaire, M.; Routamo, T.; Outa, S.; Jokiniemi, J.; Gustavsson, V.; Wennerstrom, H.; Spanier, L.; Gren, M.; Boschiero, M-H; Droulas, J-L; Friederichs, H-G; Sonnenkalb, M.

    2001-01-01

    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

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

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

    2015-01-01

    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 UO 2 –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

  11. Reliability analysis of emergency decay heat removal system of nuclear ship under various accident conditions

    International Nuclear Information System (INIS)

    Matsuoka, Takeshi

    1984-01-01

    A reliability analysis is given for the emergency decay heat removal system of the Nuclear Ship ''Mutsu'' and the emergency sea water cooling system of the Nuclear Ship ''Savannah'', under ten typical nuclear ship accident conditions. Basic event probabilities under these accident conditions are estimated from literature survey. These systems of Mutsu and Savannah have almost the same reliability under the normal condition. The dispersive arrangement of a system is useful to prevent the reduction of the system reliability under the condition of an accident restricted in one room. As for the reliability of these two systems under various accident conditions, it is seen that the configuration and the environmental condition of a system are two main factors which determine the reliability of the system. Furthermore, it was found that, for the evaluation of the effectiveness of safety system of a nuclear ship, it is necessary to evaluate its reliability under various accident conditions. (author)

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

    International Nuclear Information System (INIS)

    Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

    1986-09-01

    The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments

  13. OCCUPATIONAL ACCIDENTS AS INDICATORS OF INADEQUATE WORK CONDITIONS AND WORK ENVIRONMENT

    OpenAIRE

    Petar Babović

    2009-01-01

    Occupational accidents due to inadequate working conditions and work environment present a major problem in highly industrialised countries, as well as in developing ones. Occupational accidents are a regular and accompanying phenomenon in all human activities and one of the main health related and economic problems in modern societies.The aim of this study is the analysis of the connections of unfavourable working conditions and working environment on occupational accidents. Occurrence of oc...

  14. Core dynamics of HTR under ATWS and accident conditions

    International Nuclear Information System (INIS)

    Nabbi, R.

    1988-05-01

    The systematic classification of the ATWS has been undertaken by analogy to the considerations made for LWR. The initiating events of ATWS and protection actions of safety systems resulting from monitoring of the system variables have been described. The main emphasis of this work is the analysis of the core dynamic consequences of scram failure during the anticipated transients. The investigation has shown that because of the temperature feedback mechanisms a temperature rise during the ATWS results in a self-shutdown of the reactor. Further inherent safety features of the HTR - conditioned by the high heat capacity of the core and by the compressibility of the coolant - do effectively counteract an undesirable increase of temperature and pressure in the primary circuit. In case of the long-term failure of the forced cooling and following core heatup, neutron physical phenomena appear which determine the reactivity behaviour of the HTR. They are, for instance, the decay of Xenon 135, release of the fission products and subsiding of the top reflector. The results of the computer simulations show that a recriticality has to be excluded during the first 2 days if the reactor is shutdown by the reflector rods at the beginning of the accident. (orig./HP) [de

  15. Fuel behavior and fission product release under HTGR accident conditions

    International Nuclear Information System (INIS)

    Fukuda, K.; Hayashi, K.; Shiba, K.

    1990-01-01

    In early 1989 a final decision was made over construction of a 30 MWth HTGR called the High Temperature Engineering Test Reactor, HTTR, in Japan in order to utilize it for high temperature gas engineering tests and various nuclear material tests. The HTTR fuel is a pin-in-block type fuel element which is composed of a hexagonal graphite block with dimension of 580 mm in length and 360 mm in face-to-face distance and about 30 of the fuel rods inserted into the coolant channels drilled in the block. The TRISO coated fuel particles for HTTR are incorporated with graphite powder and phenol resin into the fuel compacts, 19 of which are encased into a graphite sleeve as a fuel rod. It is necessary for the HTTR licensing to prove the fuel stability under predicted accidents related to the high temperature events. Therefore, the release of the fission products and the fuel failure have been investigated in the irradiation---and the heating experiments simulating these conditions at JAERI. This report describes the HTTR fuel behavior at extreme temperature, made clear in these experiments

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

    International Nuclear Information System (INIS)

    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

    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

  17. Iodine/steel reactions under severe accident conditions in LWR's

    International Nuclear Information System (INIS)

    Funke, F.; Greger, G-U.; Hellman, S.; Bleier, A.; Morell, W.

    1994-01-01

    Due to large surface areas, the reaction of volatile, molecular iodine (I 2 ) with steel surfaces in the containment may play an important role in predicting the source term to the environment. Both wall retention of iodine and conversion of volatile into non-volatile iodine compounds at steel surfaces have to be considered. Two types of laboratory experiments were carried out at Siemens/KWU in order to investigate the reaction of I 2 at steel surfaces representative for German power plants. 1) For steel coupons submerged in an I 2 solution at T = 50 deg C, 90 deg C or 140 deg C the reaction rate of the I 2 /I - conversion was determined. No iodine loading was observed on the steel in the aqueous phase tests. I 2 reacts with the steel components (Fe, Cr or Ni) to form metal iodides on the surface which are all immediately dissolved in water under dissociation into the metal and the iodide ions. From these experiments, the I 2 /I - conversion rate constants over the temperature range 50 deg C - 140 deg C as well as the activation energy were determined. The measured data are suitable to be included in severe accident iodine codes such as IMPAIR. 2) Steel tubes were exposed to a steam/I 2 flow under dry air at T=120 deg C and steam-condensing conditions at T= 120 deg C and 160 deg C. In dry air I 2 was retained on the steel surface and a deposition rate constant was measured. Under steam-condensing conditions there is an effective conversion of volatile I 2 to non-volatile I - which is subsequently washed off from the steel surface. The I 2 /I - conversion rate constants suitable for modelling this process were determined. No temperature dependency was found in the range 120 deg C - 160 deg C. (author). 4 refs., 2 tabs., 7 figs

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

    International Nuclear Information System (INIS)

    Nifenecker, H.

    2011-01-01

    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

  19. Assessment of WWER fuel condition in design basis accident

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.; Sokolov, N.; Andreeva-Andrievskaya, L.; Vlasov, Yu.; Nechaeva, O.; Salatov, A.

    1994-01-01

    The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR's. 12 figs., 11 refs

  20. Assessment of WWER fuel condition in design basis accident

    Energy Technology Data Exchange (ETDEWEB)

    Bibilashvili, Yu; Sokolov, N; Andreeva-Andrievskaya, L; Vlasov, Yu; Nechaeva, O; Salatov, A [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

    1994-12-31

    The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR`s. 12 figs., 11 refs.

  1. 10 CFR 71.74 - Accident conditions for air transport of plutonium.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Accident conditions for air transport of plutonium. 71.74 Section 71.74 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL Package, Special Form, and LSA-III Tests 2 § 71.74 Accident conditions for air transport of...

  2. Chemistry of fission product iodine under nuclear reactor accident conditions

    International Nuclear Information System (INIS)

    Malinauskas, A.P.; Bell, J.T.

    1986-01-01

    The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs

  3. Robot dispatching Scenario for Accident Condition Monitoring of NPP

    International Nuclear Information System (INIS)

    Kim, Jongseog

    2013-01-01

    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

  4. Study of containment air cooler capacity in steam air environment during accident conditions

    International Nuclear Information System (INIS)

    Kansal, M.; Mohan, N.; Bhawal, R.N.; Bajaj, S.S.

    2002-01-01

    Full text: The air coolers are provided for controlling the temperature in the reactor building during normal operation. These air coolers also serve as the main heat sink for the removal of energy from high enthalpy air-steam mixture expected in reactor building under accident conditions. A subroutine COOLER has been developed to estimate the heat removal rate of the air coolers at high temperature and steam conditions. The subroutine COOLER has been attached with the code PACSR (post accident containment system response) used for containment pressure temperature calculation. The subroutine was validated using design parameters at normal operating condition. A study was done to estimate the heat removal rate for some postulated accident conditions. The study reveals that, under accident conditions, the heat removal rate of air coolers increases several times compared with normal operating conditions

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

    International Nuclear Information System (INIS)

    1993-01-01

    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

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

    International Nuclear Information System (INIS)

    Bolt, P.R.

    1984-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-15

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

  8. Response of a DSNP pressurizer model under accident conditions

    International Nuclear Information System (INIS)

    Saphier, D.; Kallfelz, J.; Belblidia, L.

    1986-01-01

    Recently a new pressurizer model was developed for the DSNP simulation language. The model was connected to a simulation of the Trojan pressurized water reactor (PWR) and tested by simulating a loss-of-off-site power (LOSP) anticipated transient without scram. The results compare well to a similar study performed using the RELAP code. The pressurizer model and its response to the LOSP accident are presented

  9. Numerical module for debris behavior under severe accident conditions

    International Nuclear Information System (INIS)

    Kisselev, A.E.; Kobelev, G.V.; Strizhov, V.F.; Vasiliev, A.D.

    2005-01-01

    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)

  10. Probabilistic Approach to Conditional Probability of Release of Hazardous Materials from Railroad Tank Cars during Accidents

    Science.gov (United States)

    2009-10-13

    This paper describes a probabilistic approach to estimate the conditional probability of release of hazardous materials from railroad tank cars during train accidents. Monte Carlo methods are used in developing a probabilistic model to simulate head ...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

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

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

    International Nuclear Information System (INIS)

    Rosa, J.C. de la; Fornós, Joan

    2016-01-01

    Highlights: • Analysis of flammability conditions in buildings outside containment. • Stepwise approach easily applicable for any kind of containment and attached buildings layout. • Detailed application for real plant conditions has been included. - Abstract: Right after the events unfolded in Fukushima Daiichi, the European Union countries agreed in subjecting Nuclear Power Plants to Stress Tests as developed by WENRA and ENSREG organizations. One of the results as implemented in many European countries derived from such tests consisted of mandatory technical instructions issued by nuclear regulatory bodies on the analysis of potential risk of flammable gases in attached buildings to containment. The current study addresses the key aspects of the analysis of flammable gases leaking to auxiliary buildings attached to Westinghouse large-dry PWR containment for the specific situation where mitigating systems to prevent flammable gases to grow up inside containment are available, and containment integrity is preserved – hence avoiding isolation system failure. It also provides a full practical exercise where lessons learned derived from the current study – hence limited to the imposed boundary conditions – are applied. The leakage of gas from the containment to the support buildings is based on separate calculations using the EPRI-owned Modular Accident Analysis Program, MAAP4.07. The FATE™ code (facility Flow, Aerosol, Thermal, and Explosion) was used to model the transport and distribution of leaked flammable gas (H_2 and CO) in the penetration buildings. FATE models the significant mixing (dilution) which occurs as the released buoyant gas rises and entrains air. Also, FATE accounts for the condensation of steam on room surfaces, an effect which acts to concentrate flammable gas. The results of the analysis show that during a severe accident, flammable conditions are unlikely to occur in compartmentalized buildings such as the one used in the

  13. Assessment of Equipment Capability to Perform Reliably under Severe Accident Conditions

    International Nuclear Information System (INIS)

    2017-07-01

    The experience from the last 40 years has shown that severe accidents can subject electrical and instrumentation and control (I&C) equipment to environmental conditions exceeding the equipment’s original design basis assumptions. Severe accident conditions can then cause rapid degradation or damage to various degrees up to complete failure of such equipment. This publication provides the technical basis to consider when assessing the capability of electrical and I&C equipment to perform reliably during a severe accident. It provides examples of calculation tools to determine the environmental parameters as well as examples and methods that Member States can apply to assess equipment reliability.

  14. Accidents and emergency conditions: Tasks of the radiation protection expert

    International Nuclear Information System (INIS)

    Hacke, J.

    1985-01-01

    This paper reviews and explains the tasks of the radiation protection expert at a given site in the event of accidents or emergencies involving a radiation hazard to the personnel. The various measures recommended discriminate between the main two types of hazards, namely external radiation or internal radiation. The paper discusses the first-aid and emergency measures recommended in various publications (BG, 1982; ICRP, 1980; MO, 1972; ME, 1980) and also cites recommendations contained therein, referring to preventive means and measures and to communications to the press and the general public. (DG) [de

  15. Detection, diagnosis, and treatment of accident conditions using response trees

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1980-01-01

    Response Trees were developed at the LOFT facility in 1978 and included in the Plant Operating Manual (POM) to assist reactor operators in selecting emergency procedures. In an emergency situation the operator would manually gather data and evaluate the trees to select the appropriate procedures. As a portion of the LOFT Augmented Operator Capability (AOC) Program, the response tree methodology has been extended so that a computer can be used to evaluate the trees and recommend an appropriate response for an accident. Techniques for diagnosing failures within a cooling mode have also been investigated. This paper summarizes these additions to the response tree methodology

  16. Effect of RCIC Operating Conditions on the Accident Scenario in Fukushima Unit 2

    International Nuclear Information System (INIS)

    Kim, Sung Il; Park, Jong Hwa; Ha, Kwang Soon

    2015-01-01

    This study was conducted by using MELCOR 1.8.6. Fukushima unit 2 accident was analyzed using MELCOR in this study, and best estimate scenario with considering RCIC operating conditions was presented. Researches on the boiling water reactor (BWR) plant with reactor core isolation cooling (RCIC) system have been conducted. Research on the RCIC operation in Fukushima unit 2 was also conducted by Sandia National Laboratory. MELCOR analysis of the Fukushima unit 2 accident was conducted in the report and energy balance in wetwell was described by considering RCIC operation. However, the effect of RCIC operation condition on the accident scenario has not been studied. The operating conditions of RCIC system affect the pressures in wetwell and drywell, and the high pressure can make leakage path of fission product from PCV to reactor building. Thus it can be directly related with the amount of fission product which released to environment. In this study, severe accident on Fukushima unit 2 was analyzed considering the operating condition of RCIC system, and best estimated scenario was presented. In addition, the effect of RCIC turbine efficiency on the accident progression was examined. Energy balance in suppression chamber was also considered with discussion on the effect of torus room flooding level. It was found that the operating condition of RCIC turbine not only affects the variation of drywell pressure but also the amount of released fission products to environment. It was also confirmed that the RCIC turbine efficiency in the accident would be less than normal operating condition

  17. Tools to support important technical decisions during accident conditions

    International Nuclear Information System (INIS)

    Tenschert, J.; Bergiers, C.

    2008-01-01

    To handle design basis and beyond design basis accidents with intact reactor core, Nuclear Power Plants are using Emergency Operating Procedures (EOP) that they may have developed based on the generic Westinghouse Emergency Response Guidelines. Even though the EOPs are very directive, some questions are left to external support, i.e. to a team of persons constituting the so-called Technical Support Center (TSC). The Pressurized Water Reactor Owner Group (PWROG, previously Westinghouse Owner Group, WOG) has developed a TSC manual to support this group in their decision making process. Because of the specific and particular design of the Beznau NPP (KKB) Safety Systems, development of a plant-specific TSC manual required a lot of additions compared to the generic material. This plant-specific TSC manual is a helpful tool for the Site Emergency Director (SED) of the KKB to better evaluate issues and potential concerns arising while executing the EOPs. The majority of considered issues are relevant for beyond design basis accidents and external events. (orig.)

  18. A review of iodine chemistry under severe accident conditions

    International Nuclear Information System (INIS)

    Clough, P.M.; Starkie, H.C.; Wren, D.J.; Paquette, J.; Wren, J.C.

    1987-01-01

    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

  19. Iodine behavior in containment under LWR accident conditions

    International Nuclear Information System (INIS)

    Wisbey, S.J.; Beahm, E.C.; Shockley, W.E.; Wang, Y.M.

    1986-01-01

    The description of containment iodine behavior in reactor accident sequences requires an understanding of iodine volatility effects, deposition and revaporization/resuspension (from surfaces and aerosols), chemical changes between species, and mass transport. The experimental work in this program has largely centered on the interactions of iodine in or with water pools. The formation of volatile iodine, as I 2 or organic iodides, is primarily dependent on radiation and solution pH. Lower pH results in increased formation of volatile iodine species; thus, for example, a pH of 3.05 resulted in a conversion of I - to I 2 that was more than two orders of magnitude greater than tests run at pH 6.1 or 6.8. The formation or organic iodides involving water pools has been linked to the presence of iodine as I 2 , the solution/gas contact, and to the type of organic material

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

    Science.gov (United States)

    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 (p<0.05). These findings can be applied in the design and implementation of a comprehensive safety and health risk management system to reduce ASR.

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

    Energy Technology Data Exchange (ETDEWEB)

    Mishima, J.; Parkhurst, M.A.; Scherpelz, R.I.

    1985-03-01

    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.

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

    International Nuclear Information System (INIS)

    1996-12-01

    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

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

    International Nuclear Information System (INIS)

    Krieg, R.

    2005-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Schenk, W; Naoumidis, A [Institute for Reactor Material, KFA Juelich (Germany)

    1985-07-01

    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 UO{sub 2}-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.

  5. Ruthenium behaviour in severe nuclear accident conditions - Progress report

    International Nuclear Information System (INIS)

    Backman, U.; Lipponen, M.; Zilliacus, R.; Auvinen, A.; Jokiniemi, J.

    2004-03-01

    In order to prevent the radioactive ruthenium from spreading in gaseous form in case of an accident in a nuclear power plant it is of interest to know how it is formed and how it behaves. In the experiments the behaviour of ruthenium in oxidising atmosphere at high temperatures is studied. The methods for trapping and analysing RuO4 has been studied. It was found that 1M NaOH is capable of trapping RuO4 totally. The determination of Ru from the solution can be made using ICP-MS (inductively coupled plasma mass spectrometry) and from the reduced precipitates on filters by INAA (instrumental neutron activation analysis). The results of the experiments carried out so far is reported. A significant difference in the decomposition rate of gaseous RuO4 depending on the tube material was found. In all experiments only a minor fraction of Ru remained in gaseous form until the bubbler. In order to achieve a better mass balance an experiment using radioactive tracer was carried out. In the decomposition of gaseous Ru needle-shaped RuO2 crystallites were formed. (au)

  6. Stable Chemical Dosimeters for Partial Reconstruction of Nuclear Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Dvornik, I.; Zec, U.; Baric, M.; Razem, D. [Ruder Boskovic Nuclear Institute, Zagreb, Yugoslavia (Croatia)

    1969-10-15

    The application of chemical dosimeters, tissue equivalent with respect to gamma rays and neutrons, is proposed for dosimetric topography of the space around nuclear devices in case of accidents. The dosimeters in the form of sealed glass ampoules have sufficient sensitivity and long-term stability and are evaluated or checked directly by conventional spectrophotometry. The sensitivity, expressed as yield per rad, is approximately equal for gamma rays and neutrons. The resolution in both cases is about one rad, and the range is up to several thousand rads. The precision of dosimetry is {+-} 1 rad or {+-} 2%, whichever is higher. In free space and unshielded the dosimeter measures the total rad-absorbed dose delivered by gamma rays and neutrons, i.e. the first collision gamma plus neutron dose. If used on- or in-phantom, especially if several dosimeters are disposed within and around the same phantom, it can give important data about the amount of the neutron component of the dose and about the effective mean energy of incident neutrons. The neutron component of the dose can be directly measured if the gamma dosimeter is used together with the chemical dosimeter. The experiments giving the change of optical density per rad and the radiation chemical yield with respect to the absorbed dose delivered by 14-MeV neutrons are described in detail. The possibility is also mentioned of applying the dosimeter as a very sensitive monitor for thermal neutrons, which is due to the chlorine content of 4.73% and activation to {sup 38}Cl. The opinion is expressed that this dosimeter deserves some attention as a part of future planning and development work on area and personnel accidental dosimetry systems. (author)

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

    International Nuclear Information System (INIS)

    Kunitomi, K.; Shinozaki, M.; Baba, O.; Saito, S.

    1992-01-01

    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)

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

    International Nuclear Information System (INIS)

    2007-12-01

    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

  9. Learning Lessons from TMI to Fukushima and Other Industrial Accidents: Keys for Assessing Safety Management Practices

    International Nuclear Information System (INIS)

    Dechy, N.; Rousseau, J.-M.; Dien, Y.; Montmayeul, R.; Llory, M.

    2016-01-01

    The main objective of the paper is to discuss and to argue about transfer, from an industrial sector to another industrial sector, of lessons learnt from accidents. It will be achieved through the discussion of some theoretical foundations and through the illustration of examples of application cases in assessment of safety management practices in Nuclear Power Plant (NPP). The nuclear energy production industry has faced three big ones in 30 years (TMI, Chernobyl, Fukushima) involving three different reactor technologies operated in three quite different cultural, organizational and regulatory contexts. Each of those accident has been the origin of questions, but also generator of lessons, some changing the worldview (see Wilpert and Fahlbruch, 1998) of what does cause an accident in addition to the engineering view about the importance of technical failures (human error, safety culture, sociotechnical interactions). Some of their main lessons were implemented such as improvements of human-machine interfaces ergonomics, recast of some emergency operating procedures, severe accident mitigation strategies and crisis management. Some lessons did not really provide deep changes. It is the case for organizational lessons such as, organizational complexity, management of production pressures, regulatory capture, and failure to learn, etc.

  10. Hydrogen jet recombination under postulated LMFBR accident conditions

    International Nuclear Information System (INIS)

    Wierman, R.W.

    1977-01-01

    Certain conditions may be postulated in LMFBR risk assessments for which the potential of hydrogen release to the reactor containment building needs to be evaluated. The inherent self-ignition characteristics of hydrogen jets entering the air atmosphere of the reactor containment building should be understood for such analyses. If hydrogen jets were to self-ignite (recombine) at the source where they enter the reactor containment building, then undesirable hydrogen accumulation would not occur. Therefore, experiments have been conducted investigating the phenomena associated with the recombination of hydrogen jets under conditions similar to those postulated for LMFBR studies. The data presented define the conditions required for self-ignition of the hydrogen jets

  11. Recriticality, a Key Phenomenon to Investigate in Core Disruptive Accident Scenarios of Current and Future Fast Reactor Designs

    International Nuclear Information System (INIS)

    Maschek, W.; Rineiski, A.; Flad, M.; Kriventsev, V.; Gabrielli, F.; Morita, K.

    2012-01-01

    Final comments and conclusions: • Modern plants, should have performed better under Fukushima type event. • In future fast reactor systems significantly higher active and passive safety features are installed, which should cope with events like Fukushima. • One important lesson: put a focus on rare initiators, accident routes and consequences that are neither expected nor have been observed, events that are categorized under ‘black swans’. • Importance of severe accident research demonstrated - both analytically and experimentally for assessing and interpreting accident scenarios and developments. Precondition for developing preventive & mitigative safety measures. Passive safety measures are in the focus of advanced design options and must work under conditions of multiple loads and aggravating events. • Fast reactor systems behavior as the SFR under severe accident conditions: – In fast spectrum systems as the SFR the core is not in its neutronically most reactive configuration and SFRs may be loaded with MAs for waste management; – Recriticalities have a high probability because of the higher enrichment levels; – Short time scales have to be envisioned for core melt-down; – Decay heat levels might be significantly higher, if MA bearing fuel is involved. • Improve design by measures for prevention and/or mitigation of recriticalities; – High reliability of simulations required for proof; • Assessment of fuel relocated on peripheral structures; • Preventive/mitigating measures should not replace containment measures

  12. Study of labor accidents in the rural environment: analysis of processes and conditions of work

    Directory of Open Access Journals (Sweden)

    Thaís Alves Brito

    2009-01-01

    Full Text Available The modernization of agriculture, that broadenned the mechanization of farming and the agrotoxic use, potentially increased some risks of accidents. The agriculture workers and cattle raising are constantly exposed to several physical, chemical and biological agents, like machine, implements, handly tools, agrotoxics, ectoparaziticides, domestic animals and poisonous animals, which can to bring accidents. The aiming the importance of this working class to economic developing of country, this study was done to identify the working process and accidents that strike the rural population. This article is composed by a specialized literature review between September and December of 2007, which was made consultations to periodical and scientific articles selected through searches in the database of Scielo and Bireme. It was founded few studies related to rural workers, as well as the main articles had as setting of investigation the Southern and Southeastern, mainly in state of São Paulo and Rio Grande do Sul. In relation to work conditions was noticed a high degree of insalubrities which the workers are exposed, such as handly tools, poisonous animals, insecure attitudes because of lack of training and the no use of equipments of individual protection. There are a prevalence of accidents among men, occurring predominantly the typical accidents, the occupational disease and commute accidents. The relationships of work have been modified along the years, being the outsourcing outstanding point, however this work relationship causes legal losses to workers, which in most of the time get without social welfare right

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Studying work practices: a key factor in understanding accidents on the level triggered by a balance disturbance.

    Science.gov (United States)

    Derosier, C; Leclercq, S; Rabardel, P; Langa, P

    2008-12-01

    Accidents on the level (AOL) rank second amongst the most numerous and serious occupational accidents with days lost in France and are a major health and safety problem in every sector of activity. The case study described in this paper was conducted at a metallurgical company with 300 employees. The aims of this work were dual: 1) to extend the general knowledge required for preventing these accidents; 2) to propose prevention measures to this company. Existing data on company occupational accidents were gathered and analysed to identify a work situation that appeared likely to cause AOL. This work situation was analysed in detail. Several risk factors were identified within this work situation, by way of interviews with 12 operators. These risk factors concerned various dimensions of the work situation, particularly its physical dimension (e.g. templates structure) and organisational dimension (e.g. parts availability). Interviews were conducted, focusing on risk factors perceived by operators and involving allo-confrontations based on accounts of four AOL occurring in this situation. Allo-confrontations were interviews confronting operators with a risk occupational situation that was accidental for one of their colleagues, the latter being absent from the interview. Results highlighted the fact that the work practices implemented are key factors in understanding these accidents. This study underlines the role of work practices in AOL causality and prevention. It also provides explanations associated with various work situation dimensions involving adoption of more or less safe work practices. AOL are serious and frequent in occupational situations. Injury claims analysis and interviews in an industrial company emphasise the specific characteristics of an occupational situation and of prevention actions forming the basis of an intervention. The need for a better understanding of factors affecting work practice is highlighted in relation to research.

  15. Historical civilian nuclear accident based Nuclear Reactor Condition Analyzer

    Science.gov (United States)

    McCoy, Kaylyn Marie

    There are significant challenges to successfully monitoring multiple processes within a nuclear reactor facility. The evidence for this observation can be seen in the historical civilian nuclear incidents that have occurred with similar initiating conditions and sequences of events. Because there is a current lack within the nuclear industry, with regards to the monitoring of internal sensors across multiple processes for patterns of failure, this study has developed a program that is directed at accomplishing that charge through an innovation that monitors these systems simultaneously. The inclusion of digital sensor technology within the nuclear industry has appreciably increased computer systems' capabilities to manipulate sensor signals, thus making the satisfaction of these monitoring challenges possible. One such manipulation to signal data has been explored in this study. The Nuclear Reactor Condition Analyzer (NRCA) program that has been developed for this research, with the assistance of the Nuclear Regulatory Commission's Graduate Fellowship, utilizes one-norm distance and kernel weighting equations to normalize all nuclear reactor parameters under the program's analysis. This normalization allows the program to set more consistent parameter value thresholds for a more simplified approach to analyzing the condition of the nuclear reactor under its scrutiny. The product of this research provides a means for the nuclear industry to implement a safety and monitoring program that can oversee the system parameters of a nuclear power reactor facility, like that of a nuclear power plant.

  16. Rapid analysis of key radionuclides in urine and estimation of internal dose for nuclear accident emergency

    International Nuclear Information System (INIS)

    Zhao Shuquan; Hu Heping; Wu Mingyu; Zhu Guoying; Huang Shibin; Liu Shiming

    2005-01-01

    Objective: To estimate the internal doses of a Chinese visiting scholar in the Chernobyl accident. Methods: The contents of 134 Cs and 137 Cs in urine were measured using a Ge(Li) γ-spectrometer. Their internal doses were estimated according to ICRP reports. Dose review of 131I was performed referring to UNSCEAR 2000 report. Results: The effective dose equivalent from 134 Cs, 137 Cs and 131 I were 66 μSv, 88 μSv and 1728 μSv respectively. Their summation was 1.9 mSv. Conclusion: The internal dose from 131 I was 10 times higher than that from 134 Cs and 137 Cs. So, the earlier estimation of internal doses for 131 I is significant in evaluation on radiation injuries of a nuclear reactor accident. (authors)

  17. Application of uncertainty analysis method for calculations of accident conditions for RP AES-2006

    International Nuclear Information System (INIS)

    Zajtsev, S.I.; Bykov, M.A.; Zakutaev, M.O.; Siryapin, V.N.; Petkevich, I.G.; Siryapin, N.V.; Borisov, S.L.; Kozlachkov, A.N.

    2015-01-01

    An analysis of some accidents using the uncertainly assessment methods is given. The list of the variable parameters incorporated the model parameters of the computer codes, initial and boundary conditions of reactor plant, neutronics. On the basis of the performed calculations of the accident conditions using the statistical method, errors assessment is presented in the determination of the main parameters comparable with the acceptance criteria. It was shown that in the investigated accidents the values of the calculated parameters with account for their error obtained from TRAP-KS and KORSAR/GP Codes do not exceed the established acceptance criteria. Besides, these values do not exceed the values obtained in the conservative calculations. A possibility in principle of the actual application of the method of estimation of uncertainty was shown to justify the safety of WWER AES-2006 using the thermal-physical codes KORSAR/GP and TRAP-KS, PANDA and SUSA programs [ru

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

    International Nuclear Information System (INIS)

    Hoover, M.D.; Newton, G.J.; Farrell, R.F.

    1996-01-01

    This qualitative hazard evaluation systematically assessed potential doses to workers during postulated accident conditions at the U.S. Department of Energy's Waste Isolation Pilot Plant (WIPP). Postulated accidents included the spontaneous ignition of a waste drum, puncture of a waste drum by a forklift, dropping of a waste drum from a forklift, and simultaneous dropping of seven drums during a crane failure. The descriptions and estimated frequencies of occurrence for these accidents were developed by the Hazard and Operability Study for CH TRU Waste Handling System (WCAP 14312). The estimated materials at risk, damage ratios, airborne release fractions and respirable fractions for these accidents were taken from the 1995 Safety Analysis Report (SAR) update and from the DOE handbook Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities (DOE-HDBK-3010-94). A Monte Carlo simulation was used to estimate the range of worker exposures that could result from each accident. Guidelines for evaluating the adequacy of defense-in-depth for worker protection at WIPP were adopted from a scheme presented by the International Commission on Radiological Protection in its publication on Protection from Potential Exposure: A Conceptual Framework (ICRP Publication 64). Probabilities of exposures greater than 5, 50, and 300 rem were less than 10 -2 , 10 -4 , and 10 -6 per year, respectively. In conformance with the guidance of DOE standard 3009-94, Appendix A (draft), we emphasize that use of these evaluation guidelines is not intended to imply that these numbers constitute acceptable limits for worker exposure 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, as well as members of the public and the environment

  19. Review of U.S. Army Aviation Accident Reports: Prevalence of Environmental Stressors and Medical Conditions

    Science.gov (United States)

    2017-10-18

    terminology related to an aforementioned stressor or medical condition. Table 1 presents the identified operational stressor with the keywords extracted...USAARL Report No. 2018-02 Review of U.S. Army Aviation Accident Reports: Prevalence of Environmental Stressors and Medical Conditions By Kathryn...Environmental Stressors and Medical Conditions N/A N/A N/A N/A N/A N/A Feltman, Kathryn A. Kelley, Amanda M. Curry, Ian P. Boudreaux, David A. Milam

  20. Failure Mode Estimation of Wolsong Unit 1 Containment Building with respect to Severe Accident Condition

    International Nuclear Information System (INIS)

    Hahm, Dae Gi; Choi, In Kil

    2009-01-01

    The containment buildings in a nuclear power plant (NPP) are final barriers against the exposure of harmful radiation materials at severe accident condition. Since the accident at Three Mile Island nuclear plant in 1979, it has become necessary to evaluate the internal pressure capacity of the containment buildings for the assessment of the safety of nuclear power plants. According to this necessity, many researchers including Yonezawa et al. and Hu and Lin analyzed the ultimate capacity of prestressed concrete containments subjected to internal pressure which can be occurred at sever accident condition. Especially in Wolsong nuclear power plant, the Unit 1 containment structures were constructed in the late 1970 to early 1980, so that the end of its service life will be reached in near future. Since that the complete decommission and reconstruction of the NPP may cause a huge expenses, an extension of the service time can be a cost-effective alternative. To extend the service time of NPP, an overall safety evaluation of the containment building under severe accident condition should be performed. In this study, we assessed the pressure capacity of Wolsong Unit 1 containment building under severe accident, and estimated the responses at all of the probable critical areas. Based on those results, we found the significant failure modes of Wolsong Unit 1 containment building with respect to the severe accident condition. On the other hand, for the aged NPP, the degradation of their structural performance must also be explained in the procedure of the internal pressure capacity evaluation. Therefore, in this study, we performed a parametric study on the degradation effects and evaluated the internal pressure capacity of Wolsong Unit 1 containment building with considering aging and degradation effects

  1. Thermalydraulic processes in the reactor coolant system of a BWR under severe accident conditions

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1990-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermal hydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom atomic power station is presented

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

    Energy Technology Data Exchange (ETDEWEB)

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Sanitation of conditioned radioactive waste after a contamination accident

    International Nuclear Information System (INIS)

    Aeppli, J.

    1980-01-01

    In June 1978, there occurred in the port of Ijmuiden, Netherlands, a contamination incident involving drums originating from Switzerkand and containing radioactive wastes intended to be dumped into the sea. The batch of 207 drums excluded from the sea-dumping action had to be sanitated for the next year dumping in such a manner, that these wastes met the international requirements and could be disposed of by sinking them into the Atlantic. As a consequence of extensive sanitation work, requiring part of the wastes to be newly conditioned and several drums to be packaged again, the total weight of the wastes ready for dumping was doubled. The total radiation exposure for the personnel that took part in the individual phases of sanitation amounted to about 10 man-rem. The main causes for this contamination incident were unusual chemical composition of the concentrate to be solidified, unsufficient quality control and a position not suitabble for transport. The measures taken after this incident intend to avoid similar occurrences in the future. (orig.) [de

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

    International Nuclear Information System (INIS)

    Deuber, H.

    1984-09-01

    Under simulated accident conditions (maximum temperature: 130 0 C) 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.) [de

  6. Analysis of effects of calandria tube uncovery under severe accident conditions in CANDU reactors

    International Nuclear Information System (INIS)

    Rogers, J.T.; Currie, T.C.; Atkinson, J.C.; Dick, R.

    1983-01-01

    A study is being undertaken for the Atomic Energy Control Board to assess the thermal and hydraulic behaviour of CANDU reactor cores under accident conditions more severe than those normally considered in the licensing process. In this paper, we consider the effects on a coolant channel of the uncovery of a calandria tube by moderator boil-off following a LOCA in a Bruce reactor unit in which emergency cooling is ineffective and the moderator heat sink is impaired by the failure of the moderator cooling system. Calandria tube uncovery and its immediate consequences, as described here, constitute only one part of the entire accident sequence. Other aspects of this sequence as well as results of the analysis of the other accident sequences studied will be described in the final report on the project and in later papers

  7. Relationships of working conditions, health problems and vehicle accidents in bus rapid transit (BRT) drivers.

    Science.gov (United States)

    Gómez-Ortiz, Viviola; Cendales, Boris; Useche, Sergio; Bocarejo, Juan P

    2018-04-01

    The aim of this study was to estimate accident risk rates and mental health of bus rapid transit (BRT) drivers based on psychosocial risk factors at work leading to increased stress and health problems. A cross-sectional research design utilized a self-report questionnaire completed by 524 BRT drivers. Some working conditions of BRT drivers (lack of social support from supervisors and perceived potential for risk) may partially explain Bogota's BRT drivers' involvement in road accidents. Drivers' mental health problems were associated with higher job strain, less support from co-workers, fewer rewards and greater signal conflict while driving. To prevent bus accidents, supervisory support may need to be increased. To prevent mental health problems, other interventions may be needed such as reducing demands, increasing job control, reducing amount of incoming information, simplifying current signals, making signals less contradictory, and revising rewards. © 2018 Wiley Periodicals, Inc.

  8. Analysis of some accident conditions in confirmation of the HTGR safety

    International Nuclear Information System (INIS)

    Grebennik, V.N.; Grishanin, E.I.; Kukharkin, N.E.; Mikhailov, P.V.; Pinchuk, V.V.; Ponomarev-Stepnoy, N.N.; Fedin, G.I.; Shilov, V.N.; Yanushevich, I.V.

    1981-01-01

    This report concerns some accident conditions for the HTGR-50 demonstrational reactor which along with the safety features common to the typical HTGR differs in design. The analyses carried out on the accident situations showed that due to the high heat capacity of the graphite core and negative temperature effect of the reactivity the HTGR-50 reactor is effectively selfcontrolled at different perturbations of the reactivity and has low sensitivity to the failure of the core cooling. The primary circuit depressurization accident should be thoroughly studied because of the dangerous consequences i.e. the core overheating and the reactivity release into the environment. As a whole, the studies now in progress show that the problem of the HTGR safety can be successfully solved

  9. Analysis of some accident conditions in confirmation of the HTGR safety

    Energy Technology Data Exchange (ETDEWEB)

    Grebennik, V. N.; Grishanin, E. I.; Kukharkin, N. E.; Mikhailov, P. V.; Pinchuk, V. V.; Ponomarev-Stepnoy, N. N.; Fedin, G. I.; Shilov, V. N.; Yanushevich, I. V. [Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow. Inst. Atomnoj Ehnergii

    1981-01-15

    This report concerns some accident conditions for the HTGR-50 demonstrational reactor which along with the safety features common to the typical HTGR differs in design. The analyses carried out on the accident situations showed that due to the high heat capacity of the graphite core and negative temperature effect of the reactivity the HTGR-50 reactor is effectively selfcontrolled at different perturbations of the reactivity and has low sensitivity to the failure of the core cooling. The primary circuit depressurization accident should be thoroughly studied because of the dangerous consequences i.e. the core overheating and the reactivity release into the environment. As a whole, the studies now in progress show that the problem of the HTGR safety can be successfully solved.

  10. Qualitative analysis of the man-organization system in accident conditions for nuclear installations

    International Nuclear Information System (INIS)

    Farcasiu, Mita; Prisecaru, Ilie

    2010-01-01

    In this paper a model of the human performance investigation of accident conditions in the operation of the nuclear installation is developed. A framework for analyses of the human action in the man-organization system context is achieved. The goal of this model is to identify the possible roots causing human errors which could occur during the evolution of the accident by the qualitative analysis of the interfaces in man-organization system. These interfaces represent the main elements which characterize the implication of the organization in human performance. The results of this paper are the interfaces of the man-organization and their circumstances in which human performance could fail. Also, another result is a pre-designed framework which could help in the investigation of an accident. (authors)

  11. MCCI study for Pressurized Heavy Water Reactor under hypothetical accident condition

    International Nuclear Information System (INIS)

    Verma, Vishnu; Mukhopadhyay, Deb; Chatterjee, B.; Singh, R.K.; Vaze, K.K.

    2011-01-01

    In case of severe core damage accident in Pressurized Heavy Water Reactor (PHWR), large amount of molten corium is expected to come out into the calandria vault due to failure of calandria vessel. Molten corium at high temperature is sufficient to decompose and ablate concrete. Such attack could fail CV by basement penetration. Since containment is ultimate barrier for activity release. The Molten Core Concrete Interaction (MCCI) of the resulting pool of debris with the concrete has been identified as an important part of the accident sequence. MCCI Analysis has been carried out for PHWR for a hypothetical accident condition where total core material is considered to be relocated in calandria vault. Concrete ablation rate in vertical and radial direction is evaluated for rectangular geometry using MEDICIS module of ASTEC Code. Amount of gases released during MCCI is also evaluated. (author)

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

    International Nuclear Information System (INIS)

    Mohamed, F.M.

    2011-01-01

    Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99 Mo 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 99 Mo 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 99 Mo 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.

  13. Comparative study of heterogeneous and homogeneous LMFBR cores in some accident conditions

    International Nuclear Information System (INIS)

    Renard, A.; Evrard, G.

    1978-01-01

    An heterogeneous design and a homogeneous one of a LMFBR core with the same power and similar dimensions are compared from the safety point-of-view. The comparison is performed for several accident conditions, such as Loss-of-Flow and Transient Overpower, with the same failure criteria and model assumptions for both cores. Qualitative trends are deduced from the behaviour of the core designs in the investigated transient conditions. (author)

  14. Limit strains for severe accident conditions. Final report of the EU-project LISSAC

    International Nuclear Information System (INIS)

    Krieg, R.; Seidenfuss, M.

    2003-10-01

    The local failure strains of essential reactor vessel components are investigated. The size influence of the components 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 static and dynamic loads. Within one test family the specimen geometries and the load conditions are similar, the temperature is the same; 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 reactor components. There are indications that for such non-uniform distributions size effects may be stronger than for uniform distributions. Thus size effects on the failure strains and failure processes are determined under realistic conditions. Several tests with nominal identical parameters are performed for small size specimens. In this way some information is obtained about the scatter. A reduced number of tests is carried out for medium size specimens and only a few tests are carried out for large size specimens to reduce the costs to an acceptable level. To manufacture all specimens sufficient material was available from the unused reactor pressure vessel Biblis C consisting of the material 22NiMoCr37. Thus variations of the mechanical material properties, which could impair the interpretation of the test results, are quite small. This has been confirmed by an adequate number of additional quality assurance tests. A key problem was the definition of failure and the determination of the local strains at failure for very different specimens under varying load conditions. Here appropriate 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

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

    International Nuclear Information System (INIS)

    2001-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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. 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

    International Nuclear Information System (INIS)

    Park, Jeong-Yong; KIM, Hyun-Gil; JUNG, Yang-Il; PARK, Dong-Jun; KOO, Yang-Hyun

    2013-01-01

    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. Al 3 Ti-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

  18. Causal Factors and Adverse Conditions of Aviation Accidents and Incidents Related to Integrated Resilient Aircraft Control

    Science.gov (United States)

    Reveley, Mary S.; Briggs, Jeffrey L.; Evans, Joni K.; Sandifer, Carl E.; Jones, Sharon Monica

    2010-01-01

    The causal factors of accidents from the National Transportation Safety Board (NTSB) database and incidents from the Federal Aviation Administration (FAA) database associated with loss of control (LOC) were examined for four types of operations (i.e., Federal Aviation Regulation Part 121, Part 135 Scheduled, Part 135 Nonscheduled, and Part 91) for the years 1988 to 2004. In-flight LOC is a serious aviation problem. Well over half of the LOC accidents included at least one fatality (80 percent in Part 121), and roughly half of all aviation fatalities in the studied time period occurred in conjunction with LOC. An adverse events table was updated to provide focus to the technology validation strategy of the Integrated Resilient Aircraft Control (IRAC) Project. The table contains three types of adverse conditions: failure, damage, and upset. Thirteen different adverse condition subtypes were gleaned from the Aviation Safety Reporting System (ASRS), the FAA Accident and Incident database, and the NTSB database. The severity and frequency of the damage conditions, initial test conditions, and milestones references are also provided.

  19. Spent fuel transport cask thermal evaluation under normal and accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Pugliese, G. [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy); Lo Frano, R., E-mail: rosa.lofrano@ing.unipi.i [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy); Forasassi, G. [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy)

    2010-06-15

    The casks used for transport of nuclear materials, especially the spent fuel element (SPE), must be designed according to rigorous acceptance criteria and standards requirements, e.g. the International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure particularly in a severe accident scenario. The aim of this work was the evaluation of the integrity of a spent fuel cask under both normal and accident scenarios transport conditions, such as impact and rigorous fire events, in according to the IAEA accident test requirements. The thermal behaviour and the temperatures distribution of a Light Water Reactor (LWR) spent fuel transport cask are presented in this paper, especially with reference to the Italian cask designed by AGN, which was characterized by a cylindrical body, with water or air inside the internal cavity, and two lateral shock absorbers. Using the finite element code ANSYS a series of thermal analyses (steady-state and transient thermal analyses) were carried out in order to obtain the maximum fuel temperature and the temperatures field in the body of the cask, both in normal and in accidents scenario, considering all the heat transfer modes between the cask and the external environment (fire in the test or air in the normal conditions) as well as inside the cask itself. In order to follow the standards requirements, the thermal analyses in accidents scenarios were also performed adopting a deformed shape of the shock absorbers to simulate the mechanical effects of a previous IAEA 9 m drop test event. Impact tests on scale models of the shock absorbers have already been conducted in the past at the Department of Mechanical, Nuclear and Production Engineering, University of Pisa, in the '80s. The obtained results, used for possible new licensing approval purposes by the Italian competent Authority of the cask for PWR spent fuel cask transport by the Italian competent Authority, are

  20. Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Sebrell, W.

    1983-07-01

    The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

  1. Potential for containment leak paths through electrical penetration assemblies under severe accident conditions

    International Nuclear Information System (INIS)

    Sebrell, W.

    1983-07-01

    The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350 0 F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage

  2. Multi-phase model development to assess RCIC system capabilities under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kirkland, Karen Vierow [Texas A & M Univ., College Station, TX (United States); Ross, Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Beeny, Bradley [Texas A & M Univ., College Station, TX (United States); Luthman, Nicholas [Texas A& M Engineering Experiment Station, College Station, TX (United States); Strater, Zachary [Texas A & M Univ., College Station, TX (United States)

    2017-12-23

    The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that the system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.

  3. Effect of marine condition on feature of natural circulation after accident in floating nuclear power plant

    International Nuclear Information System (INIS)

    Yang Fan; Zhang Dan; Tan Changlu; Ran Xu; Yu Hongxing

    2015-01-01

    The incline and swing effect on natural circulation of floating nuclear power plant under site black out (SBO) accident is studied using self-developing marine condition system code RELAP5/MC. It shows that, for floating nuclear power plant under marine condition, the pressurizer fluctuating flow rate, the parallel heat sink (steam generator) have significant influences on the direct passive reactor heat removal (PRHR) system, which is different from other secondary PRHR under marine condition. The flow exchange between the loop and the pressurizer have major effect on cooling capacity for the left side loop. (authors)

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

    International Nuclear Information System (INIS)

    Nourbakhsh, H.P.

    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

  5. The analysis of a condition of an accident rate on highways of Tajikistan

    International Nuclear Information System (INIS)

    Davlatshoev, R.A.; Tursunov, A.A.

    2005-01-01

    In this clause the results of the analysis of an accident rate on highways of Tajikistan, according to the official information State Automobile Inspection the Ministry of Internal affairs of Republic of Tajikistan, and research of safe movement of automobiles in mountain conditions are given. On the basis of the qualitative and quantitative analysis, the ways of safe movement on roads of Republic of Tajikistan are determined

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

    International Nuclear Information System (INIS)

    Santhosh; Gera, B.; Kumar, Mithilesh; Thangamani, I.; Prasad, Hari; Srivastava, A.; Dutta, Anu; Sharma, Pavan K.; Majumdar, P.; Verma, V.; Mukhopadhyay, D.; Ganju, Sunil; Chatterjee, B.; Sanyasi Rao, V.V.S.; Lele, H.G.; Ghosh, A.K.

    2009-07-01

    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)

  7. Condition monitoring a key component in the preventive maintenance

    International Nuclear Information System (INIS)

    Isar, C.

    2006-01-01

    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

  8. Limit strains for severe accident conditions. Synthesis report of the EU-project LISSAC

    International Nuclear Information System (INIS)

    Krieg, R.; Seidenfuss, M.

    2003-10-01

    The local failure strains of essential reactor vessel components are investigated. The size influence of the components 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 static and dynamic loads. Within one test family the specimen geometry and the load conditions are similar, the temperature is the same; 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 reactor components. To manufacture all specimens sufficient material was available from the unused reactor pressure vessel Biblis C. Thus variations of the mechanical material properties, which could impair the interpretation of the test results, are rather small. This has been confirmed by an adequate number of additional quality assurance tests. A key problem was to determine the local strain at failure. 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 calculations, for instance. To deepen the understanding of structural degradation and fracture and to allow extrapolations, advanced computational methods including damage models have been developed and validated. Several approaches were tried in parallel including so-called non-local concepts and descriptions of stochastic properties at grain size level. The experimental results indicate that stresses versus dimensionless deformations are approximately size independent up to failure for specimens of similar geometry under similar load conditions. Also the maximum stress is approximately size independent, if failure occurs after the maximum stress is reached

  9. Comparison of US/FRG accident condition models for HTGR fuel failure and radionuclide release

    International Nuclear Information System (INIS)

    Verfondern, K.

    1991-03-01

    The objective was to compare calculation models used in safety analyses in the US and FRG which describe fission product release behavior from TRISO coated fuel particles under core heatup accident conditions. The frist step performed is the qualitative comparison of both sides' fuel failure and release models in order to identify differences and similarities in modeling assumptions and inputs. Assumptions of possible particle failure mechanisms under accident conditions (SiC degradation, pressure vessel) are principally the same on both sides though they are used in different modeling approaches. The characterization of a standard (= intact) coated particle to be of non-releasing (GA) or possibly releasing (KFA/ISF) type is one of the major qualitative differences. Similar models are used regarding radionuclide release from exposed particle kernels. In a second step, a quantitative comparison of the calculation models was made by assessing a benchmark problem predicting particle failure and radionuclide release under MHTGR conduction cooldown accident conditions. Calculations with each side's reference method have come to almost the same failure fractions after 250 hours for the core region with maximum core heatup temperature despite the different modeling approaches of SORS and PANAMA-I. The comparison of the results of particle failure obtained with the Integrated Failure and Release Model for Standard Particles and its revision provides a 'verification' of these models in this sense that the codes (SORS and PANAMA-II, and -III, respectively) which were independently developed lead to very good agreement in the predictions. (orig./HP) [de

  10. Postulated accident conditions for air cleaning systems and radiological dose assessments for containment options

    International Nuclear Information System (INIS)

    Hilliard, R.K.; Postma, A.K.

    1975-01-01

    Ambient conditions and performance requirements for emergency air cleaning systems applicable to commercial LMFBR plants were studied. The focus of this study centered on aerosol removal under hypothetical core disruptive accident conditions. Effort completed includes a review of air cleaning systems related to LMFBR plants, selection of three reference containment system designs, postulation of the EACS design basis accident (EACS-DBA), analysis of thermal conditions resulting from the DBA, analysis of aerosol transport behavior following the DBA, and an estimate of bone dose at the site boundary for each of the reference plant designs. Reference plant concepts were a single containment system (e.g., FFTF), a double containment system (e.g., CRBRP with closed head compartment), and a containment-confinement design in which an inerted, sealed primary volume was located within a ventilated building whose exhaust was filtered. The reference design basis accident selected here involved release to the inner containment system of 1 percent of non-volatile solids and plutonium, 25 percent of core halogens, 25 percent of core volatile solids, 100 percent of core noble gases, 68 lbs of sodium vapor and 5000 lbs of liquid sodium. 13 references. (U.S.)

  11. Fission product releases at severe LWR accident conditions: ORNL/CEA measurements versus calculations

    Energy Technology Data Exchange (ETDEWEB)

    Andre, B.; Ducros, G.; Leveque, J.P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Osborne, M.F.; Lorenz, R.A. [Oak Ridge National Lab., TN (United States); Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Dept. de Protection de l`Environnement et des Installations

    1995-12-31

    Experimental programs in the United States and France have followed similar paths in supplying much of the data needed to analyze severe accidents. Both the HI/VI program, conducted at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (NRC), and the HEVA/VERCORS program, supported by IPSN-Commissariat a l`Energie Atomique (CEA) and carried out at the Centre d`Etudes Nucleaires de Grenoble, have studied fission product release from light water reactor (LWR) fuel samples during test sequences representative of severe accidents. Recognizing that more accurate data, i.e., a better defined source term, could reduce the safety margins included in the rather conservative source terms originating from WASH-1400, the primary objective of these programs has been to improve the data base concerning fission product release and behavior at high temperatures. To facilitate the comparison, a model based on fission product diffusion mechanisms that was developed at ORNL and adapted with CEA experimental data is proposed. This CEA model is compared with the ORNL experimental data in a blind test. The two experimental programs used similar techniques in out-of-pile studies. Highly irradiated fuel samples were heated in radiofrequency induction furnaces to very high temperatures (up to 2700 K at ORNL and 2750 K at CEA) in oxidizing (H{sub 2}O), reducing (H{sub 2}) or mixed (H{sub 2}O+H{sub 2}) environments. The experimental parameters, which were chosen from calculated accident scenarios, did not duplicate specific accidents, but rather emphasized careful control of test conditions to facilitate extrapolation of the results to a wide variety of accident situations. This paper presents a broad and consistent database from ORNL and CEA release results obtained independently since the early 1980`S. A comparison of CORSOR and CORSOR Booth calculations, currently used in safety analysis, and the experimental results is presented and

  12. Living conditions in the contaminated territories of Bielorussia 8 years after the Chernobyl accident

    International Nuclear Information System (INIS)

    Heriard-Dubreuil, G.; Girard, P.

    1997-01-01

    Living conditions in the contaminated territories of Bielorussia after the Chernobyl accident: evaluation of the situation in the district of Chetchersk in Bielorussia. This article presents an analysis of the social and economic aspects of radiological protection in the territories contaminated by the Chernobyl accident. It is based on the results of two surveys performed in 1994 on the living conditions of the inhabitants of a territorial community located in Bielorussia, 180 km north of Chernobyl. The first part presents the radiological post-accident situation of the district, together with an analysis of this situation's demographic impact since 1986. The second part presents a description of the modes of exposure of the inhabitants of the contaminated territories and an assessment of he various countermeasures programmes initiated by the authorities in the legislative framework of 1991. The last part addresses the economic aspects of the Chetchersk district and an evaluation of the consequences of the radiological situation on the economic, and above all agricultural, activities of the district.The conclusion highlights the difficulties that face the Byelorussian authorities today. The now definitive presence of inhabitants in a durably contaminated environment poses a new category of problems. The objectives of radiological protection have to be reshaped within a set of constraints of different types, notably social and economic. The development of radiological safety cannot be dissociated from a return to quality living in these territories. This necessarily entails re-establishing a climate of social confidence. The initial legislative plan for post-accident management must be adapted to give greater autonomy to local participants in the reconstruction of satisfactory living conditions. (authors)

  13. Key Conditions for Successful Serial Entrepreneurship in Healthcare.

    Science.gov (United States)

    Piron, Cameron

    2017-01-01

    As a serial entrepreneur in the medical device industry, the author embraces Snowdon's (2017) effort to create and stimulate dialogue among experts in health system innovation in an effort to define and support Canada's innovation agenda. In this paper, he outlines some of the attributes and skills that companies need to launch their products and scale their companies. He also identifies the main conditions of an innovation ecosystem that create the necessary infrastructure to enable and support highly successful companies while allowing them to accelerate their growth.

  14. Accident identification system with automatic detection of abnormal condition using quantum computation

    International Nuclear Information System (INIS)

    Nicolau, Andressa dos Santos; Schirru, Roberto; Lima, Alan Miranda Monteiro de

    2011-01-01

    Transient identification systems have been proposed in order to maintain the plant operating in safe conditions and help operators in make decisions in emergency short time interval with maximum certainty associated. This article presents a system, time independent and without the use of an event that can be used as a starting point for t = 0 (reactor scram, for instance), for transient/accident identification of a pressurized water nuclear reactor (PWR). The model was developed in order to be able to recognize the normal condition and three accidents of the design basis list of the Nuclear Power Plant Angra 2, postulated in the Final Safety Analysis Report (FSAR). Were used several sets of process variables in order to establish a minimum set of variables considered necessary and sufficient. The optimization step of the identification algorithm is based upon the paradigm of Quantum Computing. In this case, the optimization metaheuristic Quantum Inspired Evolutionary Algorithm (QEA) was implemented and works as a data mining tool. The results obtained with the QEA without the time variable are compatible to the techniques in the reference literature, for the transient identification problem, with less computational effort (number of evaluations). This system allows a solution that approximates the ideal solution, the Voronoi Vectors with only one partition for the classes of accidents with robustness. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    Gomes, Daniel de Souza; Baptista Filho, Benedito; Oliveira, Fabio Branco de; Giovedi, Claudia

    2015-01-01

    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)

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

    International Nuclear Information System (INIS)

    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)

  18. NPP physical protection and information security as necessary conditions for reducing nuclear and radiation accident risks

    International Nuclear Information System (INIS)

    Pogosov, O.Yu.; Derevyanko, O.V.

    2017-01-01

    The paper focuses on the fact that nuclear failures and incidents can lead to radioactive contamination of NPP premises. Nuclear and radiation hazard may be caused by malefactors in technological processes when applying computers or inadequate control in case of insufficient level of information security.The researchers performed analysis of factors for reducing risks of nuclear and radiation accidents at NPPs considering specific conditions related to information security of NPP physical protection systems. The paper considers connection of heterogeneous factors that may increase the risk of NPP accidents, possibilities and ways to improve adequate modelling of security of information with limited access directly related to the functioning of automated set of engineering and technical means for NPP physical protection. Within the overall Hutchinson formalization, it is proposed to include additional functional dependencies on indicators specific for NPPs into analysis algorithms.

  19. A dynamic model for the study of evacuation under accident conditions

    International Nuclear Information System (INIS)

    Boeri, G.C.; Caracciolo, R.; Sepede, M.; Casiroli, F.; Rodriguez, R.

    1987-01-01

    Information techniques and models are being used to simulate the evacuation of people living around a nuclear power plant and these methods are being increasingly used for planning purposes. In this study vehicular mobility on a complex road network following an accident has been considered and applied to the specifications of the emergency plans. Simulation tests of the mobility relevant to different combinations of time and meteorological conditions have been undertaken through use of the computer code TRIPS in order to assess the impact on the road network of an accident situation and the preparedness measures. The study has led to a description of the accessibility time curves related to the population gathering centres and has enabled identification of the best routes in order to achieve the minimum travel time. (author)

  20. SIMBATH 1976-1992, seventeen years of experimental investigation of key issues concerned with severe reactor accidents

    International Nuclear Information System (INIS)

    Kaiser, A.; Peppler, W.; Will, H.

    1994-01-01

    The course of the initiating phase of severe fast reactor accidents is determined by early material motion. In simulation experiments (SIMBATH, simulation experiments in fuel element mock-ups with thermite) the behavior of single pin, 7 pin, 19 pin, 37 pin bundles undergoing meltdown was investigated. Thermite (Al + Fe 2 O 3 ) filled tubes were used to simulate fuel rods, while exothermal heat of the thermite reaction simulated the nuclear heat. The energy of 3.4 kJ per centimeter of pin length resulted in melting temperature of about 3200 K. SIMBATH is an out-of-pile experimental program with non-radioactive materials which provided the possibility to perform numerous experiments. The x-ray high speed photography used in the test enabled to visualise material motion and relocation qualitatively, and furthermore to gain quantitative results by additionally installed photodiodes. The results of the experiment serve as a database to evaluate physical phenomena relevant to be modelled by computer codes (SIMMER) and to verify the codes. The experiments were carried out either in stagnant sodium with an axial temperature gradient, or in flowing sodium, simulating unprotected loss of flow (ULOF) or unprotected transient overpower accidents (UTOP) conditions, respectively

  1. Predictability of iodine chemistry in the containment of a nuclear power plant under hypothetical severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L.E.; Vela-Garcia, M.; Fontanet, J. [Unit of Nuclear Safety Research, CIEMAT, Madrid (Spain)

    2007-07-01

    One of the areas of top interest in the arena of severe accidents to get an accurate prediction of Source Term is Iodine Chemistry. In this paper an assessment of the current capability of MELCOR and ASTEC to predict iodine chemistry within containment in case of a postulated severe accident has been carried out. The experiments FPT1 and FPT2 of the PHEBUS-FP project have been used for comparisons, since they were carried out under rather different containment conditions during the chemistry phase (subcooled vs. saturated sump or acid vs. alkaline pH), which makes them very suitable to assess the current modeling capability of in-containment iodine chemistry models. The results obtained indicate that, even though, both integral codes have specific areas related to iodine chemistry that should be further developed and that their approach to the matter is drastically different, at present ASTEC-IODE allows for a more comprehensive simulation of the containment iodine chemistry. More importantly, lack of maturity of these codes would potentially maximize the so-called user-effect, so that it would be highly recommendable to perform sensitivity studies around iodine chemistry aspects when calculating Source Term scenarios. Key aspects needed of further research are: gaseous iodine chemistry (absent in MELCOR), organic iodine chemistry and adsorption/desorption on/from containment surfaces. (authors)

  2. Basic study on BWR plant behavior under the condition of severe accident (2)

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Ueda, Masataka; Sasaki, Hajime

    2016-01-01

    In this paper, we report on the results using the BWR plant simulator about the plant behavior under the condition of the two types of severe accidents that LOCA occurs but ECCS fails the water irrigation into the reactor core and SBO occurs and at the same time the reclosed failure of SRV occurs. The simulation experiments were carried out for the cases that LOCA has occurred in the main feed-water piping. As for the results about the relationship between the LOCA area and the time from LOCA occurs until the fuel temperature rise start, the effect that RCIC operated was extremely big for small and middle LOCA area. In the case of main feed-water system LOCA, the core water level suddenly decreased for large LOCA of 2000 cm"2 area, however, if the irrigation into the reactor core was carried out 30 min after LOCA occurrence, the core had little damage. In addition, the H_2 concentration in the containment vessel did not exceed both limits of H_2 explosion nor detonation. The pressure of the containment vessel was around 3 kg/cm"2 of design value, so the soundness of the containment vessel was confirmed. On the other hand, for the accident of SBO with reclosed failure of SRV, it has been shown that the accidents continue to progress rapidly as compared with the case of normally operating of SRV. Because SRV has the function that keep the inside pressure of reactor core by repeating opened and closed in response of the inside pressure and prevent the decrease of water level inside reactor core. However, if the irrigation into the reactor core was carried out 30 min after SBO occurrence, the core had little damage and also the H_2 concentration in the containment vessel did not exceed limits of H_2 explosion. Further, as for the accident of reclosed failure of SRV, it has been shown that there are very good correspondence with the simulation results of main steam piping LOCA of area 180 cm"2 corresponding to the inlet cross-sectional area SRV installed on the piping

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

    International Nuclear Information System (INIS)

    Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

    2015-01-01

    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

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

    International Nuclear Information System (INIS)

    Gokhale, Onkar; Mukhopadhyay, D.; Chatterjee, B.; Singh, R.K.

    2015-01-01

    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

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

    Directory of Open Access Journals (Sweden)

    Élise Lépy

    2016-08-01

    Full Text Available (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.

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

    International Nuclear Information System (INIS)

    Le Dai Dien; Hoang Minh Giang; Nguyen Thi Thanh Thuy; Nguyen Thi Tu Oanh; Le Thi Thu; Pham Tuan Nam; Tran Van Trung; Le Van Hong; Vo Thi Huong

    2014-01-01

    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 9 th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10 th 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

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

    International Nuclear Information System (INIS)

    Adorni, M.; Esmaili, H.; Grant, W.; Hollands, T.; Hozer, Z.; Jaeckel, B.; Munoz, M.; Nakajima, T.; Rocchi, F.; Strucic, M.; ); Tregoures, N.; Vokac, P.; Ahn, K.I.; Bourgue, L.; Dickson, R.; Douxchamps, P.A.; Herranz, L.E.; Jernkvist, L.O.; Amri, A.; Kissane, M.P.; )

    2015-01-01

    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

  8. Fission product release from HTGR fuel under core heatup accident conditions - HTR2008-58160

    International Nuclear Information System (INIS)

    Verfondern, K.; Nabielek, H.

    2008-01-01

    Various countries engaged in the development and fabrication of modern fuel for the High Temperature Gas-Cooled Reactor (HTGR) have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under operating and accidental conditions of future HTGRs. Within the IAEA directed Coordinated Research Project CRP6 on 'Advances in HTGR Fuel Technology Development' active since 2002, the 13 participating Member States have agreed upon benchmark studies on fuel performance during normal operation and under accident conditions. While the former has been completed in the meantime, the focus is now on the extension of the national code developments to become applicable to core heatup accident conditions. These activities are supported by the fact that core heatup simulation experiments have been resumed recently providing new, highly valuable data. Work on accident performance will be - similar to the normal operation benchmark - consisting of three essential parts comprising both code verification that establishes the correspondence of code work with the underlying physical, chemical and mathematical laws, and code validation that establishes reasonable agreement with the existing experimental data base, but including also predictive calculations for future heating tests and/or reactor concepts. The paper will describe the cases to be studied and the calculational results obtained with the German computer model FRESCO. Among the benchmark cases in consideration are tests which were most recently conducted in the new heating facility KUEFA. Therefore this study will also re-open the discussion and analysis of both the validity of diffusion models and the transport data of the principal fission product species in the HTGR fuel materials as essential input data for the codes. (authors)

  9. Accident management for severe accidents

    International Nuclear Information System (INIS)

    Bari, R.A.; Pratt, W.T.; Lehner, J.; Leonard, M.; Disalvo, R.; Sheron, B.

    1988-01-01

    The management of severe accidents in light water reactors is receiving much attention in several countries. The reduction of risk by measures and/or actions that would affect the behavior of a severe accident is discussed. The research program that is being conducted by the US Nuclear Regulatory Commission focuses on both in-vessel accident management and containment and release accident management. The key issues and approaches taken in this program are summarized. 6 refs

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

    International Nuclear Information System (INIS)

    Brown, G.S.; Cashwell, J.W.; Apple, M.L.

    1991-01-01

    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. This report discusses the shipment of these highly radioactive materials

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

    International Nuclear Information System (INIS)

    Norwood, K.S.; Collins, J.L.; Osborne, M.F.; Lorenz, R.A.; Wichner, R.P.

    1984-01-01

    Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000 0 C 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

  12. Evaluation of the leakage behavior of inflatable seals subject to severe accident conditions

    International Nuclear Information System (INIS)

    Parks, M.B.

    1989-11-01

    Sandia National Laboratories, under the sponsorship of the United States Nuclear Regulatory Commission, is currently developing test validated methods to predict the pressure capacity of light water reactor containment buildings when subjected to postulated severe accident conditions. These conditions are well beyond the design basis. Scale model tests of steel and reinforced concrete containments have been conducted as well as tests of typical containment penetrations. As a part of this effort, a series of tests was recently conducted to determine the leakage behavior of inflatable seals. These seals are used to prevent leakage around personnel and escape lock doors of some containments. The results of the inflatable seals tests are the subject of this report. Inflatable seals were tested at both room temperature and at elevated temperatures representative of postulated severe accident conditions. Both aged (radiation and thermal) and unaged seals were included in the test program. The internal seal pressure at the beginning of each test was varied to cover the range of seal pressures actually used in containments. For each seal pressure level, the external (containment) pressure was increased until significant leakage past the seals was observed. Parameters that were monitored and recorded during the tests were the internal seal pressure, chamber pressure, leakage past the seals, and temperature of the test chamber and fixture to which the seals were attached. 8 refs., 34 figs., 7 tabs

  13. Study of behavior of cermet fuel elements on IGR reactor under RIA type accident condition

    International Nuclear Information System (INIS)

    Vasil'ev, Yu.S.; Vurim, A.D.; Koltyshev, S.M.; Pakhnits, V.A.; Tukhvatulin, Sh.T.; Popov, V.V.; Ryzhkov, A.N.

    1996-01-01

    In 1993 December in IGR reactor of Inst. of Atomic Energy of National Nuclear Center of Republic of Kazakstan the second batch of in-pile testing of perspective cermet fuel elements under the condition, simulating RIA type accident was conducted. In the second batch of testing during eight start-ups 10 cermet fuel elements were examined. Among which 8 of monolith type and 2 fuel elements with false jacket beside cladding (FJF), as well as, 6 standard fuel elements of WWER-1000 type reactor with dioxide fuel were tested. 2 fuel elements - cermet and standard were placed into capsule filled with water. To measure energy release for the each start-up two fission monitor and inside core control gauge were placed. In all the start-ups operation mode of IGR was neutron pulse. Power of fuel element kept changing from 151 to 336 k W; energy release was 38-93 kJ/gr m 235 U; maximum temperature of cermet fuel was 1943-2173 K, of dioxide fuel - 1923-2843 K. The testing has demonstrated that operability of cermet fuel elements under reactivity accident condition with pulse width of 0,2 s is, at least, not less that operability of dioxide fuel elements, through advantages of cermet fuel under these conditions are revealed to the least extent

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

    International Nuclear Information System (INIS)

    Ha, Jun Su; Kim, Sungyeop

    2014-01-01

    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

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

    International Nuclear Information System (INIS)

    Lambert, L.D.; Parks, M.B.

    1995-10-01

    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

  16. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 Degree-Sign C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, 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 furnace system and the associated

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  18. Oxidation behavior of fuel cladding tube in spent fuel pool accident condition

    International Nuclear Information System (INIS)

    Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Nakashima, Kazuo; Tojo, Masayuki

    2017-01-01

    In spent fuel pool (SFP) under loss-of-cooling or loss-of-coolant severe accident condition, the spent fuels will be exposed to air and heated by their own residual decay heat. Integrity of fuel cladding is crucial for SFP safety therefore study on cladding oxidation in air at high temperature is important. Zircaloy-2 (Zry2) and zircaloy-4 (Zry4) were applied for thermogravimetric analyses (TGA) in different temperatures in air at different flow rates to evaluate oxidation behavior. Oxidation rate increased with testing temperature. In a range of flow rate of air which is predictable in spent fuel lack during a hypothetical SFP accident, influence of flow rate was not clearly observed below 950degC for the Zry2, or below 1050degC for Zry4. In higher temperature, oxidation rate was higher in high rate condition, and this trend was seen clearer when temperature increased. Oxide layers were carefully examined after the TGA analyses and compared with mass gain data to investigate detail of oxidation process in air. It was revealed that the mass gain data in pre-breakaway regime reflects growth of dense oxide film on specimen surface, meanwhile in post-breakaway regime, it reflects growth of porous oxide layer beneath fracture of the dense oxide film. (author)

  19. Should evacuation conditions after a nuclear accident be revised?; Faut-il revoir les conditions d'evacuation a la suite d'un accident nucleaire?

    Energy Technology Data Exchange (ETDEWEB)

    Nifenecker, H.

    2011-07-01

    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

  20. Assessment of radiation doses in normal operation, upset accident conditions at the Olkiluoto nuclear waste facility

    International Nuclear Information System (INIS)

    Rossi, J.; Raiko, H.; Suolanen, V.

    2009-09-01

    Radiation doses for workers of the facility, for inhabitants in the environment and for terrestrial ecosystem possibly caused by the encapsulation and disposal facility to be built at Olkiluoto during its operation were considered in the study. The study covers both the normal operation of the plant and some hypothetical incidents and accidents. Release through the ventilation stack is assumed to be filtered both in normal operation and in hypothetical abnormal fault and accident cases. Calculation of the offsite doses from normal operation is based on the hypothesis that on average one fuel pin per 100 fuel bundles for all batches of spent fuel transported to the encapsulation facility is leaking. The release magnitude in incidents and accidents is based on the event chains, which lead to loss of fuel pin tightness followed by a discharge of radionuclides into the handling space and to some degree to the atmosphere through the ventilation stack equipped with redundant filters. The critical group is conservatively assumed to live at the distance of 200 meters from the encapsulation and disposal plant and thus it will receive the largest doses in most dispersion conditions. The dose value to a member of the critical group was calculated on the basis of the weather data in such a way that greater dose than obtained here is caused only in 0.5 percent of dispersion conditions. The results obtained indicate that during normal operation the doses to workers remain small and the dose to the member of the critical group is less than 0,001 mSv per year. In the case of hypothetical fault and accident releases the offsite doses do not exceed either the limit values set by the safety authority. The highest dose rates to the reference organisms of the terrestrial ecosystem with conservative assumptions from the largest release were estimated to be of the order of 100 μ Gy/h at the distance of 200 m. As a chronic exposure this dose rate is expected to bring up detrimental

  1. Flow behavior of volume-heated boiling pools: implications with respect to transition phase accident conditions

    International Nuclear Information System (INIS)

    Ginsberg, T.; Jones, O.C. Jr.; Chen, J.C.

    1979-01-01

    Observations of two-phase flow fields in single-component volume-heated boiling pools were made. Photographic observations, together with pool-average void fraction measurements, indicate that the churn-turbulent flow regime is stable for superficial vapor velocities up to nearly five times the Kutateladze dispersal limit. Within this range of conditions, a churn-turbulent drift flux model provides a reasonable prediction of the pool-average void fraction data. An extrapolation of the data to transition phase accident conditions suggests that intense boilup could occur where the pool-average void fraction would be >0.6 for steel vaporization rates equivalent to power levels >1% of nominal liquid-metal fast breeder reactor power density. The extended stability of bubbly flow to unusually large vapor fluxes and void fractions, observed in some experiments, is a major unresolved issue

  2. A risk-based evaluation of LMFBR containment response under core disruptive accident conditions

    International Nuclear Information System (INIS)

    Hartung, J.; Berk, S.

    1978-01-01

    Probabilistic risk methodology is utilized to evaluate the failure modes and effects of LMFBR containment systems under Core Disruptive Accident (CDA) conditions. First, the potential causes of LMFBR containment failure under CDA conditions are discussed and categorized. Then, a simple scoping-type risk assessment of a reference design is presented to help place these potential causes of failure in perspective. The highest risk containment failure modes are identified for the reference design, and several design and research and development options which appear capable of reducing these risks are discussed. The degree to which large LMFBR containment systems must mitigate the consequences of CDA's to achieve a level of risk (for LMFBR's) comparable to the already very low risk of contemporary LWR's is explored. Based on the results of this evaluation, several suggestions are offered concerning CDA-related design goals and research and development priorities for large LMFBR's. (author)

  3. Oxidation of SiC cladding under Loss of Coolant Accident (LOCA) conditions in LWRs

    International Nuclear Information System (INIS)

    Lee, Y.; Yue, C.; Arnold, R. P.; McKrell, T. J.; Kazimi, M. S.

    2012-01-01

    An experimental assessment of Silicon Carbide (SiC) cladding oxidation rate in steam under conditions representative of Loss of Coolant Accidents (LOCA) in light water reactors (LWRs) was conducted. SiC oxidation tests were performed with monolithic alpha phase tubular samples in a vertical quartz tube at a steam temperature of 1140 deg. C and steam velocity range of 1 to 10 m/sec, at atmospheric pressure. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO 2 volatilization) was experimentally observed. The weight loss rate increased with increasing steam flow rate. Over the range of test conditions, SiC oxidation rates were shown to be about 3 orders of magnitude lower than the oxidation rates of zircaloy 4. A SiC volatilization correlation for developing laminar flow in a vertical channel is formulated. (authors)

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

    International Nuclear Information System (INIS)

    Hartig, S.H.; Wurz, D.E.; Arnitz, T.; Ruedinger, V.

    1985-01-01

    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

  5. A radiation condition in some regions with more pronounced effect of the Chernobyl accident

    International Nuclear Information System (INIS)

    Ivanov, I.V.; Ivanov, I.M.

    1993-01-01

    The radioecological condition of the Devin region situated in the Rodopes mountain (Bulgaria) has been investigated for the period October 1992 - March 1993. It is believed that the Rodopes were more significantly affected by the Chernobyl accident in comparison with other regions of Bulgaria. Some regions near Kozloduy NPP have been chosen for comparing, for which there are more detailed investigations of the anthropogenic radiation effects. Analysis of the background radiation is made, specific soil and water samples are tested. The alterations in the radiation conditions of the Devin region are analysed. Some conclusions and predictions for the trends in further alterations of the background radiation are made. As a result a draft regional program for environment protection reclamation is prepared. (V.K.)

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

    International Nuclear Information System (INIS)

    Sweet, D.W.; Roberts, G.J.

    1994-01-01

    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

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

    International Nuclear Information System (INIS)

    Girault, N.; Dickinson, S.; Funke, F.; Auvinen, A.; Herranz, L.; Krausmann, E.

    2006-01-01

    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

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

    International Nuclear Information System (INIS)

    2015-03-01

    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 (UO 2 ) had been manufactured since 1978 and were successfully demonstrated in irradiation and accident tests. In 1980, HTR fuels changed to low enriched uranium UO 2 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

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

    International Nuclear Information System (INIS)

    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

  10. Preliminary experiment design of graphite dust emission measurement under accident conditions for HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei, E-mail: pengwei@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Chen, Tao; Sun, Qi; Wang, Jie [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • A theoretical analysis is used to predict the total graphite dust release for an AVR LOCA. • Similarity criteria must be satisfied between the experiment and the actual HTGR system. • Model experiments should be conducted to predict the graphite dust resuspension rate. - Abstract: The graphite dust movement behavior is significant for the safety analyses of high-temperature gas cooled reactor (HTGR). The graphite dust release for accident conditions is an important source term for HTGR safety analyses. Depressurization release tests are not practical in HTGR because of a radioactivity release to the environment. Thus, a theoretical analysis and similarity principles were used to design a group of modeling experiments. Modeling experiments for fan start-up and depressurization process and actual experiments of helium circulator start-up in an HTGR were used to predict the rate of graphite dust resuspension and the graphite dust concentration, which can be used to predict the graphite dust release during accidents. The modeling experiments are easy to realize and the helium circulator start-up test does not harm the reactor system or the environment, so this experiment program is easily achieved. The revised Rock’n’Roll model was then used to calculate the AVR reactor release. The calculation results indicate that the total graphite dust releases during a LOCA will be about 0.65 g in AVR.

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

    Directory of Open Access Journals (Sweden)

    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

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

    International Nuclear Information System (INIS)

    Yefanov, A.D.; Kalyakin, C.G.; Loshchinin, V.M.; Pomet'ko, R.S.; Sergeev, V.V.; Shumsky, R.V.

    1996-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  14. Guidance of reactor operators and TSC personnel with the severe accident management guidance under shutdown and low power conditions

    International Nuclear Information System (INIS)

    Van Haesendonck, M.F.; Prior, R.P.

    2000-01-01

    The Westinghouse Owners Group Severe Accident Management Guidance (WOG SAMG) was developed between 1991 and 1994. The primary goals for severe accident management that form the basis of the WOG SAMG are to terminate any radioactive releases to the environment; to prevent failure of any containment fission product boundary and to return the plant to a controlled stable condition. The WOG SAMG is primarily a TSC tool for mitigation of low probability core damage events. The philosophy is that control room operators should remain focused on the prevention of core damage, whereas the TSC personnel should concentrate on the mitigation of the severe accident. The symptom based package is built up as a structured process for choosing appropriate actions based on actual plant conditions. No detailed knowledge of severe accident phenomena is required. The scope of the WOG SAMG is limited to severe accidents resulting from initiating events occurring during full power operation. However, a number of studies such as the EdF EPS 1300 Probabilistic Safety Assessment (PSA), the shutdown Probabilistic Risk Assessment (PRA) for Surry, the BERA shutdown PRA for Beznau, the EPRI/ Westinghouse ORAM methodology etc. have shown that the frequency of core damage (a severe accident) during shutdown and low power operation can be of the same order of magnitude as for full power operation. The at-power SAMG is viewed as the resolution of the severe accident issue. Similarly, it is expected that as shutdown PRAs mature, the final resolution of the severe accident issue will lie in SAMG for low power and shutdown operation. Therefore in resolution of this issue, Westinghouse has developed the Shutdown Severe Accident Management Guidance (SSAMG) which gives guidance for both control room and TSC personnel to mitigate a severe accident under shutdown or low power conditions. In the last few years, many LWR plants have been implementing SAMG. In the US, all plants have developed SAMG, and many

  15. Experiments to evaluate behavior of containment piping bellows under severe accident conditions

    International Nuclear Information System (INIS)

    Lambert, L.D.; Parks, M.B.

    1993-01-01

    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. In a severe accident they may be subjected to high pressure and temperature, and 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, Albuquerque, New Mexico. Several different bellows geometries, representative of actual containment bellows, are being subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of thirteen tests have been conducted. The tests showed that withstanding relatively large bellows are capable of deformations, up to, or near, the point of full compression before developing leakage. The test data is presented and discussed

  16. Experience in health care organization for victims of Chernobyl accident under conditions of spatial hospitals

    International Nuclear Information System (INIS)

    Nadezhina, N.M.

    1990-01-01

    Experience in organization of health care for victims of Chernobyl accidents under conditions of spatial hospitals are discussed taking into account patients with residual contamination of skin and clothe. A necessity of well-adjusted organization activites, including an inpatient clinic with well-equipped reception, dosimetric, haryological and bacteriological laboratories, an intensive care department, a surgical (burn) department, a blood transfusion laboratory and equipment for plasmopheresis and hemosorption is marked. Therapy of such patients should be developed along the following lines: 1) prevention and therapy of infectious complications; 2) blood cell substitution therapy; 3) bone marrow transplantation; 4) detoxicating therapy; 5) correction of water-electrolyte metabolism; 6) therapy of local radiation injuries

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

    International Nuclear Information System (INIS)

    Rolevitch, I.; Pachkievitch, V.; Petroviet, V.; Lepicard, S.; Livolsi, P.; Lochard, J.; Schneider, T.; Ollagnon, H.; Pupin, V.; Heriard-Dubreuil, G.; Girard, P.; Guyonnet, J.F.; Le Cardinal, G.; Monroy, M.; Pena-Vega, A.; Rigby, J.

    1999-01-01

    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.)

  18. Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Yano, Y., E-mail: yano.yasuhide@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki, 311-1393 (Japan); Tanno, T.; Oka, H.; Ohtsuka, S.; Inoue, T.; Kato, S.; Furukawa, T.; Uwaba, T.; Kaito, T. [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki, 311-1393 (Japan); Ukai, S.; Oono, N. [Materials Science and Engineering, Faculty of Engineering, Hokkaido University, N13, W-8, Kita-ku, Sapporo, Hokkaido, 060-8628 (Japan); Kimura, A. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hayashi, S. [Tokyo Institute of Technology, 2-12-1, Ookayama, Meguro-ku, Tokyo 152-8550 (Japan); Torimaru, T. [Nippon Nuclear Fuel Development Co., Ltd., 2163, Narita-cho, Oarai-machi, Ibaraki, 311-1313 (Japan)

    2017-04-15

    Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900–1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

  19. Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

    Science.gov (United States)

    Yano, Y.; Tanno, T.; Oka, H.; Ohtsuka, S.; Inoue, T.; Kato, S.; Furukawa, T.; Uwaba, T.; Kaito, T.; Ukai, S.; Oono, N.; Kimura, A.; Hayashi, S.; Torimaru, T.

    2017-04-01

    Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900-1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

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

    International Nuclear Information System (INIS)

    Lewis, B.J.; Corse, B.J.; Thompson, W.T.; Kaye, M.H.; Iglesias, F.C.; Elder, P.; Dickson, R.; Liu, Z.

    1997-01-01

    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)

  1. Nuclear Fuel Behaviour in Loss-of-coolant Accident (LOCA) Conditions

    International Nuclear Information System (INIS)

    Pettersson, Kjell; Chung, Haijung; ); Billone, Michael; Fuketa, Toyoshi; Nagase, Fumihisa; Grandjean, Claude; Hache, George; Papin, Joelle; Heins, Lothar; Hozer, Zoltan; In de Betou, Jan; Kelppe, Seppo; Mayer, Ralph; Scott, Harold; Voglewede, John; Sonnenburg, Heinz; Sunder, Sham; Valach, Mojmir; Vrtilkova, Vera; Waeckel, Nicolas; Wiesenack, Wolfgang; Zimmermann, Martin

    2009-01-01

    The NEA Working Group on Fuel Safety (WGFS) is tasked with advancing the current understanding of fuel safety issues by assessing the technical basis for current safety criteria and their applicability to high burn-up and to new fuel designs and materials. The group aims at facilitating international convergence in this area, including as regards experimental approaches and interpretation and the use of experimental data relevant for safety. In 1986, a working group of the NEA Committee on the Safety of Nuclear Installations (CSNI) issued a state-of-the-art report on water reactor fuel behaviour in design-basis accident (DBA) conditions. The 1986 report was limited to the oxidation, embrittlement and deformation of pressurised water reactor (PWR) fuel in a loss-of-coolant accident (LOCA). Since then, considerable experimental and analytical work has been performed, which has led to a broader and deeper understanding of LOCA-related phenomena. Further, new cladding alloys have been produced, which might behave differently than the previously used Zircaloy-4, both under normal operating conditions and during transients. Compared with 20 years ago, fuel burn-up has been significantly increased, which requires extending the LOCA database in order to cover the high burnup range. There was also a clear need to address LOCA performance for reactor types other than PWRs. The present report has been prepared by the WGFS and covers the following technical aspects: - Description of different LOCA scenarios for major types of reactors: BWRs, PWRs, VVERs and to a lesser extent CANDUs. - LOCA phenomena: ballooning, burst, oxidation, fuel relocation and possible fracture at quench. - Details of high-temperature oxidation behaviour of various cladding materials. - Metallurgical phase change, effect of hydrogen and oxygen on residual cladding ductility. - Methods for LOCA testing, for example two-sided oxidation and ring compression for ductility, and integral quench test for

  2. Influence of initial conditions on rod behaviour during boiling crisis phase following a reactivity initiated accident

    International Nuclear Information System (INIS)

    Georgenthum, V.; Sugiyama, T.

    2010-01-01

    In the frame of their research programs on high burn-up fuel safety, the French Institute for Radioprotection and Nuclear Safety (IRSN) and the Japan Atomic Energy Agency (JAEA) performed a large set of tests devoted to the study of PWR fuel rod behavior during Reactivity Initiated Accident (RIA) respectively in the CABRI reactor and in the NSRR reactor. The reactor test conditions are different in terms of coolant nature, temperature and pressure. In the CABRI reactor, tests were performed until now with sodium coolant at 280 Celsius degrees and 3 bar. In the NSRR reactor most of the tests were performed with stagnant water at 20 C. degrees and atmospheric pressure but recently a new high temperature high pressure capsule has been developed which allows to performed tests at up to 280 Celsius degrees and 70 bar. The paper discusses the influence of test conditions on rod behaviour during boiling phase, based on tests results and SCANAIR code calculations. The study shows that when the boiling crisis is reached, the initial inner and outer rod pressure have an essential impact on the clad straining and possible ballooning. The analysis of the different test conditions makes it possible to discriminate the influence of initial conditions on the different phases of the transient and is useful for modelling and code development. (authors)

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

    International Nuclear Information System (INIS)

    Zaiss, W.

    1999-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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. Development of stable walking robot for accident condition monitoring on uneven floors in a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Seog; Jang, You Hyun [Central Research Institute of Korea Hydro and Nuclear Power Company, Daejeon (Korea, Republic of)

    2017-04-15

    Even though the potential for an accident in nuclear power plants is very low, multiple emergency plans are necessary because the impact of such an accident to the public is enormous. One of these emergency plans involves a robotic system for investigating accidents under conditions of high radiation and contaminated air. To develop a robot suitable for operation in a nuclear power plant, we focused on eliminating the three major obstacles that challenge robots in such conditions: the disconnection of radio communication, falling on uneven floors, and loss of localization. To solve the radio problem, a Wi-Fi extender was used in radio shadow areas. To reinforce the walking, we developed two- and four-leg convertible walking, a floor adaptive foot, a roly-poly defensive falling design, and automatic standing recovery after falling methods were developed. To allow the robot to determine its location in the containment building, a bar code landmark reading method was chosen. When a severe accident occurs, this robot will be useful for accident condition monitoring. We also anticipate the robot can serve as a workman aid in a high radiation area during normal operations.

  6. Modelling of RPV lower head under core melt severe accident condition using OpenFOAM

    International Nuclear Information System (INIS)

    Madokoro, Hiroshi; Kretzschmar, Frank; Miassoedov, Alexei

    2017-01-01

    Although six years have been passed since the tragic severe accident at Fukushima Daiichi, still large uncertainties exist in modeling of core degradation and reactor pressure vessel (RPV) failure. It is extremely important to obtain a better understanding of complex phenomena in the lower head in order to improve accident management measures. The possible failure mode of reactor pressure vessel and its failure time are especially a matter of importance. Thermal behavior of the molten pool can be simulated by the Phase-change Effective Convectivity Model (PECM), which is a distributed-parameter model developed in the Royal Institute of Technology (KTH), Sweden. The model calculates convective currents not using a pure CFD approach but based on so called “characteristic velocities” that are determined by empirical correlations depending on the geometry and physical properties of the molten pool. At the Karlsruhe Institute of Technology (KIT), the PECM has been implemented in the open-source CFD software OpenFOAM in order to receive detailed predictions of a core melt behavior in the RPV lower head under severe accident conditions. An advantage of using OpenFOAM is that it is very flexible to add and modify models and physical properties. In the current work, the solver is extended to couple PECM with a structure analysis model of the vessel wall. The model considers thermal expansion, plasticity, creep and damage. The model and physical properties are based on those implemented in ANSYS. Although the previous implementation had restriction that the amount of and geometry of the melt cannot be changed, our coupled model allows flexibility of the melt amount and geometry. The extended solver was used to simulate the LIVE-L1 and -L7V experiments and has demonstrated good prediction of the temperature distribution in the molten pool and heat flux distribution through the vessel wall. Regarding the vessel failure the model was applied to one of the FOREVER tests

  7. Collaborative testing for key-term definitions under representative conditions: Efficiency costs and no learning benefits.

    Science.gov (United States)

    Wissman, Kathryn T; Rawson, Katherine A

    2018-01-01

    Students are expected to learn key-term definitions across many different grade levels and academic disciplines. Thus, investigating ways to promote understanding of key-term definitions is of critical importance for applied purposes. A recent survey showed that learners report engaging in collaborative practice testing when learning key-term definitions, with outcomes also shedding light on the way in which learners report engaging in collaborative testing in real-world contexts (Wissman & Rawson, 2016, Memory, 24, 223-239). However, no research has directly explored the effectiveness of engaging in collaborative testing under representative conditions. Accordingly, the current research evaluates the costs (with respect to efficiency) and the benefits (with respect to learning) of collaborative testing for key-term definitions under representative conditions. In three experiments (ns = 94, 74, 95), learners individually studied key-term definitions and then completed retrieval practice, which occurred either individually or collaboratively (in dyads). Two days later, all learners completed a final individual test. Results from Experiments 1-2 showed a cost (with respect to efficiency) and no benefit (with respect to learning) of engaging in collaborative testing for key-term definitions. Experiment 3 evaluated a theoretical explanation for why collaborative benefits do not emerge under representative conditions. Collectively, outcomes indicate that collaborative testing versus individual testing is less effective and less efficient when learning key-term definitions under representative conditions.

  8. Effect of air condition on AP-1000 containment cooling performance in station black out accident

    International Nuclear Information System (INIS)

    Hendro Tjahjono

    2015-01-01

    AP1000 reactor is a nuclear power plant generation III+ 1000 MWe which apply passive cooling concept to anticipate accidents triggered by the extinction of the entire supply of electrical power or Station Black Out (SBO). In the AP1000 reactor, decay heat disposal mechanism conducted passively through the PRHR-IRWST and subsequently forwarded to the reactor containment. Containment externally cooled through natural convection in the air gap and through evaporation cooling water poured on the outer surface of the containment wall. The mechanism of evaporation of water into the air outside is strongly influenced by the conditions of humidity and air temperature. The purpose of this study was to determine the extent of the influence of the air condition on cooling capabilities of the AP1000 containment. The method used is to perform simulations using Matlab-based analytical calculation model capable of estimating the power of heat transferred. The simulation results showed a decrease in power up to 5% for relative humidity rose from 10% to 95%, while the variation of air temperature of 10°C to 40°C, the power will decrease up to 15%. It can be concluded that the effect of air temperature increase is much more significant in lowering the containment cooling ability compared with the increase of humidity. (author)

  9. Post-test investigation result on the WWER-1000 fuel tested under severe accident conditions

    International Nuclear Information System (INIS)

    Goryachev, A.; Shtuckert, Yu.; Zwir, E.; Stupina, L.

    1996-01-01

    The model bundle of WWER-type were tested under SFD condition in the out-of-pile CORA installation. The objective of the test was to provide an information on the WWER-type fuel bundles behaviour under severe fuel damage accident conditions. Also it was assumed to compare the WWER-type bundle damage mechanisms with these experienced in the PWR-type bundle tests with aim to confirm a possibility to use the various code systems, worked our for PWR as applied to WWER. In order to ensure the possibility of the comparison of the calculated core degradation parameters with the real state of the tested bundle, some parameters have been measured on the bundle cross-sections under examination. Quantitative parameters of the bundle degradation have been evaluated by digital image processing of the bundle cross-sections. The obtained results are shown together with corresponding results obtained by the other participants of this investigation. (author). 3 refs, 13 figs

  10. SSYST. A code system to analyze LWR fuel rod behavior under accident conditions

    International Nuclear Information System (INIS)

    Gulden, W.; Meyder, R.; Borgwaldt, H.

    1982-01-01

    SSYST (Safety SYSTem) is a modular system to analyze the behavior of light water reactor fuel rods and fuel rod simulators under accident conditions. It has been developed in close cooperation between Kernforschungszentrum Karlsruhe (KfK) and the Institut fuer Kerntechnik und Energiewandlung (IKE), University Stuttgart, under contract of Projekt Nukleare Sicherheit (PNS) at KfK. Although originally aimed at single rod analysis, features are available to calculate effects such as blockage ratios of bundles and wholes cores. A number of inpile and out-of-pile experiments were used to assess the system. Main differences versus codes like FRAP-T with similar applications are (1) an open-ended modular code organisation, (2) availability of modules of different sophistication levels for the same physical processes, and (3) a preference for simple models, wherever possible. The first feature makes SSYST a very flexible tool, easily adapted to changing requirements; the second enables the user to select computational models adequate to the significance of the physical process. This leads together with the third feature to short execution times. The analysis of transient rod behavior under LOCA boundary conditions e.g. takes 2 mins cpu-time (IBM-3033), so that extensive parametric studies become possible

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

    International Nuclear Information System (INIS)

    1987-09-01

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

    International Nuclear Information System (INIS)

    Hishida, Makoto

    1996-01-01

    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.)

  13. A study on gap heat transfer of LWR fuel rods under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Fujishiro, Toshio

    1984-03-01

    Gap heat transfer between fuel pellet and cladding have a large influence on the LWR fuel behaviors under reactivity initiated accident (RIA) conditions. The objective of the present study is to investigate the effects of gap heat transfer on RIA fuel behaviors based on the results of the gap-gas parameter tests in NSRR and on their analysis with NSR-77 code. Through this study, transient variations of gap heat transfer, the effects of the gap heat transfer on fuel thermal behaviors and on fuel failure, effects of pellet-cladding sticking by eutectic formation, and the effects of cladding collapse under high external pressure have been clearified. The studies have also been performed on the applicability and its limit of modified Ross and Stoute equation which is extensively utilized to evaluate the gap heat transfer coefficient in the present fuel behavior codes. The method to evaluate the gap conductance to the conditions beyond the applicability limit of the Ross and Stoute equation has also been proposed. (author)

  14. Radioactive particulate release associated with the DOT specification 6M container under hypothetical accident conditions

    International Nuclear Information System (INIS)

    Taylor, J.M.; Raney, P.J.

    1986-02-01

    A testing program was conducted to determine the leakage of depleted uranium dioxide powder (DUO) from the inner containment components of the US Department of Transportation's (DOT) specification 6M container under hypothetical accident conditions. Depleted uranium dioxide was selected as a surrogate for plutonium oxide because of the similarities in the powder characteristics, density and particle size, and because of the special handling and special facilities required for plutonium oxide. The DUO was packaged inside food pack cans in three different configurations inside the 2R vessel of the 6M container. The amount of DUO powder leakage ranged from none detectable ( -7 g) to a high of 1 x 10 -3 g. The combination of gravity, vibration and pressure produced the highest leakage of DUO. Containers that had hermetic seals (leak rates -4 atm cc/min) did not leak any detectable amount ( -7 g) of DUO under the test conditions. Impact forces had no effect on the leakage of particles with the packaging configurations used. 23 refs., 24 figs., 3 tabs

  15. Development of systematic models for aerosol agglomeration and spray removal under severe accident conditions

    International Nuclear Information System (INIS)

    Kajimoto, Mitsuhiro

    2008-01-01

    Radionuclide behavior during various severe accident conditions has been addressed as one of the important issues to discuss environmental safety in nuclear power plants. The present paper deals with the development of analytical models and their validations for the agglomeration of multiple-component aerosol and spray removal that controls source terms to the environment of both aerosols and gaseous radionuclides during recirculation mode operation in a containment system for a light water reactor. As for aerosol agglomeration, the single collision kernel model that can cover all types of two-body collision of aerosol was developed. In addition, the dynamic model that can treat aerosol and vapor transfer leading to the equilibrium condition under the containment spray operation was developed. The validations of the present models for multiple-component aerosol growth by agglomeration were performed by comparisons with Nuclear Safety Pilot Plant (NSPP) experiments at Oak Ridge National Laboratory (ORNL) and AB experiments at Hanford Engineering National Laboratory (HEDL). In addition, the spray removal models were applied to the analysis of containment spray experiment (CSE) at HEDL. The results calculated by the models showed good agreements with experimental results. (author)

  16. Conditions for oxygen-deficient combustion during accidents with severe core concrete thermal attack

    International Nuclear Information System (INIS)

    Luangdilok, W.; Elicson, G.T.; Berger, W.E. Jr.

    1993-01-01

    This paper addresses the interactions between MCCI (molten core-concrete interactions)-induced offgas releases, mostly the combustible gases, natural circulation between the cavity and the lower containment based on recent research developments in the area of mixed convection flow (Epstein, et al., 1989; Epstein, 1988; Epstein, 1992) between compartments, and their effects on combustion in PWR containments during prolonged severe accidents. Specifically, large dry PWR containments undergoing severe core-concrete attack during station blackouts where the containment atmosphere is expected to be inerted are objects of this analysis. The purpose of this paper, given the conditions that oxygen can be brought to the cavity, is to demonstrate that consumption of most oxygen present in the containment can be achieved in a reasonable time scale assuming that combustion is not subject to flammability limits due to the high cavity temperatures. The conditions for cavity combustion depend on several factors including good gas flowpaths between the cavity and other containment regions, and combustion processes within the cavity with the hot debris acting as the ignition source

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

    International Nuclear Information System (INIS)

    Carlin, F.; Alba, C.; Chenion, J.; Gaussens, G.; Henry, J.Y.

    1986-10-01

    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 [fr

  18. The behavior of a container for UF6 under accident conditions

    International Nuclear Information System (INIS)

    Andreuccetti, P.; Aquaro, D.; Forasassi, G.

    1987-01-01

    Transport of uranium hexafluoride during the different phases of the fuel cycle is carried out using containers of various types that must meet the safety requirements provided for in the specific international regulations for this area. Qualification of the behavior of the 30B cylinder and its respective overpack under reference accident conditions for the purpose of design and utilization of such containers is currently a subject of interest on an international level, since it is being widely used in a number of countries. To contribute to this qualification process, a relatively complex research program was defined and developed, including, among other things, drop tests from 9 m on to an unyielding target, drop tests from a height of 1 m on to a cylindrical bar, and thermal tests in a furnace, all of which were carried out on two complete specimens of the same container with a simulated load. For analysis of the damage a series of leak tests and a water immersion test were developed to analyze the damage to the two specimens mentioned above and to a container of reduced dimensions designed for this purpose and equipped to reproduce conditions similar to the real conditions inside the container under investigation. Evaluation of the heat exchange conditions that could exist in the container given real contents of uranium hexafluoride was also conducted using a series of calculations carried out with the computer code TRUMP. The results of the different types of experiments and calculations performed and presented in detail in the present study have made it possible to draw useful conclusions for practical evaluation of the reliability of the container under investigation, also in view of the intended goal of container qualification as per the existing regulations for transport of radioactive material. 21 refs., 44 figs., 3 tabs

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

    International Nuclear Information System (INIS)

    2000-03-01

    uncertainty in calculating 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-H 2 -H 2 O-CO 2 -air mixtures, resulting in significant uncertainties in predicted pressure loads from ignition. - Experimental data on the detonation cell sizes (detonability) of CO-H 2 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 CO 2 reduces the detonation sensitivity of CO-H 2 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-H 2 mixtures are lacking

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

  1. MDEP Common Position CP-EPRWG-04. Common position on EPR containment heat removal system in accident conditions

    International Nuclear Information System (INIS)

    2015-01-01

    The importance of the integrity of the containment as a fundamental barrier to protect the people and environment against the effects of a nuclear accident is well established. In this regard, an essential objective is that the necessity for off-site counter-measures to reduce radiological consequences be limited or even eliminated. The design should provide engineering means to address those sequences which would otherwise lead to large or early releases, even in case of severe external hazards. The plant shall be designed so that it can be brought into a controlled and stable state and the containment function can be maintained, under accident conditions in which there is a significant amount of radioactive material in the containment, i.e. resulting from severe degradation of the reactor core. It is expected that due consideration to these requirements is to be given while tailoring long term loss of electrical power mitigation strategies. In order to reliably maintain the containment barrier, the regulators believe that: - safety features specifically designed for fulfilling safety functions required in core melt accidents shall be independent to the extent reasonably practicable from the Systems, Structures and Components (SSC) of the other levels of defense; - safety features specifically designed for fulfilling safety functions required in core melt accidents shall be safety classified and adequately qualified for the core melt accident environmental conditions for the time frame for which they are required to operate. In the light of the Fukushima Daiichi accident, the regulators believe that those safety features shall be designed with an adequate margin as compared to the levels of natural hazards considered for the site hazard evaluation; - the systems and components necessary for ensuring the containment function in a core melt accident shall have reliability commensurate with the function that they are required to fulfil. This may require redundancy of

  2. Behavior of irradiated ATR/MOX fuel under reactivity initiated accident conditions (Joint research)

    International Nuclear Information System (INIS)

    Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Uetsuka, Hiroshi

    2000-03-01

    Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20 MWd/kgHM were conducted at the NSRR in JAERI to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Deformation of the fuel rods due to PCMI occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet. (author)

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

    International Nuclear Information System (INIS)

    Kazuyuki, Kusagaya; Takehiko, Nakamura; Makio, Yoshinaga; Hiroshi, Akie; Toshiyuki, Yamashita; Hiroshi, Uetsuka

    2002-01-01

    Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl 2 O 4 ) 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/m 3 , which was comparable to that of un-irradiated UO 2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO 2 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/m 3 . 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)

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

    International Nuclear Information System (INIS)

    Alonso, A.; Bolado, R.; Hontanon, E.

    1991-07-01

    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

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

    International Nuclear Information System (INIS)

    Krieg, R.; Eberle, F.; Goeller, B.; Gulden, W.; Kadlec, J.; Messemer, G.; Mueller, S.; Wolf, E.

    1983-10-01

    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.) [de

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

    International Nuclear Information System (INIS)

    Seol, Wookcheol; Park, Jongwoon

    2014-01-01

    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

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

    International Nuclear Information System (INIS)

    Kljenak, I.; Mavko, B.

    2007-01-01

    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)

  8. Simulation of KAEVER experiments on aerosol behavior in a nuclear power plant containment at accident conditions with the ASTEC code

    International Nuclear Information System (INIS)

    Kljenak, I.; Mavko, B.

    2006-01-01

    Experiments on aerosol behaviour in saturated and non-saturated atmosphere, which were performed in the KAEVER experimental facility, were simulated with the severe accident computer code ASTEC CPA V1.2. The specific purpose of the work was to assess the capability of the code to model aerosol condensation and deposition in the containment of a light-water-reactor nuclear power plant at severe accident conditions, if the atmosphere saturation conditions are simulated adequately. Five different tests were first simulated with boundary conditions, obtained from the experiments. In all five tests, a non-saturated atmosphere was simulated, although, in four tests, the atmosphere was allegedly saturated. The simulations were repeated with modified boundary conditions, to obtain a saturated atmosphere in all tests. Results of dry and wet aerosol concentrations in the test vessel atmosphere for both sets of simulations are compared to experimental results. (author)

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

    International Nuclear Information System (INIS)

    Jacobus, M.J.

    1992-11-01

    This report describes the results of aging, condition monitoring, and accident testing of miscellaneous cable types. 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, properly installed, most of the various miscellaneous cable products tested should be able to survive an accident after 60 years for total aging doses of at least 150 kGy or higher (depending on the material) and for moderate ambient temperatures on the order of 45--55 degrees C (potentially higher or lower, depending on material specific activtion energies and total radiation doses). Mechanical measurements (primarily elongation, modulus, and density) were more effective than electrical measurements for monitoring age-related degradation

  10. Research activities at JAERI on core material behaviour under severe accident conditions

    International Nuclear Information System (INIS)

    Uetsuka, H.; Katanashi, S.; Ishijima, K.

    1996-01-01

    At the Japan Atomic Energy Research Institute (JAERI), experimental studies on physical phenomena under the condition of a severe accident have been conducted. This paper presents the progress of the experimental studies on fuel and core materials behaviour such as the thermal shock fracture of fuel cladding due to quenching, the chemical interaction of core materials at high temperatures and the examination of TMI-2 debris. The mechanical behaviour of fuel rod with heavily embrittled cladding tube due to the thermal shock during delayed reflooding have been investigated at the Nuclear Safety Research Reactor (NSSR) of JAERI. A test fuel rod was heated in steam atmosphere by both electric and nuclear heating using the NSSR, then the rod was quenched by reflooding at the test section. Melting of core component materials having relatively low melting points and their eutectic reaction with other materials significantly influence on the degradation and melt down of fuel bundles during severe accidents. Therefore basic information on the reaction of core materials is necessary to understand and analyze the progress of core melting and relocation. Chemical interactions have been widely investigated at high temperatures for various binary systems of core component materials including absorber materials such as Zircaloy/Inconel, Zircaloy/stainless steel, Zircaloy/(Ag-In-Cd), stainless steel B 4 C and Zircaloy/B 4 C. It was found that the reaction generally obeyed a parabolic rate law and the reaction rate was determined for each reaction system. Many debris samples obtained from the degraded core of TMI-2 were transported to JAERI for numerous examinations and analyses. The microstructural examination revealed that the most part of debris was ceramic and it was not homogeneous in a microscopic sense. The thermal diffusivity data was also obtained for the temperature range up to about 1800K. The data from the large scale integral experiments were also obtained through the

  11. Basic study on PWR plant behavior under the condition of severe accident (1)

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Ida, Shohma; Nakamura, Shinya

    2015-01-01

    In this paper, we report on the results using the PWR plant simulator about the plant behavior under the condition of the severe accident that LOCA occurs but ECCS fails the water irrigation into the reactor core. As for the results about the relationship between the LOCA area and the time from LOCA occurs until fuel temperature rise start, the time became shorter as the area was the larger. But, in LOCA area of 1000 cm 2 or more large, the time was almost constant regardless of the area. For small LOCA of 25 cm 2 area, from the results of the comparative experiments for RCS natural circulation cooling effect in the case of SG open or not, in SG open condition compared with SG not open, the effect was observed more, but the reactor water level was greatly reduced and the time until the fuel temperature rise start was shortened, so the fuel temperature at the time of water irrigation into the reactor core has become higher. On the other hand, for the large LOCA of 1000 cm 2 , the effect was not observed regardless of SG open or not. In addition, the reactor core damage was not spared in the irrigation into the reactor core after 30 minutes from LOCA, however, the hydrogen concentration in the containment building is less than the lower limit of hydrogen detonation, and also the pressure in the containment building is less than the designed value. That is, although suffered the core damage, the integrity of the containment building has been shown to be secured. (author)

  12. High burnup (41 - 61 GWd/tU) BWR fuel behavior under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Kusagaya, Kazuyuki; Yoshinaga, Makio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    High burnup boiling water reactor (BWR) fuel was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity initiated accident (RIA) conditions. Temperature, deformation, failure, and fission gas release behavior under the simulated RIA condition was studied in the tests. Fuel failure due to pellet-cladding mechanical interaction (PCMI) did not occur in the tests with typical domestic BWR fuel at burnups up to 56 GWd/tU, because they had limited cladding embrittlement due to hydrogen absorption of about 100 ppm or less. However, the cladding failure occurred in tests with fuel at a burnup of 61 GWd/tU, in which the peak hydrogen content in the cladding was above 150 ppm. This type of failure was observed for the first time in BWR fuels. The cladding failure occurred at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g), which were higher than the PCMI failure thresholds decided by the Japanese Nuclear Safety Commission. From post-test examinations of the failed fuel, it was found that the crack in the BWR cladding progressed in a manner different from the one in PWR cladding failed in earlier tests, owing to its more randomly oriented hydride distribution. Because of these differences, the BWR fuel was judged to have failed at hydrogen contents lower than those of the PWR fuel. Comparison of the test results with code calculations revealed that the PCMI failure was caused by thermal expansion of pellets, rather than by the fission gas expansion in the pellets. The gas expansion, however, was found to cause large cladding hoop deformation later after the cladding temperature escalated. (author)

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

    International Nuclear Information System (INIS)

    Rogers, J.T.

    1984-06-01

    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

  14. Study of labor accidents in the rural environment: analysis of processes and conditions of work

    OpenAIRE

    Thaís Alves Brito; Cleber Souza de Jesus

    2009-01-01

    The modernization of agriculture, that broadenned the mechanization of farming and the agrotoxic use, potentially increased some risks of accidents. The agriculture workers and cattle raising are constantly exposed to several physical, chemical and biological agents, like machine, implements, handly tools, agrotoxics, ectoparaziticides, domestic animals and poisonous animals, which can to bring accidents. The aiming the importance of this working class to economic developing of country, this ...

  15. Simulation of the transient processes of load rejection under different accident conditions in a hydroelectric generating set

    Science.gov (United States)

    Guo, W. C.; Yang, J. D.; Chen, J. P.; Peng, Z. Y.; Zhang, Y.; Chen, C. C.

    2016-11-01

    Load rejection test is one of the essential tests that carried out before the hydroelectric generating set is put into operation formally. The test aims at inspecting the rationality of the design of the water diversion and power generation system of hydropower station, reliability of the equipment of generating set and the dynamic characteristics of hydroturbine governing system. Proceeding from different accident conditions of hydroelectric generating set, this paper presents the transient processes of load rejection corresponding to different accident conditions, and elaborates the characteristics of different types of load rejection. Then the numerical simulation method of different types of load rejection is established. An engineering project is calculated to verify the validity of the method. Finally, based on the numerical simulation results, the relationship among the different types of load rejection and their functions on the design of hydropower station and the operation of load rejection test are pointed out. The results indicate that: The load rejection caused by the accident within the hydroelectric generating set is realized by emergency distributing valve, and it is the basis of the optimization for the closing law of guide vane and the calculation of regulation and guarantee. The load rejection caused by the accident outside the hydroelectric generating set is realized by the governor. It is the most efficient measure to inspect the dynamic characteristics of hydro-turbine governing system, and its closure rate of guide vane set in the governor depends on the optimization result in the former type load rejection.

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

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    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.

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

    International Nuclear Information System (INIS)

    Verfondern, K.; Mueller, D.

    1991-01-01

    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

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

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Chen Dianshan; Takeuchi, Kiyoshi.

    1992-01-01

    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.)

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

    International Nuclear Information System (INIS)

    Anderson, M.H.; Corradini, M.L.

    1998-01-01

    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.)

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

    International Nuclear Information System (INIS)

    Purcell, P.C.; Dallongeville, M.

    2004-01-01

    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

  1. 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

    International Nuclear Information System (INIS)

    Araiza M, E.; Nunez C, A.

    2001-01-01

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

  2. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    Energy Technology Data Exchange (ETDEWEB)

    Bakalov, Ivan [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Berlin (Germany); Sonnenkalb, Martin [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Koeln (Germany)

    2018-02-15

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

  3. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    International Nuclear Information System (INIS)

    Bakalov, Ivan; Sonnenkalb, Martin

    2018-01-01

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

  4. Diagnostic and prognostic system for identification of accident scenarios and prediction of 'source term' in nuclear power plants under accident conditions

    International Nuclear Information System (INIS)

    Santhosh; Gera, B.; Kumar, Mithilesh

    2014-01-01

    Nuclear power plant experiences a number of transients during its operations. These transients may be due to equipment failure, malfunctioning of process support systems etc. In such a situation, the plant may result in an abnormal state which is undesired. In case of such an undesired plant condition, the operator has to carry out diagnostic and corrective actions. When an event occurs starting from the steady state operation, instruments' readings develop a time dependent pattern and these patterns are unique with respect to the type of the particular event. Therefore, by properly selecting the plant process parameters, the transients can be distinguished. In this connection, a computer based tool known as Diagnostic and Prognostic System has been developed for identification of large pipe break scenarios in 220 MWe Pressurised Heavy Water Reactors (PHWRs) and for prediction of expected 'Source Term' and consequence for a situation where Emergency Core Cooling System (ECCS) is not available or partially available. Diagnostic and Prognostic System is essentially a transient identification and expected source term forecasting system. The system is based on Artificial Neural Networks (ANNs) that continuously monitors the plant conditions and identifies a Loss Of Coolant Accident (LOCA) scenario quickly based on the reactor process parameter values. The system further identifies the availability of injection of ECCS and in case non-availability of ECCS, it can forecast expected 'Source Term'. The system is a support to plant operators as well as for emergency preparedness. The ANN is trained with a process parameter database pertaining to accident conditions and tested against blind exercises. In order to see the feasibility of implementing in the plant for real-time diagnosis, this system has been set up on a high speed computing facility and has been demonstrated successfully for LOCA scenarios. (author)

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

    Directory of Open Access Journals (Sweden)

    Kil-Mo Koo

    2012-01-01

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

  6. Evaluation of High-Pressure RCS Natural Circulations Under Severe Accident Conditions

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Bang, Young Suk; Suh, Nam Duk

    2006-01-01

    Since TMI-2 accident, the occurrence of severe accident natural circulations inside RCS during entire in-vessel core melt progressions before the reactor vessel breach had been emphasized and tried to clarify its thermal-hydraulic characteristics. As one of consolidated outcomes of these efforts, sophisticated models have been presented to explain the effects of a variety of engineering and phenomenological factors involved during severe accident mitigation on the integrity of RCS pressure boundaries, i.e. reactor pressure vessel(RPV), RCS coolant pipe and steam generator tubes. In general, natural circulation occurs due to density differences, which for single phase flow, is typically generated by temperature differences. Three natural circulation flows can be formed during severe accidents: in-vessel, hot leg countercurrent flow and flow through the coolant loops. Each of these flows may be present during high-pressure transients such as station blackout (SBO) and total loss of feedwater (TLOFW). As a part of research works in order to contribute on the completeness of severe accident management guidance (SAMG) in domestic plants by quantitatively assessing the RCS natural circulations on its integrity, this study presents basic approach for this work and some preliminary results of these efforts with development of appropriately detailed RCS model using MELCOR computer code

  7. Early results from an experimental program to determine the behavior of containment piping penetration bellows subjected to severe accident conditions

    International Nuclear Information System (INIS)

    Lambert, L.D.; Parks, M.B.

    1994-01-01

    Containment piping penetration bellows are an integral part of the pressure boundary in steel containments in the United States (US). Their purpose is to minimize loading on the containment shell caused by differential movement between the piping and the containment. This differential movement is typically caused by thermal gradients generated during startup and shutdown of the reactor, but can be caused by earthquake, a loss-of-coolant accident (LOCA), or ''severe'' accidents. In the event of a severe accident, the bellows would be subjected to pressure, temperature, and deflection well beyond the design basis. Most bellows are installed such that they would be subjected to elevated internal pressure, elevated temperature, axial compression, and lateral deflection during a severe accident. A few bellows would be subjected to external pressure and axial elongation, as well as elevated temperature and lateral deflection. The purpose of this experimental program is to examine the potential for leakage of containment bellows during a severe accident. The test series subjects bellows to various levels and combinations of internal pressure, elevated temperature, axial compression or elongation, and lateral deformation. The experiments are being conducted in two parts. For Part 1, all bellows specimens are tested in ''like-new'' condition, without regard for the possible degrading effect of corrosion that has been observed in some containment piping bellows in the US Part I testing, which included 13 bellows tests, has been completed. The second part of the experimental program, in which bellows are subjected to simulated corrosive environments prior to testing, has just just begun. The Part I experiments have shown that bellows in ''like-new'' condition can withstand elevated temperatures and pressures along with large deformations before leaking. In most cases, the like-new bellows were fully compressed without developing any leakage

  8. Accident information needs

    International Nuclear Information System (INIS)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information

  9. Accident information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-12-31

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

  10. Accident information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

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

    Science.gov (United States)

    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. Copyright © 2014 Elsevier Ltd. All rights reserved.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  14. Fast dose assessment models, parameters and code under accident conditions for Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Zhang, Z.Y.; Hu, E.B.; Meng, X.C.; Zhang, Y.; Yao, R.T.

    1993-01-01

    According to requirement of accident emergency plan for Qinshan Nuclear Power Plant, a Gaussian straight-line model was adopted for estimating radionuclide concentration in surface air. In addition, the effects of mountain body on atmospheric dispersion was considered. By combination of field atmospheric dispersion experiment and wind tunnel modeling test, necessary modifications have been done for some models and parameters. A computer code for assessment was written in Quick BASIC (V4.5) language. The radius of assessment region is 10 km and the code is applicable to early accident assessment. (1 tab.)

  15. Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

    1990-05-01

    The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs

  16. The Influence of atmospheric conditions to probabilistic calculation of impact of radiology accident on PWR 1000 MWe

    International Nuclear Information System (INIS)

    Pande Made Udiyani; Sri Kuntjoro

    2015-01-01

    The calculation of the radiological impact of the fission products releases due to potential accidents that may occur in the PWR (Pressurized Water Reactor) is required in a probabilistic. The atmospheric conditions greatly contribute to the dispersion of radionuclides in the environment, so that in this study will be analyzed the influence of atmospheric conditions on probabilistic calculation of the reactor accidents consequences. The objective of this study is to conduct an analysis of the influence of atmospheric conditions based on meteorological input data models on the radiological consequences of PWR 1000 MWe accidents. Simulations using PC-Cosyma code with probabilistic calculations mode, the meteorological data input executed cyclic and stratified, the meteorological input data are executed in the cyclic and stratified, and simulated in Muria Peninsula and Serang Coastal. Meteorological data were taken every hour for the duration of the year. The result showed that the cumulative frequency for the same input models for Serang coastal is higher than the Muria Peninsula. For the same site, cumulative frequency on cyclic input models is higher than stratified models. The cyclic models provide flexibility in determining the level of accuracy of calculations and do not require reference data compared to stratified models. The use of cyclic and stratified models involving large amounts of data and calculation repetition will improve the accuracy of statistical calculation values. (author)

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

    International Nuclear Information System (INIS)

    Bowsher, B.R.

    1985-09-01

    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)

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

    International Nuclear Information System (INIS)

    Abouhadra, D.S.; Byrne, J.E.

    2004-01-01

    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

  19. Basic study on BWR plant behavior under the condition of severe accident

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Jyohko, Shingo; Dohgo, Hirofumi

    2015-01-01

    In this paper, we report on the results using the BWR plant simulator about the plant behavior under the condition of the severe accident that LOCA occurs but ECCS fails the water irrigation into the reactor core. The simulation experiments were carried out for the cases that LOCA has occurred in the main steam piping or in the recirculation piping, respectively. As for the results about the relationship between the LOCA area and the time from LOCA occurs until the fuel temperature rise start, the effect that RCIC operated was extremely big for LOCA area of up to 100 cm"2 for both type LOCA. In the case of main steam system LOCA, the core water level suddenly decreased for large LOCA of 2000 cm"2 area, however, if the irrigation into the reactor core was carried out 30 min after LOCA occurrence, the core had little damage. In addition, the H_2 concentration in the containment vessel did not exceed both limits of H_2 explosion nor detonation. The pressure of the containment vessel was around 3 kg/cm"2 of design value, so the soundness of the containment vessel was confirmed. On the other hand, for the recirculation system LOCA of 2000 cm"2 area, a drop of the core water level was extremely in comparison with main steam system LOCA, and the fuel assemblies were completely exposed during up to 30 min, to the irrigation from approximately 100 sec, after LOCA occurrence. Therefore, the fuel temperature during the irrigation had reached approximately 1900degC. Thus, the fuel cladding were damaged approximately less than 10%, and H_2 concentration in the containment vessel was approximately 9% which did not exceed H_2 detonation limit of 13% but exceeded H_2 explosion limit of 4%. However, the containment vessel internal pressure was settled around design pressure value of containment vessel. As the results, some core damage could not be avoided, but soundness of the containment vessel, which should take the role of 'confine', was found to be secured. (author)

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    Gillen, K.T.; Malone, G.M.

    1997-07-01

    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

  2. A scoping evaluation of severe accidents at Surry and Grand Gulf Nuclear Power Plants resulting from earthquakes during shutdown conditions

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Davis, P.R.

    1991-01-01

    This report explores the likelihood of seismic-initiated core damage accidents during refueling shutdown conditions at two nuclear power plants, Surry Unit I and Grand Gulf Unit 1. The effort is scoping in character, and has been performed primarily to establish if a potential problem exists sufficient to justify a more rigorous and more quantitative evaluation. A summary is presented of the important conclusions that have been reached. The most important conclusion is that the core-damage frequencies for earthquake-initiated accidents during shutdown at both Surry Unit I and Grand Gulf Unit I are found to be low in absolute terms. The reasons for this are that in their ability to respond to earthquakes during shutdowns, the plants both have large seismic capacities, well above their design-basis levels; and also that both sites enjoy among the lowest seismic hazards of any LWR sites in the US

  3. Prevalence of oral health-related conditions that could trigger accidents for patients with moderate-to-severe dementia.

    Science.gov (United States)

    Kobayashi, Naoki; Soga, Yoshihiko; Maekawa, Kyoko; Kanda, Yuko; Kobayashi, Eiko; Inoue, Hisako; Kanao, Ayana; Himuro, Yumiko; Fujiwara, Yumi

    2017-03-01

    This study was performed to determine the prevalence of oral health conditions unnoticed by doctors and ward staff that may increase risk of incidents and/or accidents in hospitalised patients with moderate-severe dementia. Dementia patients may not recognise risks in the mouth, such as tooth mobility or ill-fitting dental prostheses and/or dentures. In addition to the risk of choking, injury by sharp edges of collapsed teeth or prosthodontics could pose risks. However, many previous publications were limited to case reports or series. Ninety-two consecutive hospitalised dementia patients (M: 52, F: 40, median age: 82.5 years, range: 62-99 years, from 2011 to 2014), referred for dentistry for dysphagia rehabilitation, were enrolled in this study. Participants referred for dental treatment with dental problems detected by ward staff were excluded. All participants had a Global Clinical Dementia Rating Score >2. Their dental records were evaluated retrospectively for issues that may cause incidents and/or accidents. Problems in the mouth, for example tooth stumps, dental caries, and ill-fitting dentures, were detected in 51.1% of participants (47/92). Furthermore, 23.9% (22/92) showed risk factors that could lead to incidents and/or accidents, for example falling out of teeth and/or prosthodontics or injury by sharp edges of teeth and/or prosthodontics. Hospitalised moderate-severe dementia patients had a high prevalence of oral health conditions unnoticed by doctors and ward staff that may increase risk of incidents and/or accidents. © 2016 John Wiley & Sons A/S and The Gerodontology Association. Published by John Wiley & Sons Ltd.

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

    International Nuclear Information System (INIS)

    Ma, C.W.; Sit, R.C.; Zavoshy, S.J.; Jardine, L.J.; Laub, T.W.

    1992-06-01

    This preliminary preclosure radiological safety analysis assesses the scenarios, probabilities, and potential radiological consequences associated with postulated accidents in the underground facility of the potential Yucca Mountain repository. The analysis follows a probabilistic-risk-assessment approach. Twenty-one event trees resulting in 129 accident scenarios are developed. Most of the scenarios have estimated annual probabilities ranging from 10 -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

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

    International Nuclear Information System (INIS)

    Heams, T.J.; Williams, D.A.; Johns, N.A.; Mason, A.; Bixler, N.E.; Grimley, A.J.; Wheatley, C.J.; Dickson, L.W.; Osborn-Lee, I.; Domagala, P.; Zawadzki, S.; Rest, J.; Alexander, C.A.; Lee, R.Y.

    1992-12-01

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

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

    International Nuclear Information System (INIS)

    Grégoire, A.-C.; Kalilainen, J.; Cousin, F.; Mutelle, H.; Cantrel, L.; Auvinen, A.; Haste, T.; Sobanska, S.

    2015-01-01

    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

  7. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

    Frid, W.

    1991-08-01

    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)

  8. How propeller suction is the dominant factor for ship accidents at shallow water conditions

    Science.gov (United States)

    Acar, Dursun; Alpar, Bedri; Ozeren, Sinan

    2017-04-01

    The laminar flow comes to the fore with the disappearance of the several other directions in the internal displacements in the water current. Due to the dominant speed direction during the straightforward motion of the ship, the underwater hull is associated with the continuous flow of laminar currents. The open marine environment acts as a compressible liquid medium because of the presence of many variables about water volume overflow boundaries where the ship is associated. Layers of water rising over the sea surface due to ship's body and the propeller's water push provides loss of liquid lifting force for the ship. These situations change the well-known sea-floor morphology and reliable depth limits, and lead to probable accidents. If the ship block coefficient for the front side is 0.7 or higher, the "squat" will be more on the bow, because the associated factor "displacement volume" causes to the low-pressure environment due to large and rapid turbulence. Thus, the bow sinks further, which faced with liquid's weaker lift force. The vessels Gerardus Mercator, Queen Elizabeth and Costa Concordia had accidents because of unified reasons of squat, fast water mass displacement by hull push and propeller suction interaction. In the case of water mass displacement from the bow side away, that accident occurred in 2005 by the vessel Gerardus Mercator with excessive longitudinal trim angularity in the shallow water. The vessel Costa Concordia (2012), voluminous water displaced from the rear left side was an important factor because of the sharp manoeuvre of that the captain made before the accident. Observations before the accident indicate that full-speed sharp turn provided listed position for the ship from left (port side) in the direction of travel before colliding and then strike a rock on the sloping side of the seabed. The reason why the ship drifted to the left depends mainly the water discharge occurred at the left side of the hull during left-hand rudder

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

    International Nuclear Information System (INIS)

    Koundy, V.; Dupas, J.; Bonneville, H.; Cormeau, I.

    2005-01-01

    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

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

    International Nuclear Information System (INIS)

    Nishiwaki, Y.; Kawai, H.; Shono, N.; Fujita, S.; Matsuoka, H.; Fujiwara, S.; Hosoda, T.

    2000-01-01

    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

  11. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    International Nuclear Information System (INIS)

    Gamble, Kyle Allan Lawrence; Hales, Jason Dean; Barani, Tommaso; Pizzocri, Davide; Pastore, Giovanni

    2016-01-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  12. Laser-heating and Radiance Spectrometry for the Study of Nuclear Materials in Conditions Simulating a Nuclear Power Plant Accident.

    Science.gov (United States)

    Manara, Dario; Soldi, Luca; Mastromarino, Sara; Boboridis, Kostantinos; Robba, Davide; Vlahovic, Luka; Konings, Rudy

    2017-12-14

    Major and severe accidents have occurred three times in nuclear power plants (NPPs), at Three Mile Island (USA, 1979), Chernobyl (former USSR, 1986) and Fukushima (Japan, 2011). Research on the causes, dynamics, and consequences of these mishaps has been performed in a few laboratories worldwide in the last three decades. Common goals of such research activities are: the prevention of these kinds of accidents, both in existing and potential new nuclear power plants; the minimization of their eventual consequences; and ultimately, a full understanding of the real risks connected with NPPs. At the European Commission Joint Research Centre's Institute for Transuranium Elements, a laser-heating and fast radiance spectro-pyrometry facility is used for the laboratory simulation, on a small scale, of NPP core meltdown, the most common type of severe accident (SA) that can occur in a nuclear reactor as a consequence of a failure of the cooling system. This simulation tool permits fast and effective high-temperature measurements on real nuclear materials, such as plutonium and minor actinide-containing fission fuel samples. In this respect, and in its capability to produce large amount of data concerning materials under extreme conditions, the current experimental approach is certainly unique. For current and future concepts of NPP, example results are presented on the melting behavior of some different types of nuclear fuels: uranium-plutonium oxides, carbides, and nitrides. Results on the high-temperature interaction of oxide fuels with containment materials are also briefly shown.

  13. Investigations on Health Conditions of Chernobyl Nuclear Power Plant Accident Recovery Workers from Latvia in Late Period after Disaster

    Directory of Open Access Journals (Sweden)

    Reste Jeļena

    2016-10-01

    Full Text Available The paper summarises the main findings on Chernobyl Nuclear Power Plant (CNPP accident recovery workers from Latvia and their health disturbances, which have been studied by the authors during the last two decades. Approximately 6000 persons from Latvia participated in CNPP clean-up works in 1986–1991. During their work period in Chernobyl they were exposed to external as well as to internal irradiation, but since their return to Latvia they were living in a relatively uncontaminated area. Regular careful medical examinations and clinical studies of CNPP clean-up workers have been conducted during the 25 years after disaster, gathering knowledge on radiation late effects. The aim of the present review is to summarise the most important information about Latvian CNPP clean-up worker health revealed by thorough follow-up and research conducted in the period of 25 years after the accident. This paper reviews data of the Latvian State Register of Persons Exposed to Radiation due to CNPP Accident and gives insight in main health effects found by the researchers from the Centre of Occupational and Radiological Medicine (Pauls Stradiņš Clinical University Hospital and Rīga Stradiņš University in a number of epidemiological, clinical, biochemical, immunological, and physiological studies. Latvian research data on health condition of CNPP clean-up workers in the late period after disaster indicate that ionising radiation might cause premature ageing and severe polymorbidity in humans.

  14. Predicting the impact of chronic health conditions on workplace productivity and accidents: results from two US Department of Energy national laboratories.

    Science.gov (United States)

    Frey, Jodi Jacobson; Osteen, Philip J; Berglund, Patricia A; Jinnett, Kimberly; Ko, Jungyai

    2015-04-01

    Examine associations of chronic health conditions on workplace productivity and accidents among US Department of Energy employees. The Health and Work Performance Questionnaire-Select was administered to a random sample of two Department of Energy national laboratory employees (46% response rate; N = 1854). The majority (87.4%) reported having one or more chronic health conditions, with 43.4% reporting four or more conditions. A population-attributable risk proportions analysis suggests improvements of 4.5% in absenteeism, 5.1% in presenteeism, 8.9% in productivity, and 77% of accidents by reducing the number of conditions by one level. Depression was the only health condition associated with all four outcomes. Results suggest that chronic conditions in this workforce are prevalent and costly. Efforts to prevent or reduce condition comorbidity among employees with multiple conditions can significantly reduce costs and workplace accident rates.

  15. Contributing factors in construction accidents.

    Science.gov (United States)

    Haslam, R A; Hide, S A; Gibb, A G F; Gyi, D E; Pavitt, T; Atkinson, S; Duff, A R

    2005-07-01

    This overview paper draws together findings from previous focus group research and studies of 100 individual construction accidents. Pursuing issues raised by the focus groups, the accident studies collected qualitative information on the circumstances of each incident and the causal influences involved. Site based data collection entailed interviews with accident-involved personnel and their supervisor or manager, inspection of the accident location, and review of appropriate documentation. Relevant issues from the site investigations were then followed up with off-site stakeholders, including designers, manufacturers and suppliers. Levels of involvement of key factors in the accidents were: problems arising from workers or the work team (70% of accidents), workplace issues (49%), shortcomings with equipment (including PPE) (56%), problems with suitability and condition of materials (27%), and deficiencies with risk management (84%). Employing an ergonomics systems approach, a model is proposed, indicating the manner in which originating managerial, design and cultural factors shape the circumstances found in the work place, giving rise to the acts and conditions which, in turn, lead to accidents. It is argued that attention to the originating influences will be necessary for sustained improvement in construction safety to be achieved.

  16. Coupled fuel performance and thermal-hydraulics simulation with BISON and RELAP-7 at accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Martineau, R.C., E-mail: Richard.Martineau@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' RELAP-7 is expected to be the next in the RELAP nuclear reactor safety/systems analysis application series developed at the Idaho National Laboratory (INL). The development of RELAP-7 began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of Department of Energy's (DOE) Light Water Reactor Sustainability (LWRS) Program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support modern nuclear power safety analysis. RELAP-7 is built using the INL's modern scientific software development framework, MOOSE (Multi-physics Object Oriented Simulation Environment). MOOSE provides improved implicit numerical schemes, including higher-order integration in both space and time, and yielding converged second-order accuracy for RELAP-7. The code structure is based on multiple physical component models such as pipes, junctions, pumps, etc. This component-based software architecture allows RELAP-7 to quickly adopt different physical models for different applications. One of the main advantages of building RELAP-7 on the MOOSE framework is that tight coupling with other MOOSE-based applications solving physics not present in RELAP-7 requires little to no additional lines of code. For example, the RELAP-7 core channel component is based upon a one-dimensional flow channel and a three-zone two-dimensional heat structure designed to represent fuel, gap, and cladding conjugate heat transfer with the coolant. However, the RELAP-7 application does not carry the fuels performance physics to analyze irradiated fuel, especially for accident scenarios. Here, we demonstrate the tightly coupled capability of the BISON nuclear fuels performance application with RELAP-7 for the station black out (SBO) accident scenario (Fukushima type event) and

  17. Dose assessment method for control room habitability in accident condition in nuclear power plants

    International Nuclear Information System (INIS)

    Yang Dong; Tang Shaohua; Wang Jianhua

    2012-01-01

    Based on the NRC. technical requirements on NPP control room habitability assessment, and considering the characteristics of the improved second generation NPPs in China, this paper developed a complete dose assessment model for control room habitability. Contrasting to the existing model in China, this model is applicable for DBA and sever accident, and the short term atmospheric diffusion factor can be calculated using the combined wake mode. By considering the zoning of habitable area and the design characteristics of the ventilation system, the effects of un-filtrated air leakage from the building and the ventilation system on the assessment calculation can be considered. (authors)

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

    International Nuclear Information System (INIS)

    Ohnishi, Nobuaki; Ishijima, Kiyomi; Ochiai, Masaaki; Tanzawa, Sadamitsu; Uemura, Mutsumi

    1981-05-01

    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)

  19. PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Weber, P.; Umminger, K.J.; Schoen, B. [Siemens AG Power Generation Group (KWU), Erlangen (France)

    1995-09-01

    The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

  20. Accident scenario diagnostics with neural networks

    International Nuclear Information System (INIS)

    Guo, Z.

    1992-01-01

    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

  1. Development of high-performance monitoring system under severe accident condition

    International Nuclear Information System (INIS)

    Takeuchi, Tomoaki; Tsuchiya, Kunihiro; Ishihara, Masahiro; Komanome, H.; Miura, K.

    2017-01-01

    A research and development of a monitoring system for NPPs situations even during severe accidents have been performed. The R and D consists of the three objectives. The major findings are briefly summarized in the followings: 1) Radiation-resistant monitoring camera. The image sensor with the photogate and three transistors was found to be advantageous in terms of dark current and sensitivity. In addition, radiation-resistant optical parts and signal circuits were successfully fabricated. The results suggested that the monitoring camera system with 10 6 Gy in radiation resistance was possible. 2) Radiation-resistant in-water wireless transmission system. A two-dimensional LED matrix with 10 6 Gy in radiation resistance and a camera were used as the transmission devices. The results of the in-water transmission tests suggested that stable wireless transmission between 5 m distance was possible even with bubble, turbidity, or obstacles. 3) Heat-resistant signal cable. In order to develop a cable that can transmit the data inside reactor pressure vessels, heat-proof tests were performed for candidate metallic sheath materials of mineral insulation (MI) cables. The results indicated MI cables which can be used at 1000degC in air were possible. These results indicate the feasibility of the monitoring system even during severe accidents. (author)

  2. The FIRAC code - its applicability and boundary conditions for fire accident analysis in a reprocessing plant

    International Nuclear Information System (INIS)

    Roewekamp, M.

    1991-01-01

    After a short description of the modelling capabilities and the implementation of the computer code the possible applications of FIRAC are demonstrated by means of two test-examples. The so gained experiences with respect to the variation of different parameters, convergency criteria, etc. can be used for the simulation of a fire accident in the storage area for unconditioned combustible low active waste (LAW) of the planned reprocessing plant at Wackersdorf. The code is prepared for calculating direct effects (of the fire) in the fire room as well as particularly effects on adjacent rooms and ventilation systems. Source terms for the release of radioactive particles outside a building can also be investigated. The temperature and pressure curves for the fire room as well as for other areas in the facility show that no damages caused by temperature effects are expected for the considered fire of low active waste. As a result of the calculated mass and volumetric flows radioactive aerosole particles could be transported into normally non-active areas. The FIRAC code renders the possibility of a more detailed analysis of those parameters relevant for fire accidents and by this means completes the so far phenomenological procedure of the fire hazard analysis in nuclear facilities. (orig.) [de

  3. Establishing key components of yoga interventions for musculoskeletal conditions: a Delphi survey

    Science.gov (United States)

    2014-01-01

    Background Evidence suggests yoga is a safe and effective intervention for the management of physical and psychosocial symptoms associated with musculoskeletal conditions. However, heterogeneity in the components and reporting of clinical yoga trials impedes both the generalization of study results and the replication of study protocols. The aim of this Delphi survey was to address these issues of heterogeneity, by developing a list of recommendations of key components for the design and reporting of yoga interventions for musculoskeletal conditions. Methods Recognised experts involved in the design, conduct, and teaching of yoga for musculoskeletal conditions were identified from a systematic review, and invited to contribute to the Delphi survey. Forty-one of the 58 experts contacted, representing six countries, agreed to participate. A three-round Delphi was conducted via electronic surveys. Round 1 presented an open-ended question, allowing panellists to individually identify components they considered key to the design and reporting of yoga interventions for musculoskeletal conditions. Thematic analysis of Round 1 identified items for quantitative rating in Round 2; items not reaching consensus were forwarded to Round 3 for re-rating. Results Thirty-six panellists (36/41; 88%) completed the three rounds of the Delphi survey. Panellists provided 348 comments to the Round 1 question. These comments were reduced to 49 items, grouped under five themes, for rating in subsequent rounds. A priori group consensus of ≥80% was reached on 28 items related to five themes concerning defining the yoga intervention, types of yoga practices to include in an intervention, delivery of the yoga protocol, domains of outcome measures, and reporting of yoga interventions for musculoskeletal conditions. Additionally, a priori consensus of ≥50% was reached on five items relating to minimum values for intervention parameters. Conclusions Expert consensus has provided a non

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

    Directory of Open Access Journals (Sweden)

    G. L. Zakharchenko

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  6. Influence of postharvest processing and storage conditions on key antioxidants in pūhā (Sonchus oleraceus L.)

    DEFF Research Database (Denmark)

    Ou, Zong-Quan; Schmierer, David M; Strachan, Clare J

    2014-01-01

    To investigate effects of different postharvest drying processes and storage conditions on key antioxidants in Sonchus oleraceus L. leaves.......To investigate effects of different postharvest drying processes and storage conditions on key antioxidants in Sonchus oleraceus L. leaves....

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

    International Nuclear Information System (INIS)

    Tang, I.N.; Munkelwitz, H.R.

    1982-07-01

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

  8. Tritium loading in ITER plasma-facing surfaces and its release under accident conditions

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Pawelko, R.J.

    1996-01-01

    Plasma-facing surfaces of the International Thermonuclear Experimental Reactor (ITER) will take up tritium from the plasma. These surfaces will probably consist of matures of Be, C, and possibly W together with other impurities. Recent experimental results have suggested mechanisms, not previously considered in analyses, by which tritium and other hydrogen isotopes are retained in Be. This warrants revised modeling and estimation of the amount of tritium that will be deposited in ITER beryllium plasma-facing surfaces and the rates at which it can be released under postulated accident scenarios. In this paper we describe improvements in modeling and experiments planned at the Idaho National Engineering Laboratory (INEL) to investigate the tritium uptake and thermal release behavior for mixed plasma- facing materials. TMAP4 calculations were made using recent data to estimate first-wall tritium inventories in ITER. 16 refs., 1 fig

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

    International Nuclear Information System (INIS)

    Taig, A.R.

    1981-01-01

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

  10. Thermal and hydraulic behaviour of CANDU cores under severe accident conditions - final report

    International Nuclear Information System (INIS)

    Rogers, J.T.

    1984-06-01

    This volume of appendices presents listings and sample runs of the computer codes used in the study of the thermalhydraulic behaviour of CANDU reactor cores during severe loss of coolant accidents. The codes, written in standard FORTRAN, are MODBOIL, to calculate moderator temperatures, pressures and water levels; DEBRIS, to calculate the transient temperature distribution in the debris of calandria and pressure tubes and fuel pellets; MOLTENPOOL, to calculate the temperature history in a pool of molten debris; CONFILM, to calculate the behaviour of a condensing film of vaporized core debris on the calandria wall, and BLDG, to calculate the pressurization of the containment during the expulsion of moderator through pressure relief ducts. In addition there are discussions of the average condensation heat transfer coefficient for vaporized core material on the calandria wall, and of vapor explosions

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

    International Nuclear Information System (INIS)

    Kamarli, Z.; Abdulina, A.

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

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

    International Nuclear Information System (INIS)

    Papin, Joelle; Lemoine, Francette; Sato, Ikken; Struwe, Dankward; Pfrang, Werner

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Lucas, M.

    1988-01-01

    Measurements were made of I/sub 2/ formed when aqueous cesium iodide (CsI) solutions were exposed to two temperatures, 43 and 95 0 C, with irradiation. Iodine partition coefficients were obtained from the experiments. The parameters varied were dose, CsI concentration, and Cs/sub 2/CO/sub 3/ concentration, in the presence of air-carbon dioxide and air-carbon dioxide-hydrogen mixtures, to provide information to calculate the form in which iodine released from fuel as CsI in a reactor accident might reach the environment. In a series of experiments, a two-compartment cell was used to trap the gaseous iodine produced. In this case, it was found that the quantity of gaseous iodine produced increased approximately linearly with the dose (at the dose rate used)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

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

  16. Application of the SCANAIR code for VVER RIA conditions - Boron dilution accident

    International Nuclear Information System (INIS)

    Arffman, A.; Cazalis, B.

    2010-01-01

    This paper consists of two parts. In part A, RIA pulse tests conducted at the Russian BIGR reactor are being analysed at IRSN with SCANAIR V6 fuel performance code as a part of the code validation for VVER fuel. Recently a new version of the SCANAIR code was made available to VTT Technical Research Centre of Finland, and part B of the paper covers the introduction of the code version at VTT by a calculation of a hypothetical boron dilution accident in a VVER-440 power reactor. Concerning part A, it appears that the SCANAIR V6 version, including a BIGR/NSRR heat transfer model, validated by Japanese NSRR experiments, and a Norton viscoplastic clad mechanical behaviour, is able to simulate the rod thermal behaviour in BIGR tests. Concerning the clad mechanics, it has been seen that a pellet swelling model is able to simulate the average rod deformation. Nonetheless, the current clad creep model associated with the free volume equilibrium assumption is not suited to predict the maximum clad deformation and the possible post DNB rod failure because they do not simulate local balloons. Furthermore, it has been shown that the clad deformation is strongly dependent on transient gas transfer. Concerning part B, a boron dilution accident previously calculated with SCANAIR V2 was recalculated with SCANAIR V6. A limited amount of result parameters were compared with the results of VTT's neutronics code TRAB. Divergence problems encountered previously when reaching the DNB limit were not present anymore. Fuel and cladding temperatures produced by SCANAIR were in good agreement with those calculated with TRAB

  17. The influence of the crust layer on RPV structural failure under severe accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Mao, Jianfeng, E-mail: jianfeng-mao@163.com [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Li, Xiangqing [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Bao, Shiyi [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Luo, Lijia [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Gao, Zengliang [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China)

    2017-05-15

    Highlights: • The crust layer greatly affects the RPV structural behavior. • The RPV failure is investigated in depth under severe accident. • The creep and plastic damage mainly contribute to RPV failure. • An elastic core in RPV wall is essential for ensuring RPV integrity. • The multiaxial state of stress accelerates the total damage evolution. - Abstract: The so called ‘in-vessel retention (IVR)’ is regarded as a severe accident (SA) mitigation strategy, which is widely used in most of advanced nuclear power plants. The effectiveness of IVR strategy is to employ the external water flooding to cool the reactor pressure vessel (RPV). The RPV integrity has to be maintained within a required period during the IVR period. The degraded melting core is assumed to be arrested in the lower head (LH) to form the melting pool that is bounded by upper, side and lower crusts. Consequently, the existence of the crust layer greatly affects the RPV structural behavior as well as failure process. In order to disclose this influence caused by the crust layer, a detailed investigation is conducted by using numerical simulation on the two RPVs with and without crust layer respectively. Taking the RPV without crust layer as a basis for the comparison, the present study assesses the likelihood and potential failure location, time and mode of the LH under the loadings of the critical heat flux (CHF) and slight internal pressure. Due to the high temperature melt on the inside and nucleate boiling on the outside, the RPV integrity is found to be compromised by melt-through, creep, elasticity, plasticity as well as thermal expansion. Through in-depth investigation, it is found that the creep and plasticity are of vital importance to the final structural failure, and the introduction of crust layer results in a significant change on field parameters in terms of temperature, deformation, stress(strain), triaxiality factor and total damage.

  18. The influence of the crust layer on RPV structural failure under severe accident condition

    International Nuclear Information System (INIS)

    Mao, Jianfeng; Li, Xiangqing; Bao, Shiyi; Luo, Lijia; Gao, Zengliang

    2017-01-01

    Highlights: • The crust layer greatly affects the RPV structural behavior. • The RPV failure is investigated in depth under severe accident. • The creep and plastic damage mainly contribute to RPV failure. • An elastic core in RPV wall is essential for ensuring RPV integrity. • The multiaxial state of stress accelerates the total damage evolution. - Abstract: The so called ‘in-vessel retention (IVR)’ is regarded as a severe accident (SA) mitigation strategy, which is widely used in most of advanced nuclear power plants. The effectiveness of IVR strategy is to employ the external water flooding to cool the reactor pressure vessel (RPV). The RPV integrity has to be maintained within a required period during the IVR period. The degraded melting core is assumed to be arrested in the lower head (LH) to form the melting pool that is bounded by upper, side and lower crusts. Consequently, the existence of the crust layer greatly affects the RPV structural behavior as well as failure process. In order to disclose this influence caused by the crust layer, a detailed investigation is conducted by using numerical simulation on the two RPVs with and without crust layer respectively. Taking the RPV without crust layer as a basis for the comparison, the present study assesses the likelihood and potential failure location, time and mode of the LH under the loadings of the critical heat flux (CHF) and slight internal pressure. Due to the high temperature melt on the inside and nucleate boiling on the outside, the RPV integrity is found to be compromised by melt-through, creep, elasticity, plasticity as well as thermal expansion. Through in-depth investigation, it is found that the creep and plasticity are of vital importance to the final structural failure, and the introduction of crust layer results in a significant change on field parameters in terms of temperature, deformation, stress(strain), triaxiality factor and total damage.

  19. Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Park, Dong Jun, E-mail: pdj@kaeri.re.kr; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

    2016-12-15

    This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility. - Highlights: • Cr and FeCrAl were coated onto Zr fuel cladding for light water nuclear reactors. • Mo layer between FeCrAl and Zr prevented inter-diffusion at high temperatures. • Coated claddings were tested under loss-of-cooling accident conditions. • Coating improved high-temperature oxidation resistance and mechanical properties.

  20. Performance behavior of the passive containment cooling system of a natural circulation BWR during postulated accident condition

    International Nuclear Information System (INIS)

    Kumar, Mukesh; Nayak, A.K.; Jain, Vikas; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    Passive systems are playing prominent role in the development of innovative nuclear reactor systems due to their simplicity, enhanced safety, reliability and economy. These systems are being considered for normal operation as well as accidental conditions of reactor following a postulated accident scenario to preclude the scenarios arising out of failure of active systems as well as to minimize the operator intervention. Indian innovative reactor AHWR being designed for thorium utilization employs various passive safety concepts. As containment is the ultimate barrier to the release of radioactivity, passive concepts are being employed in BWRs for minimize peak containment pressure in the containment during a postulated accident condition like LOCA. The concept of passive containment cooling system (PCCS) in the AHWR comprises of inclined tube heat exchangers located underneath an elevated pool that removes the heat from the steam-air atmosphere of containment following a LOCA by natural circulation of water inside the tubes. The steam condenses on the external surface of tubes of PCCS in addition to the wall of the containment which in turn depressurizes the containment. This paper deals with the performance assessment of PCCS of AHWR during a postulated design basis LOCA by using the best estimate code RELAP5/Mod3.2. (author)

  1. Determination of gamma-ray exposure rate from short-lived fission products under criticality accident conditions

    International Nuclear Information System (INIS)

    Yanagisawa, Hiroshi; Ohno, Akio; Aizawa, Eijyu

    2002-01-01

    For the assessment of γ-ray doses from short-lived fission products (FPs) under criticality accident conditions, γ-ray exposure rates varying with time were experimentally determined in the Transient Experiment Critical Facility (TRACY). The data were obtained by reactivity insertion in the range of 1.50 to 2.93$. It was clarified from the experiments that the contribution of γ-ray from short-lived FPs to total exposure during the experiments was evaluated to be 15 to 17%. Hence, the contribution cannot be neglected for the assessment of γ-ray doses under criticality accident conditions. Computational analyses also indicated that γ-ray exposure rates from short-lived FPs calculated with the Monte Carlo code, MCNP4B, and photon sources based on the latest FP decay data, the JENDL FP Decay Data File 2000, well agreed with the experimental results. The exposure rates were, however, extremely underestimated when the photon sources were obtained by the ORIGEN2 code. The underestimation is due to lack of energy-dependent photon emission data for major short-lived FP nuclides in the photon database attached to the ORIGEN2 code. It was also confirmed that the underestimation arose in 1,000 or less of time lapse after an initial power burst. (author)

  2. Thermal hydraulic behavior of a PWR under beyond-design-basis accident conditions: Conclusions from an experimental program in a 4-loop test facility (PKL)

    International Nuclear Information System (INIS)

    Umminger, K.J.; Kastner, W.; Mandl, R.M.; Weber, P.

    1993-01-01

    Within the scope of German reactor safety research, extensive experiments covering the behavior of nuclear power plants under accident conditions have been carried out in the PKL test facility which simulates a 4-loop, 1,300 MWe KWU-designed PWR. While the investigations dealing with design-basis accidents and with the efficiency of the emergency core cooling systems have been largely completed, the main interest nowadays concentrates on the investigation of beyond-design-basis accidents to demonstrate the safety margins of nuclear power plants and to investigate the contribution of the built-in safety features for a further reduction of the residual risk. The thermal hydraulic behavior of a PWR under these extreme accident conditions was experimentally investigated within the PKL III B test program. This paper presents the fundamental findings with some of the most important results being discussed in detail. Future plans are also outlined

  3. Development of Assessment Methodology of Chemical Behavior of Volatile Iodide under Severe Accident Conditions Using EPICUR Experiments

    International Nuclear Information System (INIS)

    Oh, Jae Yong; Yun, Jong Il; Kim, Do Sam; Han Chul

    2011-01-01

    Iodine is one of the most important fission products produced in nuclear power plants. Under severe accident condition, iodine exists as a variety of species in the containment such as aqueous iodide, gaseous iodide, iodide aerosol, etc. Following release of iodine from the reactor, mostly in the form of CsI aerosol, volatile iodine can be generated from the containment sump and release to the environment. Especially, volatile organic iodide can be produced from interaction between nonvolatile iodine and organic substances present in the containment. Volatile iodide could significantly influence the alienated residents surrounding the nuclear power plant. In particular, thyroid is vulnerable to radioiodine due to its high accumulation. Therefore, it is necessary for the Korea Institute of Nuclear Safety (KINS) to develop an evaluation model which can simulate iodine behavior in the containment following a severe accident. KINS also needs to make up its methodology for radiological consequence analysis, based on MELCOR-MACCS2 calculation, by coupling a simple iodine model which can conveniently deal with organic iodides. In the long term, such a model can contribute to develop an accident source term, which is one of urgent domestic needs. Our strategy for developing the model is as follows: 1. Review the existing methodologies, 2. Develop a simple stand-alone model, 3. Validate the model using ISTP-EPICUR (Experimental Program on Iodine Chemistry under Radiation) and OECD-BIP (Behavior of Iodine Project) experimental data. In this paper we present the context of development and validation of our model named RAIM (Radio-active iodine chemistry model)

  4. Development of Assessment Methodology of Chemical Behavior of Volatile Iodide under Severe Accident Conditions Using EPICUR Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jae Yong; Yun, Jong Il [KAIST, Daejeon (Korea, Republic of); Kim, Do Sam; Han Chul [Korea Institue of Nuclear Safety, Daejeon (Korea, Republic of)

    2011-05-15

    Iodine is one of the most important fission products produced in nuclear power plants. Under severe accident condition, iodine exists as a variety of species in the containment such as aqueous iodide, gaseous iodide, iodide aerosol, etc. Following release of iodine from the reactor, mostly in the form of CsI aerosol, volatile iodine can be generated from the containment sump and release to the environment. Especially, volatile organic iodide can be produced from interaction between nonvolatile iodine and organic substances present in the containment. Volatile iodide could significantly influence the alienated residents surrounding the nuclear power plant. In particular, thyroid is vulnerable to radioiodine due to its high accumulation. Therefore, it is necessary for the Korea Institute of Nuclear Safety (KINS) to develop an evaluation model which can simulate iodine behavior in the containment following a severe accident. KINS also needs to make up its methodology for radiological consequence analysis, based on MELCOR-MACCS2 calculation, by coupling a simple iodine model which can conveniently deal with organic iodides. In the long term, such a model can contribute to develop an accident source term, which is one of urgent domestic needs. Our strategy for developing the model is as follows: 1. Review the existing methodologies, 2. Develop a simple stand-alone model, 3. Validate the model using ISTP-EPICUR (Experimental Program on Iodine Chemistry under Radiation) and OECD-BIP (Behavior of Iodine Project) experimental data. In this paper we present the context of development and validation of our model named RAIM (Radio-active iodine chemistry model)

  5. Some Examples of the Relationship Between Containment and Other Engineered Safeguard Requirements, Accident Analyses and Site Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Vinck, W. F.; Maurer, H. [EURATOM, Brussels (Belgium)

    1967-09-15

    The paper refers primarily to nuclear power reactors for which EURATOM has performed safety reviews in co-operation with national technical advisory organizations concerned in the licensing procedures. Comparative data are tabulated on a number of containment concepts and other engineered safeguards to protect against or to limit the consequences of major hypothetical accidents for a number of power reactors. Main environmental data, such as magnitudes of exclusion areas and population densities, are also presented. A number of topics of particular interest which were encountered during the safety analysis and which find widespread application in the assessment of the siting conditions and emergency planning are discussed. In this discussion, emphasis is placed on containment, engineered safeguards and emergency equipment. These items are considered in general with emphasis on: (a ) the importance of meeting operability and high efficiency requirements (reliability ) when needed through design and layout; (b) periodic testing and/or inspection possibilities and requirements in order to maintain high availability standards. Attention is drawn to some difficulties which have arisen in connection with design, material choice and construction of steel containment structures and which in the interests of safety as well as of economic optimization justify in the future more care in the use of possibly uniform or single-code requirements. Examples of uncertainties encountered in some accident analyses and their influence on siting considerations are discussed, with emphasis on safeguards intended to retain radioactive material in the plant, such as extent of core damage, iodine plate-out, filtering efficiency, wash-out effects. The means by which some of the main uncertainties have been or can in the future be eliminated through appropriate experimental programmes performed throughout the world are discussed. Some examples are also given of the influence of factors which

  6. Analysis of ex-vessel melt jet breakup and coolability. Part 1: Sensitivity on model parameters and accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kiyofumi; Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr; Hwang, Byoungcheol; Jung, Woo Hyun

    2016-06-15

    Highlights: • Application of JASMINE code to melt jet breakup and coolability in APR1400 condition. • Coolability indexes for quasi steady state breakup and cooling process. • Typical case in complete breakup/solidification, film boiling quench not reached. • Significant impact of water depth and melt jet size; weak impact of model parameters. - Abstract: The breakup of a melt jet falling in a water pool and the coolability of the melt particles produced by such jet breakup are important phenomena in terms of the mitigation of severe accident consequences in light water reactors, because the molten and relocated core material is the primary heat source that governs the accident progression. We applied a modified version of the fuel–coolant interaction simulation code, JASMINE, developed at Japan Atomic Energy Agency (JAEA) to a plant scale simulation of melt jet breakup and cooling assuming an ex-vessel condition in the APR1400, a Korean advanced pressurized water reactor. Also, we examined the sensitivity on seven model parameters and five initial/boundary condition variables. The results showed that the melt cooling performance of a 6 m deep water pool in the reactor cavity is enough for removing the initial melt enthalpy for solidification, for a melt jet of 0.2 m initial diameter. The impacts of the model parameters were relatively weak and that of some of the initial/boundary condition variables, namely the water depth and melt jet diameter, were very strong. The present model indicated that a significant fraction of the melt jet is not broken up and forms a continuous melt pool on the containment floor in cases with a large melt jet diameter, 0.5 m, or a shallow water pool depth, ≤3 m.

  7. Raim – A model for iodine behavior in containment under severe accident condition

    Directory of Open Access Journals (Sweden)

    Han-Chul Kim

    2015-12-01

    Full Text Available Following a severe accident in a nuclear power plant, iodine is a major contributor to the potential health risks for the public. Because the amount of iodine released largely depends on its volatility, iodine's behavior in containment has been extensively studied in international programs such as International Source Term Programme-Experimental Program on Iodine Chemistry under Radiation (EPICUR, Organization for Economic Co-operation and Development (OECD-Behaviour of Iodine Project, and OECD-Source Term Evaluation and Mitigation. Korea Institute of Nuclear Safety (KINS has joined these programs and is developing a simplified, stand-alone iodine chemistry model, RAIM (Radio-Active Iodine chemistry Model, based on the IMOD methodology and other previous studies. This model deals with chemical reactions associated with the formation and destruction of iodine species and surface reactions in the containment atmosphere and the sump in a simple manner. RAIM was applied to a simulation of four EPICUR tests and one Radioiodine Test Facility test, which were carried out in aqueous or gaseous phases. After analysis, the results show a trend of underestimation of organic and molecular iodine for the gas-phase experiments, the opposite of that for the aqueous-phase ones, whereas the total amount of volatile iodine species agrees well between the experiment and the analysis result.

  8. RAIM-A model for iodine behavior in containment under severe accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han Chul; Cho, Yeong Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-12-15

    Following a severe accident in a nuclear power plant, iodine is a major contributor to the potential health risks for the public. Because the amount of iodine released largely depends on its volatility, iodine's behavior in containment has been extensively studied in international programs such as International Source Term Programme-Experimental Program on Iodine Chemistry under Radiation (EPICUR), Organization for Economic Co-operation and Development (OECD)-Behaviour of Iodine Project, and OECD-Source Term Evaluation and Mitigation. Korea Institute of Nuclear Safety (KINS) has joined these programs and is developing a simplified, stand-alone iodine chemistry model, RAIM (Radio-Active Iodine chemistry Model), based on the IMOD methodology and other previous studies. This model deals with chemical reactions associated with the formation and destruction of iodine species and surface reactions in the containment atmosphere and the sump in a simple manner. RAIM was applied to a simulation of four EPICUR tests and one Radioiodine Test Facility test, which were carried out in aqueous or gaseous phases. After analysis, the results show a trend of underestimation of organic and molecular iodine for the gas-phase experiments, the opposite of that for the aqueous-phase ones, whereas the total amount of volatile iodine species agrees well between the experiment and the analysis result.

  9. Environmental gamma radiation measurements in Finland and the influence of the meteorological conditions after the Chernobyl accident in 1986

    International Nuclear Information System (INIS)

    Arvela, H.; Blomqvist, L.; Lemmelae, H.; Savolainen, A.L.; Sarkkula, S.

    1987-06-01

    Results from a survey of environmental gamma radiation levels in Finland after the Chernobyl accident 1986 were presented. The measurements were made by means of sensitive Geiger-counters and a gamma-spectrometer placed in cars. The results presented the level of external radiation caused by the cesium fallout on the first of October 1986. In the center of Southern Finland there are wide areas with exposure levels exceeding 0.04 μSv h -1 , areas exceeding 0.2 μSv h -1 being very rare. The surface area weighted mean dose rate for the 461 municipalities in Finland was 0.037 $mu$Sv h -1 (range 0-0.23 μSv h -1 ). The corresponding estimated surface activity of 137 Cs was 10.7 kBq m -2 . The population weighted mean dose rate was 0.051 μSv h -1 . Results from measurements at eight dose rate monitoring stations were presented as daily dose rate recordings in 1985-1986, the rate of decrease of the excess dose rate demonstrating quite large variations in the period from May to August. This indicated that the composition of the short-lived nuclides in the fallout varied from place to place. The influence of the meteorological conditions were reported with precipitation data from six days after the accident. There was a clear correlation between the results from precipitation and radiation measurements in different parts of Finland

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

    International Nuclear Information System (INIS)

    Sosnovsky, S.; Danilicnev, V.; Nikiforov, A.; Zybina, N.; Nesteruk, L.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sosnovsky, S; Danilicnev, V [Military Medical Academy, Saint-Petersburg (Russian Federation); Nikiforov, A; Zybina, N [All-Russian Center of Emergency and Radiation Medicine, Saint-Petersburg (Russian Federation); Nesteruk, L [Helmholtz Inst. of Eyes Diseases, Moscow (Russian Federation)

    1997-11-01

    84 liquidators of consequences of Chernobyl APS accident from the age of 28 to 58 were examined. The control group was made with 22 men from the age of 28 to 52. A certain increase of infringement of a transparency of lens without typical attributes of radiating cataract is revealed in the experimental group. Electrophysiological investigation (EPI) shows a certain reduction of amplitude of a wave ``a`` of macular electroretinogram (ERG) on green stimulus, amplitude of a main component and lengthening of an interpeak time interval of flicker ERG 10 Hz is revealed. These changes indicate the tendency to reduction of functional activity of a retina (first of all at a level of photoreceptors) in paramacular and in a smaller degree in peripheral zones among liquidators. The parameters of contrast sensitivity are definitely reduced in the experimental group for all stimuli on all spatial frequencies. Luminous and colour sensitivity to stimuli of different colour in the experimental group is definitely reduced in all central field of sight, but in paracentral zone the degree of reduction is higher. We investigated the parameters of oxidative stress in both groups. Definite increase of production of the reactive oxygen species and disbalance of a glutathione link of antioxidant protection are revealed. Authentic correlation dependences are revealed: moderate direct correlation - between a level of glutathione reductase and amplitude of a main component of flicker ERG 10 Hz, between a level of oxidized glutathione and interpeak time interval of flicker ERG 10 Hz, inverse correlation - between the level of oxidized glutathione and amplitude of a main component of flicker ERG 10 Hz. In view of large spontaneous activity of free radical processes in a retina in norm the received results can explain revealed changes of an organ of vision. (author). 5 refs.

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

    International Nuclear Information System (INIS)

    Almerico, F.; Machiels, A.J.; Ornberg, S.C.; Lahti, G.P.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Hassan, Yassin

    2013-10-01

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

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

    International Nuclear Information System (INIS)

    Glumac, B.; Ravnik, M.; Logar, M.

    1997-01-01

    Additional criticality safety analysis of a pool-type storage for TRIGA spent fuel at the Jozef Stefan Institute in Ljubljana, Slovenia, is presented. Previous results have shown that subcriticality is not guaranteed for some postulated accidents (earthquake with subsequent fuel rack disintegration resulting in contact fuel pitch) under the assumption that the fuel rack is loaded with fresh 12 wt% standard fuel. To mitigate this deficiency, a study was done on replacing a certain number of fuel elements in the rack with cadmium-loaded absorber rods. The Monte Carlo computer code MCNP4A with an ENDF/B-V library and detailed three-dimensional geometrical model of the spent-fuel rack was used for this purpose. First, a minimum critical number of fuel elements was determined for contact pitch, and two possible geometries of rack disintegration were considered. Next, it was shown that subcriticality can be ensured when pitch is decreased from a rack design pitch of 8 cm to contact, if a certain number of fuel elements (8 to 20 out of 70) are replaced by absorber rods, which are uniformly mixed into the lattice. To account for the possibility that random mixing of fuel elements and absorber rods can occur during rack disintegration and result in a supercritical configuration, a probabilistic study was made to sample the probability density functions for random absorber rod lattice loadings. Results of the calculations show that reasonably low probabilities for supercriticality can be achieved (down to 10 -6 per severe earthquake, which would result in rack disintegration and subsequent maximum possible pitch decrease) even in the case where fresh 12 wt% standard TRIGA fuel would be stored in the spent-fuel pool

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-06

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

  16. Natural circulation cooling in a PWR geometry under accident-induced conditions

    International Nuclear Information System (INIS)

    Shimeck, D.J.; Johnsen, G.W.

    1983-01-01

    The characteristics and limits of natural circulation heat rejection over a wide range of conditions were experimentally investigated in a small-scale model of a pressurized water reactor. Conditions that were varied included primary and secondary coolant inventory, decay heat power, and primary noncondensable gas content. The results have defined three distinct modes of natural circulation, their limits and transition points, and the characteristic signatures accompanying natural circulation behavior. Particular emphasis is focused on the limits of natural circulation under severely degraded primary and secondary conditions

  17. The role of seasonal, climatic and meteorological conditions in modifying nuclear accident consequences

    International Nuclear Information System (INIS)

    Mueller, H.; Proehl, G.

    1989-01-01

    One of the most important factors which influence the ingestion doses after an accidental release of radionuclides is the season of the year at which the release occurs. This is demonstrated with some examples for German conditions. This seasonal effect depends strongly on the growing periods of the different plants. Therefore it is influenced by the climatic conditions which vary to a large degree in the different countries causing very different growing periods. The influence of the meteorological conditions during and after the passing of a radioactive cloud on the initial contamination of the plants is discussed

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

    International Nuclear Information System (INIS)

    Haste, T.J.

    1982-07-01

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

  19. Sizing of type B package tie-downs on the basis of criteria related to hypothetical road transport accident conditions

    International Nuclear Information System (INIS)

    Phalippou, C.

    1986-01-01

    The aim is to guarantee intactness of the type B package containment system under hypothetical road accident conditions. Some experiments performed in France have led to analytical studies taking into account: a) the head-on collision, which is modelised by a uniform deceleration of 35 g, b) the side-on collision, which is modelised by a colliding object 3 times heavier than the package and an impact at 31.9 km/h. In the first case, the adopted criterion is the holding of the package on the vehicle by the strenght of the stowing members (tie-downs and chocks). In the second case, the adopted criterion is the desired breaking of the tie-downs in order to undamage package containment system; therefore it is assumed that no chock is acting against lateral impacts. Analytical and abacus methods have been developed for sizing the strenght of the stowing members in respect with the two above criteria [fr

  20. The effect of self-leveling on debris bed coolability under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Basso, S.; Konovalenko, A. [Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, D5, Stockholm 106 91 (Sweden); Yakush, S.E. [Institute for Problems in Mechanics of the Russian Academy of Sciences, Ave. Vernadskogo 101 Bldg 1, Moscow 119526 (Russian Federation); Kudinov, P. [Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, D5, Stockholm 106 91 (Sweden)

    2016-08-15

    Highlights: • A model for coolability of a self-leveling, variable-shape debris bed is proposed. • Sensitivity analysis is performed to screen out the less influential input parameters. • A small fraction of scenarios has initially a non-coolable debris bed configuration. • The fraction of non-coolable scenarios decreases with time due to self-leveling. - Abstract: Nordic-type boiling water reactors employ melt fragmentation, quenching, and long term cooling of the debris bed in a deep pool of water under the reactor vessel as a severe accident (SA) mitigation strategy. The height and shape of the bed are among the most important factors that determine if decay heat can be removed from the porous debris bed by natural circulation of water. The debris bed geometry depends on its formation process (melt release, fragmentation, sedimentation and settlement on the containment basemat), but it also changes with time afterwards, due to particle redistribution promoted by coolant flow (self-leveling). The ultimate goal of this work is to develop an approach to the assessment of the probability that debris in such a variable-shape bed can reach re-melting (which means failure of SA mitigation strategy), i.e. the time necessary for the slumping debris bed to reach a coolable configuration is larger than the time necessary for the debris to reach the re-melting temperature. For this purpose, previously developed models for particulate debris spreading by self-leveling and debris bed dryout are combined to assess the time necessary to reach a coolable state and evaluate its uncertainty. Sensitivity analysis was performed to screen out less important input parameters, after which Monte Carlo simulation was carried out in order to collect statistical characteristics of the coolability time. The obtained results suggest that, given the parameters ranges typical of Nordic BWRs, only a small fraction of debris beds configurations exhibits the occurrence of dryout. Of the

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

    Science.gov (United States)

    Ohnishi, Takeo

    2012-07-01

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

  2. Modelling management process of key drivers for economic sustainability in the modern conditions of economic development

    Directory of Open Access Journals (Sweden)

    Pishchulina E.S.

    2017-01-01

    Full Text Available The text is about issues concerning the management of driver for manufacturing enterprise economic sustainability and manufacturing enterprise sustainability assessment as the key aspect of the management of enterprise economic sustainability. The given issues become topical as new requirements for the methods of manufacturing enterprise management in the modern conditions of market economy occur. An economic sustainability model that is considered in the article is an integration of enterprise economic growth, economic balance of external and internal environment and economic sustainability. The method of assessment of economic sustainability of a manufacturing enterprise proposed in the study allows to reveal some weaknesses in the enterprise performance, and untapped reserves, which can be further used to improve the economic sustainability and efficiency of the enterprise. The management of manufacturing enterprise economic sustainability is one of the most important factors of business functioning and development in modern market economy. The relevance of this trend is increasing in accordance with the objective requirements of the growing volumes of production and sale, the increasing complexity of economic relations, changing external environment of an enterprise.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-15

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

  4. The five essential ('key') elements of severe accident management. To be developed as part of a SAMG industry standard

    International Nuclear Information System (INIS)

    Vayssier, George

    2017-01-01

    The Fukushima-Daiichi accident has caused a renewed interest in tools and guidelines to mitigate severe accidents. Notably, industry approaches have been reviewed and features added from the lessons learned. The various severe accident management approaches vary considerably: they have different measures, different priorities for the various actions, different staff responsibilities and different sorts of communication to the off-site authorities. It appears that there is no common basis from which the approaches have been developed. In this paper, the five elements are treated which the author considers essential for proper tools to terminate severe accidents and mitigate their consequences. These five elements should be trained in well-developed drills/exercises, involving all functions of accident management. An industrial standard to define a minimum common basis, to which individual approaches should adhere and so decrease the large scatter in these approaches present now.

  5. The five essential ('key') elements of severe accident management. To be developed as part of a SAMG industry standard

    Energy Technology Data Exchange (ETDEWEB)

    Vayssier, George [NSC Netherlands, Hansweert (Netherlands)

    2017-07-15

    The Fukushima-Daiichi accident has caused a renewed interest in tools and guidelines to mitigate severe accidents. Notably, industry approaches have been reviewed and features added from the lessons learned. The various severe accident management approaches vary considerably: they have different measures, different priorities for the various actions, different staff responsibilities and different sorts of communication to the off-site authorities. It appears that there is no common basis from which the approaches have been developed. In this paper, the five elements are treated which the author considers essential for proper tools to terminate severe accidents and mitigate their consequences. These five elements should be trained in well-developed drills/exercises, involving all functions of accident management. An industrial standard to define a minimum common basis, to which individual approaches should adhere and so decrease the large scatter in these approaches present now.

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

    International Nuclear Information System (INIS)

    Beuzet, Emilie; Lamy, Jean-Sylvestre; Perron, Hadrien; Simoni, Eric; Ducros, Gérard

    2012-01-01

    Highlights: ► We developed a new modelling of fuel oxidation and ruthenium release in the EDF version of the MAAP4 code. ► We validated this model against some VERCORS experiments. ► Ruthenium release prediction quantitatively and qualitatively well reproduced under air and steam atmospheres. - Abstract: In a nuclear power plant (NPP), a severe accident is a low probability sequence that can lead to core fusion and fission product (FP) release to the environment (source term). For instance during a loss-of-coolant accident, water vaporization and core uncovery can occur due to decay heat. These phenomena enhance core degradation and, subsequently, molten materials can relocate to the lower head of the vessel. Heat exchange between the debris and the vessel may cause its rupture and air ingress. After lower head failure, steam and air entering in the vessel can lead to degradation and oxidation of materials that are still intact in the core. Indeed, Zircaloy-4 cladding oxidation is very exothermic and fuel interaction with the cladding material can decrease its melting temperature by several hundred of Kelvin. FP release can thus be increased, noticeably that of ruthenium under oxidizing conditions. Ruthenium is of particular interest because of its high radio-toxicity due to 103 Ru and 106 Ru isotopes and its ability to form highly volatile compounds, even at room temperature, such as gaseous ruthenium tetra-oxide (RuO 4 ). It is consequently of great need to understand phenomena governing steam and air oxidation of the fuel and ruthenium release as prerequisites for the source term issues. A review of existing data on these phenomena shows relatively good understanding. In terms of oxygen affinity, the fuel is oxidized before ruthenium, from UO 2 to UO 2+x . Its oxidation is a rate-controlling surface exchange reaction with the atmosphere, so that the stoichiometric deviation and oxygen partial pressure increase. High temperatures combined with the presence

  7. Hypothetical accident conditions free drop and thermal tests USA/5791/BLF (ERDA-AL)

    International Nuclear Information System (INIS)

    Blankenship, R.W.

    1980-05-01

    The USA/5791/BLF (ERDA-AL) shipping container with rolled-top food pack cans as inner containers is evaluated under conditions required by 10 CFR 71.42. One kilogram of depleted uranium as UO 2 was packaged in each of the inner containers. After completion of a free drop test and a simulated thermal test, the maximum observed leakage of UO 2 for the following week was 3.0 μg. This leakage is well below the allowable leakage per week for most plutonium isotopic mixtures. Using the examples provided, any plutonium isotopic mixture can be easily compared with the allowable leakage per week. Test conditions and results are reported

  8. Volume-heated boiling pool behavior and application to transition phase accident conditions

    International Nuclear Information System (INIS)

    Ginsberg, T.; Jones, O.C. Jr.; Chen, J.C.

    1978-01-01

    Observations of two-phase flow fields in volume-heated boiling pools are reported. Photographic observations, together with pool-average void fraction measurements are presented. Flow regime transition criterial derived from the measurements are discussed. The churn-turbulent flow regime was the dominant regime for superficial vapor velocity. Within this range of conditions, a churn-turbulent drift flux model provides a reasonable prediction of the pool-average void fraction data. The results of the experiment and analysis are extrapolated to transition phase conditions. It is shown that intense pool boil-up could occur where the pool-average void fraction would be greater than 0.6 for steel vaporization rates equivalent to power levels greater than one percent of nominal LMFBR power density. (author)

  9. Volume-heated boiling pool flow behavior and application to transition phase accident conditions

    International Nuclear Information System (INIS)

    Ginsberg, T.; Jones, O.C. Jr.; Chen, J.C.

    1978-01-01

    Observations of two-phase flow fields in volume-heated boiling pools are reported. Photographic observations, together with pool-average void fraction measurements are presented. Flow regime transition criteria derived from the measurements are discussed. The churn-turbulent flow regime was the dominant regime for superficial vapor velocities up to nearly five times the Kutateladze dispersal velocity. Within this range of conditions, a churn-turbulent drift flux model provides a reasonable prediction of the pool-average void fraction data. The results of the experiment and analyses are extrapolated to transition phase conditions. It is shown that intense pool boil-up could occur where the pool-average void fraction would be greater than 0.6 for steel vaporization rates equivalent to power levels greater than one percent of nominal LMFBR power density

  10. Hypothetical accident conditions, free drop and thermal tests: Specification 6M

    International Nuclear Information System (INIS)

    Blankenship, R.W.

    1980-05-01

    The 30 gallon Specification 6M shipping container with rolled-top food pack cans as inner containers is evaluated under conditions required by 10 CFR 71.42. One kilogram of depleted uranium as UO 2 was packaged in each of the inner containers. After completion of a free drop test and a simulated thermal test, the maximum observed leakage of UO 2 for the following week was 3.2 μg. This leakage is well below the allowable leakage per week for most plutonium isotopic mixtures. Using the examples provided, any plutonium isotopic mixture can be easily compared with the allowable leakage per week. Test conditions and results are reported

  11. SSYST, Modular System for Transient Fuel Rod Behaviour Under Accident Condition

    International Nuclear Information System (INIS)

    Gulden, W.; Meyder, R.; Borgwaldt, H.

    1987-01-01

    1 - Description of problem or function: SSYST is a code system for analyzing transient fuel rod behaviour under off-normal conditions, developed jointly by the Institut fuer Kernenergetik und Energie-systeme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract for the Projekt Nukleare Sicherheit (PNS) at KfK. Main differences versus codes with similar applications are: (1) an open-ended modular code organisation; (2) a preference for simple models, wherever possible. While feature (1) makes SSYST a very flexible tool, easily adapted to changing requirements, feature (2) leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 minutes CPU time on IBM 3033, so that extensive parametric studies are feasible. Main differences between SSYST-3 and previous versions are related to a general clean-up of the code system, which reduces the implementation effort: - advanced modules for cladding deformation and oxidation and reflooding conditions are included; - an input processor thoroughly checks all input data

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

    Directory of Open Access Journals (Sweden)

    Ikanović Jela

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Lambert, L.D.; Parks, M.B.

    1994-09-01

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

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

    International Nuclear Information System (INIS)

    Busser, Vincent

    2009-01-01

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

  15. Behavior of defective LWR-type fuel rods irradiated under postulated accident conditions

    International Nuclear Information System (INIS)

    Hobbins, R.R.; Croucher, D.W.; Seiffert, S.L.; Cook, B.A.; Kerwin, D.K.; Mehner, A.S.; Ploger, S.A.

    1979-05-01

    The irradiation experiments reported here have been conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., for the United States Nuclear Regulatory Commission in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Five of the rods were irradiated in PCM tests and one in a LOC test. During these tests, the six rods lost cladding integrity prior to or during the transient phase of the test due to either manufacturing defects or intentional rod design and operation. Of the five defective rods tested under PCM conditions, one (Rod IE-008, Test IE-1) had a hydride rupture below the region of the rod, which was in film boiling during the transient; two (Rod A-0021, Test PCM-3 and Rod IE-019, Test IE-5) contained defects (a pin hole and a small axial crack, respectively) within the film boiling zone; and two (Rod 201-1, Test PCM-1 and Rod 205-8, Test PCM-5) failed by cladding embrittlement within the film boiling zone. Rod 312-3 was waterlogged before being subjected to LOC conditions in Test LLR-3

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

    International Nuclear Information System (INIS)

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

    1990-06-01

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

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

    International Nuclear Information System (INIS)

    Bhawal, R.N.; Bajaj, S.S.

    1998-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-06-01

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

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

    International Nuclear Information System (INIS)

    1989-09-01

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

  20. KUPOL-M code for simulation of the VVER's accident localization system under LOCA conditions

    International Nuclear Information System (INIS)

    Efanov, A.D.; Lukyanov, A.A.; Shangin, N.N.; Zajtsev, A.A.; Solov'ev, S.L.

    2004-01-01

    Computer code KUPOL-M is developed for analysis of thermodynamic parameters of medium within full pressure containment for NPPs with VVER under LOCA conditions. The analysis takes into account the effects of non-stationary heat-mass transfer of gas-drop mixture in the containment compartments with natural convection, volume and surface steam condensation in the presence of noncondensables, heat-mass exchange of the compartment atmosphere with water in the sumps. The operation of the main safety systems like a spray system, hydrogen catalytic recombiners, emergency core cooling pumps, valves and a fan system is simulated in KUPOL-M code. The main results of the code verification including the ones of the participation in ISP-47 International Standard Problem on containment thermal-hydraulics are presented. (author)

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

    International Nuclear Information System (INIS)

    Cheng, S.C.; Groeneveld, D.C.; Leung, L.K.H.; Wong, Y.L.; Nguyen, C.

    1987-07-01

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

  2. Study on the behavior of waterside corroded PWR fuel rods under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Sasajima, Hideo

    1989-06-01

    One of the highlighted problems from the fuel reliability point of view is a waterside corrosion of fuel cladding which becomes more significant at extended burnup stages. To date, at highly burned fuel, waterside corrosion was recognized as important because cladding oxidation increased with increasing burn-up. In experiments, as the basic research for the study of high burn-up fuel, the test fuel rods were prepressurized to ranges from 3.47 to 3.55 MPa, oxidized artificially to both 10 and 20 μm in thickness. Regarding fabricated oxide thickness of 10 μm, it is corresponded to be transition point from cubic law to linear law as a function of burn-up. Pulse irradiation experiments by NSRR were carried out to study the behavior of waterside corroded PWR type fuels under RIA conditions. Obtained results are: (1) The failure threshold of tested fuels was 110 cal/g·fuel (0.46 KJ/g·fuel) in enthalpy. This showed that the failure threshold of tested fuels was same as that of the past NSRR experimental data. (2) The failure mechanisms of the tested fuel rods was cladding rupture induced by ballooning. No differences in failure mechanisms existed between the past NSRR prepressurized standard fuel and the tested fuels. (3) Cracks were existed without propagating into cladding matrix, so that it was judged that these were not initiation of failure. (4) Whithin this experimental condition, reduction of cladding thickness being attributed to the increase of oxidation did not failure threshold. (author)

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

    International Nuclear Information System (INIS)

    Cheng, S.C.; Groeneveld, D.C.; Leung, L.K.H.; Wong, Y.L.; Nguyen, C.

    1987-07-01

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

  4. Thermodynamic evaluation of the solidification phase of molten core–concrete under estimated Fukushima Daiichi nuclear power plant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kitagaki, Toru, E-mail: kitagaki.toru@jaea.go.jp; Yano, Kimihiko; Ogino, Hideki; Washiya, Tadahiro

    2017-04-01

    The solidification phases of molten core–concrete under the estimated molten core–concrete interaction (MCCI) conditions in the Fukushima Daiichi Nuclear Power Plant Unit 1 were predicted using the thermodynamic equilibrium calculation tool, FactSage 6.2, and the NUCLEA database in order to contribute toward the 1F decommissioning work and to understand the accident progression via the analytical results for the 1F MCCI products. We showed that most of the U and Zr in the molten core–concrete forms (U,Zr)O{sub 2} and (Zr,U)SiO{sub 4}, and the formation of other phases with these elements is limited. However, the formation of (Zr,U)SiO{sub 4} requires a relatively long time because it involves a change in the crystal structure from fcc-(U,Zr)O{sub 2} to tet-(U,Zr)O{sub 2}, followed by the formation of (Zr,U)SiO{sub 4} by reaction with SiO{sub 2}. Therefore, the formation of (Zr,U)SiO{sub 4} is limited under quenching conditions. Other common phases are the oxide phases, CaAl{sub 2}Si{sub 2}O{sub 8}, SiO{sub 2}, and CaSiO{sub 3}, and the metallic phases of the Fe–Si and Fe–Ni alloys. The solidification phenomenon of the crust under quenching conditions and that of the molten pool under thermodynamic equilibrium conditions in the 1F MCCI progression are discussed.

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

    International Nuclear Information System (INIS)

    Watanabe, Atsushi; Nagasaka, Hideo; Yokobori, Seiichi; Akinaga, Makoto

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Ohai, Dumitru

    2007-01-01

    Full text: Depending on the residence time, fuel burnup, and fuel rack configuration, there may be sufficient decay heat for the fuel clad to heat up, swell, and burst in case of a loss of pool water. Initiating event categories can be: loss of offsite power from events initiated by severe weather, internal fire, loss of pool cooling, loss of coolant inventory, seismic event, aircraft impact, tornado, missile attack. The breach in the clad releases the radioactive gases present in the gap between the fuel and clad, what is called 'gap release'. If the fuel continues to heat up, the zirconium clad will reach the point of rapid oxidation in air. This reaction of zirconium and air, or zirconium and steam is exothermic. The energy released from the reaction, combined with the fuel's decay energy, can cause the reaction to become self-sustaining and ignite the zirconium. The increase in heat from the oxidation reaction can also raise the temperature in adjacent fuel assemblies and propagate the oxidation reaction. Simultaneously, the sintered UO 2 pellets resulting from pins destroying are oxidized. Due to the self-disintegration of pellets by oxidation, fission gases and low volatile fission products are released. The release rate, the chemical nature and the amount of fission products depend on powder granulation distribution and environmental conditions. The zirconium burning and pellets self-disintegration will result in a significant release of spent fuel fission products that will be dispersed from the reactor site. (author)

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

    International Nuclear Information System (INIS)

    Otomo, Takashi; Hashimoto, Masao; Kawasaki, Satoru; Furuta, Teruo; Uetsuka, Hiroshi

    1983-06-01

    Multi-rod burst test of No.7808 bundle was performed in steam to estimate quantitative coolant flow channel restriction caused by the ballooning of zircaloy claddings in a fuel assembly during a LOCA transient in LWRs. The test was conducted under the condition that the initial internal pressure in each rod was 35kg/cm 2 (RT) and the heating rate was 9 0 C/s in steam with flow rate of 0.4g/cm 2 .min. The following results were obtained; (1) Maximum and burst pressures in rods were in the range 45 to 48kg/cm 2 and 41 to 45kg/cm 2 , respectively. The burst temperature of cladding were estimated to be 850 to 880 0 C. (2) Axial portions of tubes with greater than 34% strain were observed in the range 0 to 40mm in most rod. The mean length was 19mm in the bundle. (3) The degree of maximum increase in cross-sectional area is 54.2% in the bundle(7 x 7) and 66.9% in the internal rods(5 x 5). (4) Maximum channel area restriction was 40.5% in the bundle(7 x 7) and 51.4% in the internal rods(5 x 5). (author)

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

    International Nuclear Information System (INIS)

    Karwat, H.

    1992-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Karwat, H

    1992-08-15

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

  10. Accident conditions analysis of spent fuel storage pool RA research reactor in Vinca; Analiza udesnih stanja u odlagalistu isluzenog goriva istrazivackog rektora RA u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Jovic, V; Jovic, L [Institute of Nuclear Sciences VINCA, Belgrade (Serbia and Montenegro)

    2000-07-01

    Based on Safety analysis of the spent fuel pool RA research reactor in Vinca, conditions and possibilities accident sequences in present configuration storage facility are considered (author) [Serbo-Croat] Na osnovu Analize sigurnosti odlagalista isluzenog goriva istrazivackog reaktora RA u Vinci razmatraju se uslovi i mogucnosti pojave udesnih stanja u postojecoj konfiguraciji odlagalista (author)

  11. Abnormal Signal Analysis for a Change of the R-C Passive Elements in a Equivalent Circuit Modeling under a High Temperature Accident Condition

    International Nuclear Information System (INIS)

    Koo, Kil-Mo; Song, Yong-Mann; Ahan, Kwang-Il; Ha, Jea-Joo

    2007-01-01

    An electrical signal should be checked to see whether it lies within its expected electrical range when there is a doubtful condition. The normal signal level for pressure, flow, level and resistance temperature detector sensors is 4 - 20mA for most instruments as an industrial process control standard. In the case of an abnormal signal level from an instrument under a severe accident condition, it is necessary to obtain a more accurate signal validation to operate a system in a control room in NPPs. Diagnostics and analysis for some abnormal signals have been performed through an important equivalent circuits modeling for passive elements under severe accident conditions. Unlike the design basis accidents, there are some inherent uncertainties for the instrumentation capabilities under severe accident conditions. In this paper, to implement a diagnostic analysis for an equivalent circuits modeling, a kind of linked LabVIEW program for each PSpice and MULTISim code is introduced as a one body order system, which can obtain some abnormal signal patterns by a special function such as an advanced simulation tool for each PSpice and Multi-SIM code as a means of a function for a PC based ASSA (abnormal signal simulation analyzer) module

  12. Abnormal Signal Analysis for a Change of the R-C Passive Elements in a Equivalent Circuit Modeling under a High Temperature Accident Condition

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Kil-Mo; Song, Yong-Mann; Ahan, Kwang-Il; Ha, Jea-Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    An electrical signal should be checked to see whether it lies within its expected electrical range when there is a doubtful condition. The normal signal level for pressure, flow, level and resistance temperature detector sensors is 4 - 20mA for most instruments as an industrial process control standard. In the case of an abnormal signal level from an instrument under a severe accident condition, it is necessary to obtain a more accurate signal validation to operate a system in a control room in NPPs. Diagnostics and analysis for some abnormal signals have been performed through an important equivalent circuits modeling for passive elements under severe accident conditions. Unlike the design basis accidents, there are some inherent uncertainties for the instrumentation capabilities under severe accident conditions. In this paper, to implement a diagnostic analysis for an equivalent circuits modeling, a kind of linked LabVIEW program for each PSpice and MULTISim code is introduced as a one body order system, which can obtain some abnormal signal patterns by a special function such as an advanced simulation tool for each PSpice and Multi-SIM code as a means of a function for a PC based ASSA (abnormal signal simulation analyzer) module.

  13. [Accidents and injuries at work].

    Science.gov (United States)

    Standke, W

    2014-06-01

    In the case of an accident at work, the person concerned is insured by law according to the guidelines of the Sozialgesetzbuch VII as far as the injuries have been caused by this accident. The most important source of information on the incident in question is the accident report that has to be sent to the responsible institution for statutory accident insurance and prevention by the employer, if the accident of the injured person is fatal or leads to an incapacity to work for more than 3 days (= reportable accident). Data concerning accidents like these are sent to the Deutsche Gesetzliche Unfallversicherung (DGUV) as part of a random sample survey by the institutions for statutory accident insurance and prevention and are analyzed statistically. Thus the key issues of accidents can be established and used for effective prevention. Although the success of effective accident prevention is undisputed, there were still 919,025 occupational accidents in 2011, with clear gender-related differences. Most occupational accidents involve the upper and lower extremities. Accidents are analyzed comprehensively and the results are published and made available to all interested parties in an effort to improve public awareness of possible accidents. Apart from reportable accidents, data on the new occupational accident pensions are also gathered and analyzed statistically. Thus, additional information is gained on accidents with extremely serious consequences and partly permanent injuries for the accident victims.

  14. Accidents (FARS) (National)

    Data.gov (United States)

    Department of Transportation — Accident - (1975-current): This data file (NTAD) contains information about crash characteristics and environmental conditions at the time of the crash. There is one...

  15. Accidents - Chernobyl accident

    International Nuclear Information System (INIS)

    2004-01-01

    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.)

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

    International Nuclear Information System (INIS)

    Kim, H. D.; Kim, D. H.; Park, S. Y.; Park, J. H.

    2005-10-01

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

  17. Feasibility study of superconducting power cables for DC electric railway feeding systems in view of thermal condition at short circuit accident

    Science.gov (United States)

    Kumagai, Daisuke; Ohsaki, Hiroyuki; Tomita, Masaru

    2016-12-01

    A superconducting power cable has merits of a high power transmission capacity, transmission losses reduction, a compactness, etc., therefore, we have been studying the feasibility of applying superconducting power cables to DC electric railway feeding systems. However, a superconducting power cable is required to be cooled down and kept at a very low temperature, so it is important to reveal its thermal and cooling characteristics. In this study, electric circuit analysis models of the system and thermal analysis models of superconducting cables were constructed and the system behaviors were simulated. We analyzed the heat generation by a short circuit accident and transient temperature distribution of the cable to estimate the value of temperature rise and the time required from the accident. From these results, we discussed a feasibility of superconducting cables for DC electric railway feeding systems. The results showed that the short circuit accident had little impact on the thermal condition of a superconducting cable in the installed system.

  18. Safety analysis methodology for Chinshan nuclear power plant spent fuel pool under Fukushima-like accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Hao-Tzu [Institute of Nuclear Energy Research, Taoyuan, Taiwan (China). Research Atomic Energy Council; Li, Wan-Yun; Wang, Jong-Rong; Tseng, Yung-Shin; Chen, Hsiung-Chih; Shih, Chunkuan; Chen, Shao-Wen [National Tsing Hua Univ., HsinChu, Taiwan (China). Inst. of Nuclear Engineering and Science

    2017-03-15

    Chinshan nuclear power plant (NPP), a BWR/4 plant, is the first NPP in Taiwan. After Fukushima NPP disaster occurred, there is more concern for the safety of NPPs in Taiwan. Therefore, in order to estimate the safety of Chinshan NPP spent fuel pool (SFP), by using TRACE, MELCOR, CFD, and FRAPTRAN codes, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of Chinshan NPP SFP. There were two main steps in this research. The first step was the establishment of Chinshan NPP SFP models. And the transient analysis under the SFP cooling system failure condition (Fukushima-like accident) was performed. In addition, the sensitive study of the time point for water spray was also performed. The next step was the fuel rod performance analysis by using FRAPTRAN and TRACE's results. Finally, the animation model of Chinshan NPP SFP was presented by using the animation function of SNAP with MELCOR analysis results.

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

    International Nuclear Information System (INIS)

    Conte, M.C.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Andrews, W.S.; Lewis, B.J.; Cox, D.S.

    1997-01-01

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

  1. Driving force of PCMI failure under reactivity initiated accident conditions and influence of hydrogen embrittlement on failure limit

    International Nuclear Information System (INIS)

    Tomiyasu, Kunihiko; Sugiyama, Tomoyuki; Nakamura, Takehiko; Fuketa, Toyoshi

    2005-09-01

    In order to clarify the driving force of PCMI (Pellet/Cladding Mechanical Interaction) failure on high burnup fuels and to investigate the influence of hydrogen embrittlement on failure limit under RIA (Reactivity Initiated Accident) conditions, RIA-simulation experiments were performed on fresh fuel rods in the NSRR (Nuclear Safety Research Reactor). The driving force of PCMI was restricted only to thermal expansion of pellet by using fresh UO 2 pellets. Fresh claddings were pre-hydrided to simulate hydrogen absorption of high burnup fuel rods. In seven experiments out of fourteen, test rods resulted in PCMI failure, which has been observed in the NSRR tests on high burnup PWR fuels, in terms of the transient behavior and the fracture configuration. This indicates that the driving force of PCMI failure is sufficiently explained with thermal expansion of pellet and a contribution of fission gas on it is small. A large number of incipient cracks were generated in the outer surface of the cladding even on non-failed fuel rods, and they stopped at the boundary between hydride rim, which was a hydride layer localized in the periphery of the cladding, and metallic layer. It suggests that the integrity of the metallic layer except for the hydride rim has particular importance for failure limit. Fuel enthalpy at failure correlates with the thickness of hydride rim, and tends to decrease with thicker hydride layer. (author)

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

    International Nuclear Information System (INIS)

    1996-07-01

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

  3. Postulated accidents

    International Nuclear Information System (INIS)

    Ullrich, W.

    1980-01-01

    This lecture on 'Postulated Accidents' is the first of a series of lectures on the dynamic and transient behaviour of nuclear power plants, especially pressurized water reactors. The main points covered will be: Reactivity Accidents, Transients (Intact Loop) and Loss of Cooland Accidents (LOCA) including small leak. This lecture will discuss the accident analysis in general, the definition of the various operational phases, the accident classification, and, as an example, an accident sequence analysis on the basis of 'Postulated Accidents'. (orig./RW)

  4. An Innovative Continuing Nursing Education Program Targeting Key Geriatric Conditions for Hospitalized Older People in China

    Science.gov (United States)

    Xiao, Lily Dongxia; Shen, Jun; Wu, Haifeng; Ding, Fu; He, Xizhen; Zhu, Yueping

    2013-01-01

    A lack of knowledge in registered nurses about geriatric conditions is one of the major factors that contribute to these conditions being overlooked in hospitalized older people. In China, an innovative geriatric continuing nursing education program aimed at developing registered nurses' understanding of the complex care needs of hospitalized…

  5. The mutation frequency of Drosophila melanogaster populations living under conditions of increased background radiation due to the Chernobyl accident

    International Nuclear Information System (INIS)

    Zainullin, V.G.; Rakin, A.O.; Shevchenko, V.A.; Myasnyankina, E.N.; Generalova, M.V.

    1992-01-01

    One of the problems facing the program in the wake of the Chernobyl accident is the estimation of genetic damage to plants and animals. Special attention was directed to studying the influence of radioactive pollutants at the accident site by means of an appropriate test system, using standard genetic subjects. The present study describes such investigations. Levels of persistent genetic damage in natural populations of Drosophila melanogaster found in the vicinity of the Chernobyl accident site were examined from August 1986-September 1989. Evidence is presented which indicates a relationship between the levels of radioactive pollution resulting from the Chernobyl accident and increasing genetic damage to exposed populations. The possible reasons for the decrease of mutation frequency observed in 1988 and 1989 are also discussed. Furthermore, evidence is presented which suggests that radiosensitive Drosophila mutants may be particularly sensitive indicators of radioactive pollution. (author). 16 refs.; 6 figs

  6. Robust and predictive fuzzy key performance indicators for condition-based treatment of squats in railway infrastructures

    NARCIS (Netherlands)

    Jamshidi, A.; Nunez Vicencio, Alfredo; Dollevoet, R.P.B.J.; Li, Z.

    2017-01-01

    This paper presents a condition-based treatment methodology for a type of rail surface defect called squat. The proposed methodology is based on a set of robust and predictive fuzzy key performance indicators. A fuzzy Takagi-Sugeno interval model is used to predict squat evolution for different

  7. [Assessment on the changing conditions of ecosystems in key ecological function zones in China].

    Science.gov (United States)

    Huang, Lin; Cao, Wei; Wu, Dan; Gong, Guo-li; Zhao, Guo-song

    2015-09-01

    In this paper, the dynamics of ecosystem macrostructure, qualities and core services during 2000 and 2010 were analyzed for the key ecological function zones of China, which were classified into four types of water conservation, soil conservation, wind prevention and sand fixation, and biodiversity maintenance. In the water conservation ecological function zones, the areas of forest and grassland ecosystems were decreased whereas water bodies and wetland were increased in the past 11 years, and the water conservation volume of forest, grassland and wetland ecosystems increased by 2.9%. This region needs to reverse the decreasing trends of forest and grassland ecosystems. In the soil conservation ecological function zones, the area of farmland ecosystem was decreased, and the areas of forest, grassland, water bodies and wetland ecosystems were increased. The total amount of the soil erosion was reduced by 28.2%, however, the soil conservation amount of ecosystems increased by 38.1%. In the wind prevention and sand fixation ecological function zones, the areas of grassland, water bodies and wetland ecosystems were decreased, but forest and farmland ecosystems were increased. The unit amount of the soil. wind erosion was reduced and the sand fixation amount of ecosystems increased lightly. In this kind of region that is located in arid and semiarid areas, ecological conservation needs to reduce farmland area and give priority to the protection of the original ecological system. In the biodiversity maintenance ecological function zones, the areas of grassland and desert ecosystems were decreased and other types were increased. The human disturbances showed a weakly upward trend and needs to be reduced. The key ecological function zones should be aimed at the core services and the protecting objects, to assess quantitatively on the effectiveness of ecosystem conservation and improvement.

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

    International Nuclear Information System (INIS)

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

    1992-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Stempniewicz, M; Marks, P; Salwa, K

    1992-06-01

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

  10. Phosphatidylinositol 3-kinase is a key mediator of central sensitization in painful inflammatory conditions

    Science.gov (United States)

    Pezet, Sophie; Marchand, Fabien; D'Mello, Richard; Grist, John; Clark, Anna K.; Malcangio, Marzia; Dickenson, Anthony H.; Williams, Robert J.; McMahon, Stephen B.

    2010-01-01

    Here we show that phosphatidylinositol 3-kinase (PI3K) is a key player in the establishment of central sensitization, the spinal cord phenomenon associated with persistent afferent inputs and contributing to chronic pain states. We demonstrated electrophysiologically that PI3K is required for the full expression of spinal neuronal wind-up. In an inflammatory pain model, intrathecal administration of LY294002, a potent PI3K inhibitor, dose-dependently inhibited pain related behavior. This effect was correlated with a reduction of the phosphorylation of extracellular signal-regulated kinase (ERK) and CaMKinase II. In addition, we observed a significant decrease in the phosphorylation of the NMDA receptor subunit NR2B, decreased translocation to the plasma membrane of the GluR1 AMPA receptor subunit in the spinal cord and a reduction of evoked neuronal activity as measured using c-Fos immunohistochemistry. Our study suggests that PI3K is a major factor in the expression of central sensitization after noxious inflammatory stimuli. PMID:18417706

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

    Directory of Open Access Journals (Sweden)

    Neubauer Peter

    2011-05-01

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

  12. Increased fitness of a key appendicularian zooplankton species under warmer, acidified seawater conditions.

    Directory of Open Access Journals (Sweden)

    Jean-Marie Bouquet

    Full Text Available Ocean warming and acidification (OA may alter the fitness of species in marine pelagic ecosystems through community effects or direct physiological impacts. We used the zooplanktonic appendicularian, Oikopleura dioica, to assess temperature and pH effects at mesocosm and microcosm scales. In mesocosms, both OA and warming positively impacted O. dioica abundance over successive generations. In microcosms, the positive impact of OA, was observed to result from increased fecundity. In contrast, increased pH, observed for example during phytoplankton blooms, reduced fecundity. Oocyte fertility and juvenile development were equivalent under all pH conditions, indicating that the positive effect of lower pH on O. dioica abundance was principally due to increased egg number. This effect was influenced by food quantity and quality, supporting possible improved digestion and assimilation at lowered pH. Higher temperature resulted in more rapid growth, faster maturation and earlier reproduction. Thus, increased temperature and reduced pH had significant positive impacts on O. dioica fitness through increased fecundity and shortened generation time, suggesting that predicted future ocean conditions may favour this zooplankton species.

  13. The Key Role of Government in Addressing the Pandemic of Micronutrient Deficiency Conditions in Southeast Asia

    Directory of Open Access Journals (Sweden)

    Theodore H. Tulchinsky

    2015-04-01

    Full Text Available Micronutrient deficiency conditions are a major global public health problem. While the private sector has an important role in addressing this problem, the main responsibility lies with national governments, in cooperation with international agencies and donors. Mandatory fortification of basic foods provides a basic necessary intake for the majority and needs to be supported by provision of essential vitamin and mineral supplements for mothers and children and other high risk groups. Fortification by government mandate and regulation is essential with cooperation by private sector food manufacturers, and in the context of broader policies for poverty reduction, education and agricultural reform. Iron, iodine, vitamin A, vitamin B complex, folic acid, zinc, vitamin D and vitamin B12 are prime examples of international fortification experience achieved by proactive governmental nutrition policies. These are essential to achieve the Millennium Development Goals and their follow-up sustainable global health targets. National governmental policies for nutritional security and initiatives are essential to implement both food fortification and targeted supplementation policies to reduce the huge burden of micronutrient deficiency conditions in Southeast Asia and other parts of the world.

  14. Consideration of Command and Control Performance during Accident Management Process at the Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Nisrene M. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Sok Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    The accident at the Fukushima Daiichi nuclear power plants shifted the nuclear safety paradigm from risk management to on-site management capability during a severe accident. The kernel of on-site management capability during an accident at a nuclear power plant is situation awareness and agility of command and control. However, little consideration has been given to accident management. After the events of September 11, 2001 and the catastrophic Fukushima nuclear disaster, agility of command and control has emerged as a significant element for effective and efficient accident management, with many studies emphasizing accident management strategies, particularly man-machine interface, which is considered a key role in ensuring nuclear power plant safety during severe accident conditions. This paper proposes a conceptual model for evaluating command and control performance during the accident management process at a nuclear power plant. Communication and information processing while responding to an accident is one of the key issues needed to mitigate the accident. This model will give guidelines for accurate and fast communication response during accident conditions.

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

    International Nuclear Information System (INIS)

    Krajnc, B.

    1995-01-01

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

  16. Development and Application of the Key Technologies for the Quality Control and Inspection of National Geographical Conditions Survey Products

    Science.gov (United States)

    Zhao, Y.; Zhang, L.; Ma, W.; Zhang, P.; Zhao, T.

    2018-04-01

    The First National Geographical Condition Survey is a predecessor task to dynamically master basic situations of the nature, ecology and human activities on the earth's surface and it is the brand-new mapping geographic information engineering. In order to ensure comprehensive, real and accurate survey results and achieve the quality management target which the qualified rate is 100 % and the yield is more than 80 %, it is necessary to carry out the quality control and result inspection for national geographical conditions survey on a national scale. To ensure that achievement quality meets quality target requirements, this paper develops the key technology method of "five-in-one" quality control that is constituted by "quality control system of national geographical condition survey, quality inspection technology system, quality evaluation system, quality inspection information management system and national linked quality control institutions" by aiming at large scale, wide coverage range, more undertaking units, more management levels, technical updating, more production process and obvious regional differences in the national geographical condition survey and combining with novel achievement manifestation, complicated dependency, more special reference data, and large data size. This project fully considering the domestic and foreign related research results and production practice experience, combined with the technology development and the needs of the production, it stipulates the inspection methods and technical requirements of each stage in the quality inspection of the geographical condition survey results, and extends the traditional inspection and acceptance technology, and solves the key technologies that are badly needed in the first national geographic survey.

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

    International Nuclear Information System (INIS)

    Funayama, Kyoko; Kajimoto, Mitsuhiro; Nagayoshi, Takuji; Tanaka, Nobuo

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Faoro, Franco; Iriti, Marcello

    2009-01-01

    An account of histo-cytological and ultrastructural studies on ozone effect on crop and forest species in Italy is given, with emphasis on induced cell death and the underlying mechanisms. Cell death phenomena possibly due to ambient O 3 were recorded in crop and forest species. In contrast, visible O 3 effects on Mediterranean vegetation are often unclear. Microscopy is thus suggested as an effective tool to validate and evaluate O 3 injury to Mediterranean vegetation. A DAB-Evans blue staining was proposed to validate O 3 symptoms at the microscopic level and for a pre-visual diagnosis of O 3 injury. The method has been positively tested in some of the most important crop species, such as wheat, tomato, bean and onion and, with some restriction, in forest species, and it also allows one to gain some very useful insights into the mechanisms at the base of O 3 sensitivity or tolerance. - Ozone-induced cell death is a frequent phenomenon in Mediterranean conditions, not only in the most sensitive crops but also in forest species.

  20. Effects of Different Drying Conditions on Key Quality Parameters of Pink Peppercorns (Schinus terebinthifolius Raddi

    Directory of Open Access Journals (Sweden)

    Bruno Guzzo Silva

    2017-01-01

    Full Text Available Pink peppercorns are among the most sophisticated condiments in the international cuisine. This culinary spice is obtained from dried fruits of Schinus terebinthifolius Raddi, a species native to South America. In this work, a methodology for the assessment of pink peppercorn quality under various drying conditions was defined. Experiments were performed in a pilot tray dryer, which ensured integrity of the product. A central composite rotatable design with 11 experiments was devised to study the influence of drying air temperature (35–75°C and air velocity (0.3–0.9 m/s on product quality, assessed by moisture content, color (CIELAB system, and volatile compounds. The essential oils of fresh and dried fruits were extracted by hydrodistillation and analyzed by gas chromatography coupled to mass spectrometry. Air temperature had the greatest influence on the quality parameters under study, while air velocity had no statistically significant effect. Considering all quality criteria, temperatures between 40 and 55°C provided the best compromise, yielding an adequate moisture content in the dried product without dramatic degradation of color and essential oil.

  1. Overview of core disruptive accidents

    International Nuclear Information System (INIS)

    Marchaterre, J.F.

    1977-01-01

    An overview of the analysis of core-disruptive accidents is given. These analyses are for the purpose of understanding and predicting fast reactor behavior in severe low probability accident conditions, to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features. The methods are used to analyze core-disruptive accidents from initiating event to complete core disruption, the effects of the accident on reactor structures and the resulting radiological consequences are described

  2. Accident management

    International Nuclear Information System (INIS)

    Lutz, R.J.; Monty, B.S.; Liparulo, N.J.; Desaedeleer, G.

    1989-01-01

    The foundation of the framework for a Severe Accident Management Program is the contained in the Probabilistic Safety Study (PSS) or the Individual Plant Evaluations (IPE) for a specific plant. The development of a Severe Accident Management Program at a plant is based on the use of the information, in conjunction with other applicable information. A Severe Accident Management Program must address both accident prevention and accident mitigation. The overall Severe Accident Management framework must address these two facets, as a living program in terms of gathering the evaluating information, the readiness to respond to an event. Significant international experience in the development of severe accident management programs exist which should provide some direction for the development of Severe Accident Management in the U.S. This paper reports that the two most important elements of a Severe Accident Management Program are the Emergency Consultation process and the standards for measuring the effectiveness of individual Severe Accident Management Programs at utilities

  3. Unavoidable Accident

    OpenAIRE

    Grady, Mark F.

    2009-01-01

    In negligence law, "unavoidable accident" is the risk that remains when an actor has used due care. The counterpart of unavoidable accident is "negligent harm." Negligence law makes parties immune for unavoidable accident even when they have used less than due care. Courts have developed a number of methods by which they "sort" accidents to unavoidable accident or to negligent harm, holding parties liable only for the latter. These sorting techniques are interesting in their own right and als...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-06-01

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

  6. DEVELOPMENT AND APPLICATION OF THE KEY TECHNOLOGIES FOR THE QUALITY CONTROL AND INSPECTION OF NATIONAL GEOGRAPHICAL CONDITIONS SURVEY PRODUCTS

    Directory of Open Access Journals (Sweden)

    Y. Zhao

    2018-04-01

    Full Text Available The First National Geographical Condition Survey is a predecessor task to dynamically master basic situations of the nature, ecology and human activities on the earth’s surface and it is the brand-new mapping geographic information engineering. In order to ensure comprehensive, real and accurate survey results and achieve the quality management target which the qualified rate is 100 % and the yield is more than 80 %, it is necessary to carry out the quality control and result inspection for national geographical conditions survey on a national scale. To ensure that achievement quality meets quality target requirements, this paper develops the key technology method of “five-in-one” quality control that is constituted by “quality control system of national geographical condition survey, quality inspection technology system, quality evaluation system, quality inspection information management system and national linked quality control institutions” by aiming at large scale, wide coverage range, more undertaking units, more management levels, technical updating, more production process and obvious regional differences in the national geographical condition survey and combining with novel achievement manifestation, complicated dependency, more special reference data, and large data size. This project fully considering the domestic and foreign related research results and production practice experience, combined with the technology development and the needs of the production, it stipulates the inspection methods and technical requirements of each stage in the quality inspection of the geographical condition survey results, and extends the traditional inspection and acceptance technology, and solves the key technologies that are badly needed in the first national geographic survey.

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

    International Nuclear Information System (INIS)

    Mathe, Emmanuel

    2014-01-01

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

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

    Science.gov (United States)

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

    2014-02-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  10. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, Jason Dean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barani, Tommaso [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pizzocri, Davide [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

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

    International Nuclear Information System (INIS)

    Milovanovic, D.; Zlvkovic, Z.

    1996-01-01

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

  12. Description of the blowdown test facility COG program on in-reactor fission product release, transport, and deposition under severe accident conditions

    International Nuclear Information System (INIS)

    Fehrenbach, P.J.; Wood, J.C.

    1987-06-01

    Loss-of-coolant accidents with additional impairment of emergency cooling would probably result in high fuel temperatures leading to severe fuel damage (SFD) and significant fission product activity would then be transported along the PHTS to the break where a fraction of it would be released and transport under such conditions, there are many interacting and sometimes competing phenomena to consider. Laboratory simulations are being used to provide data on these individual phenomena, such as UO 2 oxidation and Zr-UO 2 interaction, from which mathematical models can be constructed. These are then combined into computer codes to include the interaction effects and assess the overall releases. In addition, in-reactor tests are the only source of data on release and transport of short-lived fission product nuclides, which are important in the consequence analysis of CANDU reactor accidents. Post-test decontamination of an in-reactor test facility also provides a unique opportunity to demonstrate techniques and obtain decontamination data relevant to post-accident rehabilitation of CANDU power reactors. Specialized facilities are required for in-reactor testing because of the extensive release of radioactive fission products and the high temperatures involved (up to 2500 degrees Celsius). To meet this need for the Canadian program, the Blowdown Test Facility (BTF) has been built in the NRU reactor at Chalk River. Between completion of construction in mid-1987 and the first Zircaloy-sheathed fuel test in fiscal year 1987/88, several commissioning tests are being performed. Similarly, extensive development work has been completed to permit application of instrumentation to irradiated fuel elements, and in support of post-test fuel assembly examination. A program of decontamination studies has also been developed to generate information relevant to post-accident decontamination of power reactors. The BTF shared cost test program funded by the COG High Temperature

  13. Severe accident analysis and management in nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Golshan, Mina

    2013-01-01

    Within the UK regulatory regime, assessment of risks arising from licensee's activities are expected to cover both normal operations and fault conditions. In order to establish the safety case for fault conditions, fault analysis is expected to cover three forms of analysis: design basis analysis (DBA), probabilistic safety assessment (PSA) and severe accident analysis (SAA). DBA should provide a robust demonstration of the fault tolerance of the engineering design and the effectiveness of the safety measures on a conservative basis. PSA looks at a wider range of fault sequences (on a best estimate basis) including those excluded from the DBA. SAA considers significant but unlikely accidents and provides information on their progression and consequences, within the facility, on the site and off site. The assessment of severe accidents is not limited to nuclear power plants and is expected to be carried out for all plant states where the identified dose targets could be exceeded. This paper sets out the UK nuclear regulatory expectation on what constitutes a severe accident, irrespective of the type of facility, and describes characteristics of severe accidents focusing on nuclear fuel cycle facilities. Key rules in assessment of severe accidents as well as the relationship to other fault analysis techniques are discussed. The role of SAA in informing accident management strategies and offsite emergency plans is covered. The paper also presents generic examples of scenarios that could lead to severe accidents in a range of nuclear fuel cycle facilities. (authors)

  14. Plant habitability assessment for Point Lepreau Generating Station during a severe accident resulting from station blackout conditions

    International Nuclear Information System (INIS)

    Mullin, D.

    2015-01-01

    In response to the CNSC Fukushima Action Plan, the CANDU Owners Group (COG) developed a methodology for assessing nuclear power plant habitability under Joint Project 4426 and to determine if any improvement actions are necessary to provide a high degree of assurance that a severe accident can be managed from a human and organizational performance perspective. NB Power has applied the methodology considering a station black-out scenario (representative case), and assessed the effects of non-radiological hazards and radiological hazards in the context of operator dose relative to emergency dose limits. The paper will discuss the overall methodology, findings and recommendations. (author)

  15. Plant habitability assessment for Point Lepreau Generating Station during a severe accident resulting from station blackout conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mullin, D., E-mail: dmullin@nbpower.com [New Brunswick Power Corporation, Point Lepreau Generating Station, Lepreau, NB (Canada)

    2015-07-01

    In response to the CNSC Fukushima Action Plan, the CANDU Owners Group (COG) developed a methodology for assessing nuclear power plant habitability under Joint Project 4426 and to determine if any improvement actions are necessary to provide a high degree of assurance that a severe accident can be managed from a human and organizational performance perspective. NB Power has applied the methodology considering a station black-out scenario (representative case), and assessed the effects of non-radiological hazards and radiological hazards in the context of operator dose relative to emergency dose limits. The paper will discuss the overall methodology, findings and recommendations. (author)

  16. Evaluation of Passive Containment Cooling System design of SMART built in GBS for ocean environment under the Fukushima Accident Condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Gil; Kim, Seong Gu; Lee, Jeong Ik; Lee, Kang Heon; Lee, Phil Seung [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    The ONPP can be towed to the installation site after the SMART is constructed within the GBS in a dry dock. And, by incorporating IPSS (Integrated Passive Safety System), proposed by KAIST, in the ocean SMART the safety of the whole system will be significantly increased which can potentially eliminate any possibility of repeating Fukushima accident again. In recent years, KAIST research team is developing a very advanced concept of ocean NPPs which can avoid natural disasters while potentially increasing economy and enhancing public acceptance. Authors chose Korean small reactor, SMART as a reference system to demonstrate the feasibility of the proposed ocean NPP. Ocean SMART is mounted on GBS (Gravity Based Structure)

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

    International Nuclear Information System (INIS)

    Deuber, H.

    1985-01-01

    No significant difference was found in the retention of I-131 loaded as I 2 , by various impregnated activated carbons that had been aged in the containment exhaust air of a pressurized water reactor over a period of 12 months. In all the cases, the I-131 passing through deep beds of carbon was in a nonelemental form. It was concluded that a minimum retention of 99.99%, as required by new guidelines for certain accident filters, can be equally well achieved with various carbons in deep beds

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

    International Nuclear Information System (INIS)

    An, S.; Togo, Y.

    1976-01-01

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

  19. Features of RAPTA-SFD code modelling of chemical interactions of basic materials of the WWER active zone in accident conditions with severe fuel damage

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Sokolov, N.B.; Salatov, A.V.; Nechaeva, O.A.; Andreyeva-Andrievskaya, L.N.; Vlasov, F.Yu.

    1996-01-01

    A brief description of RAPTA-SFD code intended for computer simulations of WWER-type fuel elements (simulator or absorber element) in conditions of accident with severe damage of fuel. Presented are models of chemical interactions of basic materials of the active zone, emphasized are special feature of their application in carrying out of the CORA-W2 experiment within the framework of International Standard Problem ISP-36. Results obtained confirm expediency of phenomenological models application. (author). 6 refs, 7 figs, 1 tab

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

    International Nuclear Information System (INIS)

    Sengpiel, W.

    1980-12-01

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

  1. Preventing accidents

    Science.gov (United States)

    2005-08-01

    As the most effective strategy for improving safety is to prevent accidents from occurring at all, the Volpe Center applies a broad range of research techniques and capabilities to determine causes and consequences of accidents and to identify, asses...

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

    International Nuclear Information System (INIS)

    Harada, Y.; Maruyama, Y.; Chino, E.; Shibazaki, H.; Kudo, T.; Hidaka, A.; Hashimoto, K.; Sugimoto, J.

    2000-01-01

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

  3. The role of grain boundary fission gases in high burn-up fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Lemoine, F.; Papin, J.; Frizonnet, J.M.; Cazalis, B.; Rigat, H.

    2002-01-01

    In the frame of reactivity-initiated accidents (RIA) studies, the CABRI REP-Na programme is currently performed, focused on high burn-up UO 2 and MOX fuel behaviour. From 1993 to 1998, seven tests were performed with UO 2 fuel and three with MOX fuel. In all these tests, particular attention has been devoted to the role of fission gases in transient fuel behaviour and in clad loading mechanisms. From the analysis of experimental results, some basic phenomena were identified and a better understanding of the transient fission gas behaviour was obtained in relation to the fuel and clad thermo-mechanical evolution in RIA, but also to the initial state of the fuel before the transient. A high burn-up effect linked to the increasing part of grain boundary gases is clearly evidenced in the final gas release, which would also significantly contribute to the clad loading mechanisms. (authors)

  4. Fission products and nuclear fuel behaviour under severe accident conditions part 2: Fuel behaviour in the VERDON-1 sample

    Science.gov (United States)

    Geiger, E.; Le Gall, C.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

    2017-11-01

    Within the framework of the International Source Term Programme (ISTP), the VERDON programme aims at quantifying the source term of radioactive materials in case of a hypothetical severe accident in a light water reactor (LWR). Tests were performed in a new experimental laboratory (VERDON) built in the LECA-STAR facility (CEA Cadarache). The VERDON-1 test was devoted to the study of a high burn-up UO2 fuel and FP releases at very high temperature (≈2873 K) in a reducing atmosphere. Post-test qualitative and quantitative characterisations of the VERDON-1 sample led to the proposal of a scenario explaining the phenomena occurring during the experimental sequence. Hence, the fuel and the cladding may have interacted which led to the melting of UO2-ZrO2 alloy. Although no relocation was observed during the test, it may have been imminent.

  5. Analysis of an out-of-pile experiment for materials redistribution under core disruptive accident condition of fast breeder reactors

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Ninokata, Hisashi; Shimizu, Akinao

    1995-01-01

    Calculation of one of the SIMBATH experiments was performed using the SIMMER-II code. The experiments were intended to simulate the fuel pin disintegration, the molten materials relocation and following materials redistribution that could occur during core disruptive accidents assumed in fast breeder reactors. The calculation by SIMMER-II showed that the incorporated step-wise fuel pin disintegration model and the modified particle jamming model were capable of reproducing the course of materials relocation within the identified ranges of the parameters which governed the blockages formation, i.e. the characteristic radius of solid particles jamming and/or sieving out in the flow and the effective particle viscosity. In particular the final materials redistribution calculated by SIMMER-II very well reproduced the experiment. This fact made it possible to interpret theoretically the mechanisms of flow blockages formation and related materials redistribution. (author)

  6. Status and results of the theoretical and experimental investigations on the LWR fuel rod behavior under accident conditions

    International Nuclear Information System (INIS)

    Bocek, M.; Hofmann, P.; Leistikow, S.; Class, G.; Meyder, R.; Raff, S.; Erbacher, F.; Hofmann, G.; Ihle, P.; Karb, E.; Fiege, A.

    1978-09-01

    In this report the status of knowledge is described which has been gathered up to the end of 1977 of the LWR fuel rod behavior in loss-of-coolant accidents. The majority of results indicated have been derived from studies on the fuel rod behavior performed within the framework of the Nuclear Safety Project (PNS); partly, also the results of cooperating research establishments and fm international exchange of experience are referred to. The report has been subdivided into two complete parts: Part I provides a survey of the most significant results of the theoretical and experimental research projects on fuel rod behavior. Part II describes by detailed individual presentations the status as well as the results with respect to the major central subjects. (orig.) 891 RW 892 AP [de

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

    International Nuclear Information System (INIS)

    Bastek, H.

    1984-01-01

    In the core of the high temperature reactor the fuel element and the coated particles contained herein provide the safest enclosure for fission products. The complex process of fission product transport out of the particle kernel, through the particle coating and within the fuel element graphite is described in a simplified form by the Fick's diffusion. The effective diffusion coefficient is used for calculation. Starting from the existing ideas of fission product transport five burn-up and temperature-dependent diffusion coefficients for Cesium in (Th,U)O 2 -kernels are derived in this study. The results have been gained from several fuel element radiation experiments in recent years, which showed extreme variation in regard to burn-up, temperature cycle, neutron flux and operation time. Cs-137 release measurements from single particle kernels were present from all the experiments. Furthermore, annealing tests of AVR-fuel elements were analyzed. Heat-temperatur and heating-time, the fuel element burn-up in the AVR-reactor, as well as the measured Cs-137 inventory of the fuel elements before and after annealing, are included in the investigation as essential parameters. With the aid of the derived diffusion coeffizients and already present data sets the Cs-137 release of fuel elements into a small reactor core is investigated under unrestricted core heat-up. While the released Cs-137 is derived mainly from defective particles at accident temperatures up to 1600 0 C, the main part diffuses through the particle coating at higher accident temperatures. (orig./HP) [de

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

    International Nuclear Information System (INIS)

    Fedotov, A.

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Hyman, C.R.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Alonso, A.; Gonzalez, C.

    1991-07-01

    This state of the art report contains information on the behaviour of tellurium and its compounds in the reactor pressure vessel and the reactor coolant system under light water reactor severe accident conditions. To characterise tellurium behaviour, it is necessary the previous knowledge of the species of tellurium released from the core, and simultaneity of its release with that of other materials which can alter the transport, for instance, control rod and structural materials. Release and transport experiments have been reviewed along with the models implemented in the codes which are used in the international community: TRAPMELT, RAFT, VICTORIA and SOPHIE. From the experiments, it can be concluded that other species different to Te 2 , such as tin telluride and cesium telluride, may be released from the fuel. That is why they must be considered in the transport phenomena. There is also experimental evidence of the strong interaction of Te 2 with Inconel 600 and stainless steel of the pipe walls and structures, however this strong interaction is in competition with the interaction of tellurium with aerosols, which under severe accident conditions may represent an area greater than that of the primary system. It is for the absence of significant tellurium species in the transport models, and also for the interaction of tellurium with aerosols, for which some codes show the greatest deficiencies

  11. Rhizopus-associated soft tissue infection in an immunocompetent air-conditioning technician after a road traffic accident: A case report and review of the literature

    Directory of Open Access Journals (Sweden)

    Nada B. Rabie

    2012-02-01

    Full Text Available Summary: Rhinocerebral or sinopulmonary mucromycosis is a well-recognized human fungal infection found among immunocompromised and diabetic patients. However, the infection is rare among immunocompetent hosts. We are reporting the case of an adult immunocompetent male patient working as an air-conditioning technician. The patient was a victim of a road traffic accident (RTA and sustained multiple fractures in the proximal part of the left tibia, distal femur, and scapula. Two weeks postoperatively, Rhizopus microspores were isolated from an infected traumatic wound over the distal femur. Surgical debridement was performed, and the patient was started on amphotericin B. Occupational exposure history and workplace environmental sanitation are crucial for the prevention of this potentially fatal yet preventable infection. Keywords: Rhizopus, Immunocompetent, Air conditioning

  12. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1987-01-01

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

  13. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Dong Wook Jerng; Choong Sup Byun

    1998-01-01

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

  15. 49 CFR 195.50 - Reporting accidents.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Reporting accidents. 195.50 Section 195.50 Transportation Other Regulations Relating to Transportation (Continued) PIPELINE AND HAZARDOUS MATERIALS SAFETY... PIPELINE Annual, Accident, and Safety-Related Condition Reporting § 195.50 Reporting accidents. An accident...

  16. Probability analysis of WWER-1000 fuel elements behavior under steady-state, transient and accident conditions of reactor operation

    International Nuclear Information System (INIS)

    Tutnov, A.; Alexeev, E.

    2001-01-01

    'PULSAR-2' and 'PULSAR+' codes make it possible to simulate thermo-mechanical and thermo-physical parameters of WWER fuel elements. The probabilistic approach is used instead of traditional deterministic one to carry out a sensitive study of fuel element behavior under steady-state operation mode. Fuel elements initial parameters are given as a density of the probability distributions. Calculations are provided for all possible combinations of initial data as fuel-cladding gap, fuel density and gas pressure. Dividing values of these parameters to intervals final variants for calculations are obtained . Intervals of permissible fuel-cladding gap size have been divided to 10 equal parts, fuel density and gas pressure - to 5 parts. Probability of each variant realization is determined by multiplying the probabilities of separate parameters, because the tolerances of these parameters are distributed independently. Simulation results are turn out in the probabilistic bar charts. The charts present probability distribution of the changes in fuel outer diameter, hoop stress kinetics and fuel temperature versus irradiation time. A normative safety factor is introduced for control of any criterion realization and for determination of a reserve to the criteria failure. A probabilistic analysis of fuel element behavior under Reactivity Initiating Accident (RIA) is also performed and probability fuel element depressurization under hypothetical RIA is presented

  17. Nuclear accidents

    International Nuclear Information System (INIS)

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Cartmill, C.E.

    1978-01-01

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

  19. Mechanical energy release and fuel fragmentation in high energy deposition into fuel under a reactivity initiated accident condition

    International Nuclear Information System (INIS)

    Tsuruta, Takaharu; Saito, Shinzo; Ochiai, Masaaki

    1985-01-01

    The fuel fragmentation is one of important subjects to be studied, since it is one of basic processes of molten fuel-coolant interaction (MFCI) and it has not yet been made clear enough. Accordingly, UO 2 fuel fragmentation was studied in the NSRR experiments simulating a reactivity initiated accident (RIA). As results of the experiments, the distribution of the size of fuel fragments was obtained and the mechanism of fuel fragmentation was discussed as described below. It was revealed that the distribution was well displayed in the form of logarithmic Rosin-Rammler's distribution law. It was shown that the conversion ratio from thermal energy to mechanical in the experiment was in inverse propotion to the volume-surface mean diameter defined as a ratio of the total volume of fragments to the total surface. Consequently, it was confirmed that the mean diameter was proper as an index for the degree of the fuel fragmentation. It was also pointed out that the Weber-type hydraulic instability model for fragmentation was consistent with the experimental results. The mechanism of the fuel fragmentation is understood as follows. Cladding tube is ruptured due to the increase in rod pressure when fuel is molten, and then molten fuel spouts through the openings in the form of jet. As a result of molten fuel spouting, fuel is fragmented by the Weber-type of hydraulic instability. The model well explains the effects of experimental parameters as heat deposition, subcooling of cooling water and capsule diameter, on the fuel fragmentation. According to the model, fuel fragments have to be spherical. There were many spherical particles which had hollow and burst crack. This may be due to internal burst during solidification process. The items which should be studied further are also described in the end of this report. (author)

  20. Effect of culturing conditions on the expression of key enzymes in the proteolytic system of Lactobacillus bulgaricus *

    Science.gov (United States)

    Hou, Jun-cai; Liu, Fei; Ren, Da-xi; Han, Wei-wei; Du, Yue-ou

    2015-01-01

    The proteolytic system of Lactobacillus bulgaricus breaks down milk proteins into peptides and amino acids, which are essential for the growth of the bacteria. The aim of this study was to determine the expressions of seven key genes in the proteolytic system under different culturing conditions (different phases, initial pH values, temperatures, and nitrogen sources) using real-time polymerase chain reaction (RT-PCR). The transcriptions of the seven genes were reduced by 30-fold on average in the stationary phase compared with the exponential growth phase. The transcriptions of the seven genes were reduced by 62.5-, 15.0-, and 59.0-fold in the strains KLDS 08006, KLDS 08007, and KLDS 08012, respectively, indicating that the expressions of the seven genes were significantly different among strains. In addition, the expressions of the seven genes were repressed in the MRS medium containing casein peptone. The effect of peptone supply on PepX transcription was the weakest compared with the other six genes, and the impact on OppD transcription was the strongest. Moreover, the expressions of the seven genes were significantly different among different strains (PLactobacillus bulgaricus at the transcription level. PMID:25845365

  1. ngVLA Key Science Goal 2: Probing the Initial Conditions for Planetary Systems and Life with Astrochemistry

    Science.gov (United States)

    McGuire, Brett; ngVLA Science Working Group 1

    2018-01-01

    One of the most challenging aspects in understanding the origin and evolution of planets and planetary systems is tracing the influence of chemistry on the physical evolution of a system from a molecular cloud to a solar system. Existing facilities have already shown the stunning degree of molecular complexity present in these systems. The unique combination of sensitivity and spatial resolution offered by the ngVLA will permit the observation of both highly complex and very low-abundance chemical species that are exquisitely sensitive to the physical conditions and evolutionary history of their sources, which are out of reach of current observatories. In turn, by understanding the chemical evolution of these complex molecules, unprecedentedly detailed astrophysical insight can be gleaned from these astrochemical observations.This poster will overview a number of key science goals in astrochemistry which will be enabled by the ngVLA, including:1) imaging of the deepest, densest regions in protoplanetary disks and unveiling the physical history through isotopic ratios2) probing the ammonia snow line in these disks, thought to be the only viable tracer of the water snowline3) observations of the molecular content of giant planet atmospheres4) detections of new, complex molecules, potentially including the simplest amino acids and sugars5) tracing the origin of chiral excess in star-forming regions

  2. Influence of postharvest processing and storage conditions on key antioxidants in pūhā (Sonchus oleraceus L.).

    Science.gov (United States)

    Ou, Zong-Quan; Schmierer, David M; Strachan, Clare J; Rades, Thomas; McDowell, Arlene

    2014-07-01

    To investigate effects of different postharvest drying processes and storage conditions on key antioxidants in Sonchus oleraceus L. leaves. Fresh leaves were oven-dried (60°C), freeze-dried or air-dried (∼25°C) for 6 h, 24 h and 3 days, respectively. Design of experiments (DOE) was applied to study the stability of antioxidants (caftaric, chlorogenic and chicoric acids) in S. oleraceus leaves and leaf extracts stored at different temperatures (4, 25 and 50°C) and relative humidities (15%, 43% and 75%) for 180 days. The concentration of antioxidants was quantified by a HPLC-2,2'-diphenylpicrylhydrazyl post-column derivatisation method. Antioxidant activity was assessed by a cellular antioxidant activity assay. The three antioxidants degraded to unquantifiable levels after oven-drying. More than 90% of the antioxidants were retained by freeze-drying and air-drying. Both leaf and extract samples retained >90% of antioxidants, except those stored at 75% relative humidity. Leaf material had higher antioxidant concentrations and greater cellular antioxidant activity than corresponding extract samples. Freeze-drying and air-drying preserved more antioxidants in S. oleraceus than oven-drying. From DOE analysis, humidity plays an important role in degradation of antioxidants during storage. To preserve antioxidant activity, it is preferable to store S. oleraceus as dried leaf material. © 2014 Royal Pharmaceutical Society.

  3. A Narrow-Linewidth Atomic Line Filter for Free Space Quantum Key Distribution under Daytime Atmospheric Conditions

    Science.gov (United States)

    Brown, Justin; Woolf, David; Hensley, Joel

    2016-05-01

    Quantum key distribution can provide secure optical data links using the established BB84 protocol, though solar backgrounds severely limit the performance through free space. Several approaches to reduce the solar background include time-gating the photon signal, limiting the field of view through geometrical design of the optical system, and spectral rejection using interference filters. Despite optimization of these parameters, the solar background continues to dominate under daytime atmospheric conditions. We demonstrate an improved spectral filter by replacing the interference filter (Δν ~ 50 GHz) with an atomic line filter (Δν ~ 1 GHz) based on optical rotation of linearly polarized light through a warm Rb vapor. By controlling the magnetic field and the optical depth of the vapor, a spectrally narrow region can be transmitted between crossed polarizers. We find that the transmission is more complex than a single peak and evaluate peak transmission as well as a ratio of peak transmission to average transmission of the local spectrum. We compare filters containing a natural abundance of Rb with those containing isotopically pure 87 Rb and 85 Rb. A filter providing > 95 % transmission and Δν ~ 1.1 GHz is achieved.

  4. Effect of culturing conditions on the expression of key enzymes in the proteolytic system of Lactobacillus bulgaricus.

    Science.gov (United States)

    Hou, Jun-cai; Liu, Fei; Ren, Da-xi; Han, Wei-wei; Du, Yue-ou

    2015-04-01

    The proteolytic system of Lactobacillus bulgaricus breaks down milk proteins into peptides and amino acids, which are essential for the growth of the bacteria. The aim of this study was to determine the expressions of seven key genes in the proteolytic system under different culturing conditions (different phases, initial pH values, temperatures, and nitrogen sources) using real-time polymerase chain reaction (RT-PCR). The transcriptions of the seven genes were reduced by 30-fold on average in the stationary phase compared with the exponential growth phase. The transcriptions of the seven genes were reduced by 62.5-, 15.0-, and 59.0-fold in the strains KLDS 08006, KLDS 08007, and KLDS 08012, respectively, indicating that the expressions of the seven genes were significantly different among strains. In addition, the expressions of the seven genes were repressed in the MRS medium containing casein peptone. The effect of peptone supply on PepX transcription was the weakest compared with the other six genes, and the impact on OppD transcription was the strongest. Moreover, the expressions of the seven genes were significantly different among different strains (Pproteolytic system genes in Lactobacillus bulgaricus at the transcription level.

  5. Estimation of risk due to accidents for the transport of radioactive wastes to the conditioning and storage facilities in the Research Center of Seibersdorf

    International Nuclear Information System (INIS)

    Krejsa, P.

    1977-02-01

    By the use of an American statistic of accidents on roads the risk of body burden is estimated resulting from the transport of radioactive wastes to the central collection, conditioning and storage facilities in Seibersdorf. It is shown that the risk of the transport from power stations up to 1990 is below that of other producers of radioactive wastes (hospitals, industry and research laboratories). The risk of the individual body burden is estimated to be in 1976: 1,1 . 10 -10 mrem/a; 1978: 2,8 . 10 -10 mrem/a; 1985: 3,0 . 10 -10 mrem/a; 1995: 3,3 . 10 -10 mrem/a. These results are so much below the natural radiation in the environment, that they cannot be seen as an increase in the given potential hazard. (author)

  6. Development of a kinetic model, including rate constant estimations, on iodine and caesium behaviour in the primary circuit of LWR's under accident conditions

    International Nuclear Information System (INIS)

    Alonso, A.; Buron, J.M.; Fernandez, S.

    1991-07-01

    In this report, a kinetic model has been developed with the aim to try to reproduce the chemical phenomena that take place in a flowing system containing steam, hydrogen and iodine and caesium vapours. The work is divided into two different parts. The first part consists in the estimation, through the Activited Complex Theory, of the reaction rate constants, for the chosen reactions, and the development of the kinetic model based on the concept of ideal tubular chemical reactor. The second part deals with the application of such model to several cases, which were taken from the Phase B 'Scoping Calculations' of the Phebus-FP Project (sequence AB) and the SFD-ST and SFD1.1 experiments. The main conclusion obtained from this work is that the assumption of instantaneous equilibrium could be inacurrate in order to estimate the iodine and caesium species distribution under severe accidents conditions

  7. Rhizopus-associated soft tissue infection in an immunocompetent air-conditioning technician after a road traffic accident: a case report and review of the literature.

    Science.gov (United States)

    Rabie, Nada B; Althaqafi, Abdulhakeem O

    2012-03-01

    Rhinocerebral or sinopulmonary mucromycosis is a well-recognized human fungal infection found among immunocompromised and diabetic patients. However, the infection is rare among immunocompetent hosts. We are reporting the case of an adult immunocompetent male patient working as an air-conditioning technician. The patient was a victim of a road traffic accident (RTA) and sustained multiple fractures in the proximal part of the left tibia, distal femur, and scapula. Two weeks postoperatively, Rhizopus microspores were isolated from an infected traumatic wound over the distal femur. Surgical debridement was performed, and the patient was started on amphotericin B. Occupational exposure history and workplace environmental sanitation are crucial for the prevention of this potentially fatal yet preventable infection. Copyright © 2011 King Saud Bin Abdulaziz University for Health Sciences. Published by Elsevier Ltd. All rights reserved.

  8. Procedure for qualification of electric equipment installed in containments for pressurized water reactors subject to accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    This generic norm is usable for electrical equipment installed in containment building of PWR subject to accidental conditions. She defines the qualification methods and the general rules usable for the test specifications of qualification for these materials

  9. Factors affecting vertical distribution of Fukushima accident-derived radiocesium in soil under different land-use conditions

    International Nuclear Information System (INIS)

    Koarashi, Jun; Atarashi-Andoh, Mariko; Matsunaga, Takeshi; Sato, Tsutomu; Nagao, Seiya; Nagai, Haruyasu

    2012-01-01

    The Fukushima Dai-ichi nuclear power plant accident in Japan, triggered by a big earthquake and the resulting tsunami on 11 March 2011, caused a substantial release of radiocesium ( 137 Cs and 134 Cs) and a subsequent contamination of soils in a range of terrestrial ecosystems. Identifying factors and processes affecting radiocesium retention in these soils is essential to predict how the deposited radiocesium will migrate through the soil profile and to other biological components. We investigated vertical distributions of radiocesium and physicochemical properties in soils (to 20 cm depth) at 15 locations under different land-use types (croplands, grasslands, and forests) within a 2 km × 2 km mesh area in Fukushima city. The total 137 Cs inventory deposited onto and into soil was similar (58.4 ± 9.6 kBq m −2 ) between the three different land-use types. However, aboveground litter layer at the forest sites and herbaceous vegetation at the non-forested sites contributed differently to the total 137 Cs inventory. At the forest sites, 50–91% of the total inventory was observed in the litter layer. The aboveground vegetation contribution was in contrast smaller ( 137 Cs in mineral soil layers; 137 Cs penetrated deeper in the forest soil profiles than in the non-forested soil profiles. We quantified 137 Cs retention at surface soil layers, and showed that higher 137 Cs retention can be explained in part by larger amounts of silt- and clay-sized particles in the layers. More importantly, the 137 Cs retention highly and negatively correlated with soil organic carbon content divided by clay content across all land-use types. The results suggest that organic matter inhibits strong adsorption of 137 Cs on clay minerals in surface soil layers, and as a result affects the vertical distribution and thus the mobility of 137 Cs in soil, particularly in the forest ecosystems. - Highlights: ► Vertical distribution of radiocesium was investigated for 15 soils. ► Forest

  10. Interrogations to Learn from the Fukushima Accident

    International Nuclear Information System (INIS)

    Gisquet, E.; Jeffroy, F.

    2016-01-01

    support was more difficult to provide than expected due to the conditions of the emergency. Different key issues would be proposed to support ERC for coordination and innovation in extreme situations. 3. What is the dynamic decision of the crisis? Beyond the firm’s organization will be examined the relationship of the utility with a still larger organization involved in the response to the accident: the Japanese government. How the different stakeholders are able to cooperate in addressing the challenges entailed by the accident, adjusting their actions and making decisions accordingly? (author)

  11. Culture conditions for equine bone marrow mesenchymal stem cells and expression of key transcription factors during their differentiation into osteoblasts

    Science.gov (United States)

    2013-01-01

    Background The use of equine bone marrow mesenchymal stem cells (BMSC) is a novel method to improve fracture healing in horses. However, additional research is needed to identify optimal culture conditions and to determine the mechanisms involved in regulating BMSC differentiation into osteoblasts. The objectives of the experiments were to determine: 1) if autologous or commercial serum is better for proliferation and differentiation of equine BMSC into osteoblasts, and 2) the expression of key transcription factors during the differentiation of equine BMSC into osteoblasts. Equine BMSC were isolated from the sterna of 3 horses, treated with purchased fetal bovine serum (FBS) or autologous horse serum (HS), and cell proliferation determined. To induce osteoblast differentiation, cells were incubated with L-ascorbic acid-2-phosphate and glycerol-2-phosphate in the presence or absence of human bone morphogenetic protein2 (BMP2), dexamethasone (DEX), or combination of the two. Alkaline phosphatase (ALP) activity, a marker of osteoblast differentiation, was determined by ELISA. Total RNA was isolated from differentiating BMSC between d 0 to 18 to determine expression of runt-related transcription factor2 (Runx2), osterix (Osx), and T-box3 (Tbx3). Data were analyzed by ANOVA. Results Relative to control, FBS and HS increased cell number (133 ± 5 and 116 ± 5%, respectively; P  0.8). Runt-related transcription factor2 expression increased 3-fold (P equine BMSC into osteoblasts. In addition, expression of Runx2 and osterix increased and expression of Tbx3 is reduced during differentiation. PMID:24169030

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

    International Nuclear Information System (INIS)

    Höhne, Thomas; Grahn, Alexander; Kliem, Sören; Rohde, Ulrich; Weiss, Frank-Peter

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Garcia Rodenas, L.; Ali, S.P.; Liberman, S.J.

    1987-01-01

    H 2 , HD and D 2 radiolytic yield in heavy and light water mixtures has been determined to supply the necessary data which will allow to make a realistic estimation of the solution of such gas under LOCA conditions as a function of time. (Author)

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

    International Nuclear Information System (INIS)

    Smidts, C.; Shen, S.H.; Mosleh, A.

    1997-01-01

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

  15. ROAD ACCIDENT AND SAFETY STUDY IN SYLHET REGION OF BANGLADESH

    Directory of Open Access Journals (Sweden)

    B. K. BANIK

    2011-08-01

    Full Text Available Roads, highways and streets are fundamental infrastructure facilities to provide the transportation for passenger travel and goods movement from one place to another in Sylhet, north–eastern division of Bangladesh with rapid growth of road vehicle, being comparatively developed economic tourist prone area faces severe road traffic accident. Such severe road accidents cause harsh safety hazards on the roads of Sylhet area. This research work presents an overview of the road traffic accident and degraded road safety situation in Sylhet zone which in particular, discusses the key road accident problem characteristics identifying the hazardous roads and spots, most responsible vehicles and related components, conditions of drivers and pedestrians, most victims of accident, effects of accident on society, safety priorities and options available in Sylhet. In this regard, a comprehensive questionnaire survey was conducted on the concerned groups of transportation and detailed accident data was collected from a popular local newspaper. Analysis of the study reveals that Dhaka- Sylhet highway is the most hazardous in road basis and Sylhet Sador thana is the most vulnerable in thana basis in Sylhet region.

  16. Understanding of the operation behaviour of a Passive Autocatalytic Recombiner (PAR) for hydrogen mitigation in realistic containment conditions during a severe Light Water nuclear Reactor (LWR) accident

    International Nuclear Information System (INIS)

    Payot, Frédéric; Reinecke, Ernst-Arndt; Morfin, Franck; Sabroux, Jean-Christophe; Meynet, Nicolas; Bentaib, Ahmed; March, Philippe; Zeyen, Roland

    2012-01-01

    Highlights: ► Recombineur operation in the presence of fission products (severe accident conditions). ► Operation of catalysts in the integral and small-scale tests. ► The catalyst performance was observed by measuring the coupon temperature increase. ► The experimental observations were corroborated by numerical calculations (SPARK code). - Abstract: In the context of hydrogen risk mitigation in nuclear power plants (NPPs), experimental studies of a possible poisoning of Passive Autocatalytic Recombiners (PARs) by fission products (FPs) and aerosols released during a core meltdown accident were mainly conducted in the past with non-radioactive fission product surrogates (e.g., in the H2PAR facility at Cadarache, France). The decision was taken in 1997 to complete these studies by a test in the Phébus facility, a research nuclear reactor also at Cadarache: it was a rare opportunity to expose catalyst samples to an atmosphere as representative as possible of a core meltdown accident, containing gaseous fission products and aerosols released during the degradation of an actual irradiated nuclear fuel bundle. Before testing in Phébus during the FPT3 experiment, reference and qualification tests were performed in the H2PAR facility using the same samples — the so-called “coupons” — and coupons holder to check that the apparatus was functional and correctly designed for avoiding to tamper with the thermal-hydraulics and chemical conditions in the Phébus containment. The correct operation of catalysts was checked by measuring the surface temperature increase of the coupons due to the exothermic reaction between hydrogen and oxygen. After the Phébus FPT3 test (November 2004), REKO-1 tests were initiated at Jülich, Germany, to confirm the discrepancy in coupons temperature observed in Phébus FPT3 and H2PAR PHEB-03 tests, and to study the operation behaviour of PARs. Besides, before REKO-1 tests, a first interpretation of H2PAR and Phébus experiments

  17. Current status of the FASTGRASS/PARAGRASS models for fission product release from LWR fuel during normal and accident conditions

    International Nuclear Information System (INIS)

    Rest, J.; Zawadski, S.A.; Piasecka, M.

    1983-10-01

    The theoretical FASTGRASS model for the prediction of the behavior of the gaseous and volatile fission products in nuclear fuels under normal and transient conditions has undergone substantial improvements. The major improvements have been in the atomistic and bubble diffusive flow models, in the models for the behavior of gas bubbles on grain surfaces, and in the models for the behavior of the volatile fission products iodine and cesium. The thoery has received extensive verification over a wide range of fuel operating conditions, and can be regarded as a state-of-the-art model based on our current level of understanding of fission product behavior. PARAGRASS is an extremely efficient, mechanistic computer code with the capability of modeling steady-state and transient fission-product behavior. The models in PARAGRASS are based on the more detailed ones in FASTGRASS. PARAGRASS updates for the FRAPCON (PNL), FRAP-T (INEL), and SCDAP (INEL) codes have recently been completed and implemented. Results from an extensive FASTGRASS verification are presented and discussed for steady-state and transient conditions. In addition, FASTGRASS predictions for fission product release rate constants are compared with those in NUREG-0772. 21 references, 13 figures

  18. Normal accidents

    International Nuclear Information System (INIS)

    Perrow, C.

    1989-01-01

    The author has chosen numerous concrete examples to illustrate the hazardousness inherent in high-risk technologies. Starting with the TMI reactor accident in 1979, he shows that it is not only the nuclear energy sector that bears the risk of 'normal accidents', but also quite a number of other technologies and industrial sectors, or research fields. The author refers to the petrochemical industry, shipping, air traffic, large dams, mining activities, and genetic engineering, showing that due to the complexity of the systems and their manifold, rapidly interacting processes, accidents happen that cannot be thoroughly calculated, and hence are unavoidable. (orig./HP) [de

  19. Fission products and nuclear fuel behaviour under severe accident conditions part 1: Main lessons learnt from the first VERDON test

    Science.gov (United States)

    Pontillon, Y.; Geiger, E.; Le Gall, C.; Bernard, S.; Gallais-During, A.; Malgouyres, P. P.; Hanus, E.; Ducros, G.

    2017-11-01

    This paper describes the first VERDON test performed at the end of September 2011 with special emphasis on the behaviour of fission products (FP) and actinides during the accidental sequence itself. Two other papers discuss in detail the post-test examination results (SEM, EPMA and SIMS) of the VERDON-1 sample. The first VERDON test was devoted to studying UO2 fuel behaviour and fission product releases under reducing conditions at very high temperature (∼2883 K), which was able to confirm the very good performance of the VERDON loop. The fuel sample did not lose its integrity during this test. According to the FP behaviour measured by the online gamma station (fuel sight), the general classification of the FP in relation to their released fraction is very accurate, and the burn-up effect on the release rate is clearly highlighted.

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

    International Nuclear Information System (INIS)

    Kryger, B.; Ducamp, F.; Combette, P.

    1981-08-01

    We analyse the mixed oxide fast fuel response to off normal conditions obtained by means of an out-of-pile transient simulation apparatus designed to provide direct observations (temperature, pressure, fuel motion) of fuel fission gas phenomena that might occur during the transients. The results are concerning fast transient tests (0,1 to 1 second) obtained with high gas concentration irradiated fuel (4 to 7 at % burn up, 0,4 cm 3 Xe + Kr /g.UPuO 2 ). The kinetics of fission gas release during the transients have been directly measured and then compared with the calculated results issued of the Chicago model. This model agrees, quite well, with other experiments done in the silene prompt reactor. Other gases than xenon and krypton (such as hydrogen and carbon monoxide) do not play any role in fuel behavior, since they have been completely ruled out

  1. Evaluation of transient natural circulation behavior during accident in low power/shutdown condition of YGN units 3/4

    International Nuclear Information System (INIS)

    Bang, Young Seok; Kim, Kap; Seul, Kwang Won; Kim, Hho Jung

    1997-01-01

    A transient natural circulation behavior during a LOCA at hot-standby operation is evaluated for YGN Units 3/4. The plant initial condition is determined within the EOP limitation as suitable to hot-standby mode and the transient scenario is prepared as relevant to evaluation of transient natural circulation. A 0.4% cold leg break with loss of off-site power is calculated with RELAP5/MOD3.2, whose predictability has been verified for SBLOCA natural circulation test, S-NC-8B. Through one hour transient analysis, it is found that the plant has its own decay heat removal capability by natural circulation following a LOCA at hot-standby mode. Additional calculation is performed to investigate an effect of HPSI flow on natural circulation

  2. Accident Statistics

    Data.gov (United States)

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

  3. Radiation accidents

    International Nuclear Information System (INIS)

    Nenot, J.C.

    1996-01-01

    Analysis of radiation accidents over a 50 year period shows that simple cases, where the initiating events were immediately recognised, the source identified and under control, the medical input confined to current handling, were exceptional. In many cases, the accidents were only diagnosed when some injuries presented by the victims suggested the radiological nature of the cause. After large-scale accidents, the situation becomes more complicated, either because of management or medical problems, or both. The review of selected accidents which resulted in severe consequences shows that most of them could have been avoided; lack of regulations, contempt for rules, human failure and insufficient training have been identified as frequent initiating parameters. In addition, the situation was worsened because of unpreparedness, insufficient planning, unadapted resources, and underestimation of psychosociological aspects. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  5. Sports Accidents

    CERN Multimedia

    Kiebel

    1972-01-01

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

  6. Radiation accidents

    International Nuclear Information System (INIS)

    Poplavskij, K.K.; Smorodintseva, G.I.

    1978-01-01

    On the basis of a critical analysis of the available data on causes and consequences of radiation accidents (RA), a classification of RA by severity (five groups of accidents) according to biomedical consequences and categories of exposed personnel is proposed. A RA is defined and its main characteristics are described. Methods of RA prevention are proposed, as is a plan of specific measures to deal with RA in accordance with the proposed classification

  7. Accident management information needs

    International Nuclear Information System (INIS)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R.

    1990-04-01

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

  8. Accident management information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-04-01

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

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

    International Nuclear Information System (INIS)

    Velkov, K.

    2002-01-01

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

  10. Core loss during a severe accident (COLOSS)

    International Nuclear Information System (INIS)

    Adroguer, B.; Bertrand, F.; Chatelard, P.; Cocuaud, N.; Van Dorsselaere, J.P.; Bellenfant, L.; Knocke, D.; Bottomley, D.; Vrtilkova, V.; Belovsky, L.; Mueller, K.; Hering, W.; Homann, C.; Krauss, W.; Miassoedov, A.; Schanz, G.; Steinbrueck, M.; Stuckert, J.; Hozer, Z.; Bandini, G.; Birchley, J.; Berlepsch, T. von; Kleinhietpass, I.; Buck, M.; Benitez, J.A.F.; Virtanen, E.; Marguet, S.; Azarian, G.; Caillaux, A.; Plank, H.; Boldyrev, A.; Veshchunov, M.; Kobzar, V.; Zvonarev, Y.; Goryachev, A.

    2005-01-01

    The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H 2 production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO 2 and MOX by molten Zircaloy (b) simultaneous dissolution of UO 2 and ZrO 2 (c) oxidation of U-O-Zr mixtures (d) degradation-oxidation of B 4 C control rods. Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B 4 C control rods and in the TMI-2 accident. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Breakthroughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO 2 and MOX dissolution and oxidation of U-O-Zr and B 4 C-metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H 2 production observed during the reflooding of degraded cores under severe accident conditions. The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results. Main results and recommendations for future R and D activities are summarized in this paper

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

    International Nuclear Information System (INIS)

    Tentner, A.M.; Parma, E.; Wei, T.; Wigeland, R.

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-01

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

  13. Part A: Countermeasures to be taken after 1990 to ensure safe living conditions for the population affected by the Chernobyl accident in the USSR. A first evaluation of costs and doses averted

    International Nuclear Information System (INIS)

    Lochard, J.; Schneider, T.

    1992-01-01

    This part presents a first estimate of the cost and averted collective exposure of the potential relocation of the population from the affected territories of the BSSR, the RSFSR and the UKrSSR, to improve their living conditions following the Chernobyl accident. It is an input to the evaluation of the radiological consequences of the Chernobyl accident in the USSR. The general objective was to assess 'the concept which the USSR has evolved to enable the population to live safely in areas affected by radioactive contamination following the Chernobyl accident, and an evaluation of the effectiveness of the steps taken in these areas to safeguard the health of the population'. Specifically, this work aimed at evaluating protective measures from 1990 onwards

  14. Models and criteria for prediction of Deflagration-to-Detonation Transition (DDT) in hydrogen-air-steam systems under severe accident conditions. Final report

    International Nuclear Information System (INIS)

    Klein, R.; Rehm, W.

    1999-01-01

    The European Commission in Brussels supported a joint project on Deflagration-to-Detonation Transition (DDT) studies for hydrogen safety within the framework programme on nuclear fission safety. The project was initiated by the Forschungszentrum Juelich based on the results of a pilot project. The following main project was coordinated by the Freie Universitaet Berlin involving seven european partners. The partners came from universities, research centers and industry, as follows: FU-Berlin, RWTH-Aachen, CNRS-Marseille, IPSN-Saclay, FZ-Juelich, FZ-Karlsruhe, and NNC-Knutsford, which worked closely together. The working period was two years (1997-1998). The aim of the project was to develop models and criteria for prediction of deflagration-to-detonation transition (DDT) in hydrogen-air-steam systems under severe accident conditions. The results obtained are documented in this final report, which was finished in 1999. The report consists of seven chapters, concerning: - Introduction - Experimental Investigations - Modelling and Numerics - Validation - Mitigation - Further Deliverables - Summary and Conclusion. The final report presents special experimental, theoretical, and computational aspects of the complex DDT phenomena for hydrogen safety studies, and it should be a solid basis for end user applications and further developments. (orig.)

  15. Effects of indoor and outdoor cultivation conditions on 137 Cs concentrations in cultivated mushrooms produced after the Fukushima Daiichi Nuclear Power Plant accident.

    Science.gov (United States)

    Tagami, Keiko; Uchida, Shigeo; Ishii, Nobuyoshi

    2017-01-01

    Radiocesium ( 134 Cs and 137 Cs) in mushrooms has been a matter of public concern after the accident at Fukushima Daiichi Nuclear Power Plant. To minimize the internal dose by ingestion of cultivated mushrooms, the Japanese government set a guideline level with respect to the radiocesium concentration in bed-logs and mushroom beds; however, the effects of indoor and outdoor cultivation methods on radiocesium concentrations in cultivated mushrooms were not clear. The effects of indoor and outdoor cultivation on the radiocesium concentrations in mushroom were examined using published food monitoring data. 137 Cs concentration data in Lentinula edodes from the Aizu area in Fukushima Prefecture and seven prefectures outside Fukushima were used for the analysis. No statistically significant 137 Cs concentration differences were found between these two cultivation methods. Using detected 137 Cs data in shiitake, the geometric means from each prefecture were less than one-quarter of the standard limit (100 Bq kg -1 ) for total radiocesium under both cultivation conditions. It was suspected that re-suspended radiocesium might have been taken up by mushrooms or that radiocesium might have been absorbed into the mushrooms from the soil in the outdoor cultures. However, neither effect was significant for cultivated mushrooms in the areas examined. © 2016 Society of Chemical Industry. © 2016 Society of Chemical Industry.

  16. CAMDYN: a new model to describe the axial motion of molten fuel inside the pin of a fast breeder reactor during accident conditions

    International Nuclear Information System (INIS)

    Peter, G.

    1991-01-01

    The new in-pin fuel motion model CAMDYN (Cavity Material Dynamics) describes the axial motion of both partially and fully molten fuel inside the pin of a fast breeder reactor during accident conditions. The motion of the two types of molten fuel and the imbedded fission gas bubbles is treated both before and after cladding failure. The basic modelling approach consists of the treatment of two one-dimensional flows which are coupled by interaction terms. Each of these flows is treated compressively and with axially variable flow cross sections. The mass and energy equations of both fields are solved explicitly using upwind differencing on a fixed Eulerian grid. The two momentum equations are solved simultaneously, using the convective momentum fluxes of the previous timestep. Both partially and fully molten fuel can move axially into a central hole extending to the plenum in the case of certain hollow pellet designs. The fuel temperature calculation includes the determination of a radial temperature profile. A simple conduction freezing model is included. After cladding failure, ejection into the coolant channel is modeled

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

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

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung

    1997-01-01

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

  20. Lizimetric investigation of vertical transportation of Cs-137 in the different soil types for the different raining conditions: a new solution proposal and application for the Chernobyl accident

    International Nuclear Information System (INIS)

    Varinlioglu, A.

    2000-12-01

    The aim of this Ph.D. thesis is to trace the vertical transportation of water in soil by using the Cs-137 as radiotracer. We have studied 3 different soil types clay, loamy and sandy and 3 different raining conditions. High, normal and low raining regimes that are typical especially for the Black Sea Region were selected for experimental parameters. These experiments have been realized in lizimetric conditions. As a result, the type of applications for the combinations of soil type and raining regime contain experimental originality. Humidity observation technique was also used. The result of the experiments show that in every raining regime the activities of different types of soils can be classified as sand, loam and clay in descending activity order. When the results of the experiments were evaluated according to the raining regime it can be seen that the relative activity for every type of soil is always towards higher to lower raining conditions. These results are what we already have expected and are related to the water permeability of the soil. The surface soil activities have also been measured. The surface soil activities were arranged as clayey, loamy and sandy in descending activity order for every laining regime. The surface soil activities were found to be higher in the soil types in which water could not penetrate deeper. When the relative activity values were compared with the results of the surface soil activities it was recognized that the order of soils changes upside down. This is an evident of harmony for the experimental results. For the evaluation of surface activity it can be seen that the relative activity values for every kind of soil type is always in the same order from the highest to the lowest raining regimes. As a result, it can be said that the all results have significant and have harmonies with each other. Experimental results were comparatively examined by using the s emi infinite convection dispersion model . All the conclusions

  1. Criticality accident:

    International Nuclear Information System (INIS)

    Canavese, Susana I.

    2000-01-01

    A criticality accident occurred at 10:35 on September 30, 1999. It occurred in a precipitation tank in a Conversion Test Building at the JCO Tokai Works site in Tokaimura (Tokai Village) in the Ibaraki Prefecture of Japan. STA provisionally rated this accident a 4 on the seven-level, logarithmic International Nuclear Event Scale (INES). The September 30, 1999 criticality accident at the JCO Tokai Works Site in Tokaimura, Japan in described in preliminary, technical detail. Information is based on preliminary presentations to technical groups by Japanese scientists and spokespersons, translations by technical and non-technical persons of technical web postings by various nuclear authorities, and English-language non-technical reports from various news media and nuclear-interest groups. (author)

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

    International Nuclear Information System (INIS)

    Girard, P.; Dubreuil, G.H.

    1996-01-01

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

  3. Identification of key factors in Accelerated Low Water Corrosion through experimental simulation of tidal conditions: influence of stimulated indigenous microbiota

    NARCIS (Netherlands)

    Marty, F.; Gueuné, H.; Malard, E.; Sánchez-Amaya, J.M.; Sjögren, L.; Abbas, B.; Quillet, L.; van Loosdrecht, M.C.M.; Muyzer, G.

    2014-01-01

    Biotic and abiotic factors favoring Accelerated Low Water Corrosion (ALWC) on harbor steel structures remain unclear warranting their study under controlled experimental tidal conditions. Initial stimulation of marine microbial consortia by a pulse of organic matter resulted in localized corrosion

  4. Criticality accident of nuclear fuel facility. Think back on JCO criticality accident

    International Nuclear Information System (INIS)

    Naito, Keiji

    2003-09-01

    This book is written in order to understand the fundamental knowledge of criticality safety or criticality accident of nuclear fuel facility by the citizens. It consists of four chapters such as critical conditions and criticality accident of nuclear facility, risk of criticality accident, prevention of criticality accident and a measure at an occurrence of criticality accident. A definition of criticality, control of critical conditions, an aspect of accident, a rate of incident, damage, three sufferers, safety control method of criticality, engineering and administrative control, safety design of criticality, investigation of failure of safety control of JCO criticality accident, safety culture are explained. JCO criticality accident was caused with intention of disregarding regulation. It is important that we recognize the correct risk of criticality accident of nuclear fuel facility and prevent disasters. On the basis of them, we should establish safety culture. (S.Y.)

  5. Occupational accidents among mototaxi drivers.

    Science.gov (United States)

    Amorim, Camila Rego; de Araújo, Edna Maria; de Araújo, Tânia Maria; de Oliveira, Nelson Fernandes

    2012-03-01

    The use of motorcycles as a means of work has contributed to the increase in traffic accidents, in particular, mototaxi accidents. The aim of this study was to estimate and characterize the incidence of occupational accidents among the mototaxis registered in Feira de Santana, BA. This is a cross-sectional study with descriptive and census data. Of the 300 professionals registered at the Municipal Transportation Service, 267 professionals were interviewed through a structured questionnaire. Then, a descriptive analysis was conducted and the incidence of accidents was estimated based on the variables studied. Relative risks were calculated and statistical significance was determined using the chi-square test and Fisher's exact test, considering p accidents were observed in 10.5% of mototaxis. There were mainly minor injuries (48.7%), 27% of them requiring leaves of absence from work. There was an association between the days of work per week, fatigue in lower limbs and musculoskeletal complaints, and accidents. Knowledge of the working conditions and accidents involved in this activity can be of great importance for the adoption of traffic education policies, and to help prevent accidents by improving the working conditions and lives of these professionals.

  6. Proceedings of the CSNI workshop on International Standard Problem 48 - Analysis of 1:4-scale prestressed concrete containment vessel model under severe accident conditions

    International Nuclear Information System (INIS)

    2005-01-01

    At the CSNI meeting in June 2002, the proposal for an International Standard Problem on containment integrity (ISP 48) based on the NRC/NUPEC/Sandia test was approved. Objectives were to extend the understanding of capacities of actual containment structures based on results of the recent PCCV Model test and other previous research. The ISP was sponsored by the USNRC, and results had been made available thanks to NUPEC and to the USNRC. Sandia National Laboratory was contracted to manage the technical aspects of the ISP. At the end of the ISP48, a workshop was organized in Lyon, France on April 6-7, 2005 hosted by Electricite de France. Its overall objective was to present results obtained by participants in the ISP 48 and to assess the current practices and the state of the art with respect to the calculation of concrete structures under severe accident conditions. Experience from other areas in civil engineering related to the modelling of complex structures was greatly beneficial to all. Information obtained as a result of this assessment were utilized to develop a consensus on these calculations and identify issues or 'gaps' in the present knowledge for the primary purpose of formulating and prioritizing research needs on this topic. The ISP48 exercise was published in the report referenced NEA/CSNI/R(2005)5 in 3 volumes. Volume 1 contains the synthesis of the exercise; Volumes 2 and 3 contain individual contributions of participating organizations. The CSNI Working Group on the Integrity and Ageing and in particular its sub-group on the behaviour of concrete structures has produced extensive material over the last few years. The complete list of references is given in this document. These proceedings gather the papers and presentations given by the participants at the Lyon workshop

  7. Technical Requirements and Principles for the Standards Development of the Key Parts for Rotor Air-conditioning Compressors

    Institute of Scientific and Technical Information of China (English)

    Sun Min; Wen Yun; Fan Zhangzeng

    2011-01-01

    ntroductionSince 2000,air-conditioning sales continues to grow,and the development of air-conditioning market makes a booming market of compressor.At the present time,compressor production rising all the way,and the sales steps up the new steps constantly.Tendency chart is shown in figure 1.Rotor compressor with its simple structure,small volume,light weight,easy processed mechanical parts,reliable operation and other excellent characteristics occupied the dominant position in the market.Compared with reciprocating compressor on the same application situation,decreased in the size by 40%~50%,weight was reduced by 40%~50%.But there were also disadvantages,mainly large friction loss,friction power consumption was about 10%of compressor's total power input.

  8. Study of the impact on PSA success criteria of the variability of the initial liquid level in case of the loss of the RHR system accident scenario under mid-loop operating conditions

    International Nuclear Information System (INIS)

    Villanueva, J.F.; Carlos, S.; Martorell, S.; Serradell, V.; Pelayo, F.; Mendizabal, R.; Cirauqui, C.; Sol, I.

    2005-01-01

    Probabilistic safety assessment (PSA) is recognized nowadays as an important tool to support risk-informed decision-making aimed at providing both operational flexibility and plant safety [1]. Experience of current PSA studies shows the importance of some risky scenarios with the plant at low power and shutdown conditions as compared to the accident scenarios with the plant operating at full power. In particular, current low power and shutdown PSA (LPSA) studies shows that the loss of the Residual Heat Removal System (RHRS) transient is one of the most risk-significant events under low power conditions [2]. This accident type is supposed to occur for various plant operating states, of which mid-loop operation represents one of the main contributors [3]. LPSA has widely used methods for thermal-hydraulic analysis that play an important role in determining success criteria of safety-related functions involved to mitigate the severity of accident scenarios with the plant operating in such conditions. Various best estimate thermal-hydraulic analysis codes have been used to analyze the loss of the RHRS during low power and shutdown conditions [4, 5]. It is known that RELAP code can give good results as derived after a number of benchmark exercises using results from experiments at research facilities (e.g. ROSA-IV, BETHSY, PKL). [6] Previous research has shown how thermal-hydraulic phenomena after the loss of the RHRS, e.g. peak reactor coolant system pressure, are sensitive to the initial liquid level at the time of loss of the RHRS [2]. This paper presents the results of the study of the thermalhydraulic analysis of the accident scenarios after the loss of the RHRS under mid-loop conditions paying particular attention to the analysis of the effect of the variability of the initial liquid level on the success criteria of the safety-related functions considered in a typical LPSA [3]. (author)

  9. Chernobyl accident. Exposures and effects

    International Nuclear Information System (INIS)

    Bennett, B.; Bouville, A.; Hall, P.; Savkin, M.; Storm, H.

    2000-01-01

    The Chernobyl accident that occurred in Ukraine in April 1986 happened during an experimental test of the electrical control system as the reactor was being shut down for routine maintenance. The operators, in violation of safety regulations, had switched off important control systems and allowed the reactor to reach unstable, low-power conditions. A sudden power surge caused a steam explosion that ruptured the reactor vessel and allowed further violent fuel-steam interactions that destroyed the reactor and the reactor building. The Chernobyl accident was the most serious to have ever occurred in the nuclear power industry. The accident caused the early death of 30 power plant employees and fire fighters and resulted in widespread radioactive contamination in areas of Belarus, the Russian Federation, and Ukraine inhabited by several million people. Radionuclides released from the reactor that caused exposure of individuals were mainly iodine-131, caesium-134 and caesium-137. Iodine-131 has a short radioactive half-life (8 days), but it can be transferred relatively rapidly through milk and leafy vegetables to humans. Iodine becomes localized in the thyroid gland. For reasons of intake of these foods, size of thyroid gland and metabolism, the thyroid doses are usually greater to infants and children than to adults. The isotopes of caesium have relatively long half-lives (caesium-134: 2 years; caesium-137: 30 years). These radionuclides cause long-term exposures through the ingestion pathway and from external exposure to these radionuclides deposited on the ground. In addition to radiation exposure, the accident caused long-term changes in the lives of people living in the contaminated regions, since measures intended to limit radiation doses included resettlements, changes in food supplies, and restrictions in activities of individuals and families. These changes were accompanied by major economic, social and political changes in the affected countries resulting

  10. Tchernobyl accident

    International Nuclear Information System (INIS)

    1986-06-01

    First, R.M.B.K type reactors are described. Then, safety problems are dealt with reactor control, behavior during transients, normal loss of power and behavior of the reactor in case of leak. A possible scenario of the accident of Tchernobyl is proposed: events before the explosion, possible initiators, possible scenario and events subsequent to the core meltdown (corium-concrete interaction, interaction with the groundwater table). An estimation of the source term is proposed first from the installation characteristics and the supposed scenario of the accident, and from the measurements in Europe; radiological consequences are also estimated. Radioactivity measurements (Europe, Scandinavia, Western Europe, France) are given in tables (meteorological maps and fallouts in Europe). Finally, a description of the site is given [fr

  11. Guidelines for calculating radiation doses to the public from a release of airborne radioactive material under hypothetical accident conditions in nuclear reactors

    International Nuclear Information System (INIS)

    1989-09-01

    This Standard provides guidelines and a methodology for calculating effective doses and thyroid doses to people (either individually or collectively) in the path of airborne radioactive material released from a nuclear facility following a hypothetical accident. The specific radionuclides considered in the Standard are those associated with substances having the greatest potential for becoming airborne in reactor accidents (eg, tritium (HTO), noble gases and their daughters (Kr-Rb, Xe-Cs), and radioiodines (I)); and certain radioactive particulates (eg, Cs, Ru, Sr, Te) that may become airborne under exceptional circumstances

  12. Will seasonal and climatic conditions influence living habits and socio-economic activities in such a way that nuclear accident are affected

    International Nuclear Information System (INIS)

    Baeverstam, U.

    1989-01-01

    The paper discusses to which extent climatic and seasonal effects can influence living habits and socio-economic activities in such a way that consequences of a nuclear accident might be affected. A number of examples from Sweden are given, related to dwellings (building standards and location), diet, seasonal effects in agriculture and tourism. The reindeer are discussed separately. Although climate and season do change man's habits in a way relevant to accident consequences, the conclusion of this paper is that in most cases this mechanism is severely mixed with other, sometimes more important ones

  13. Business incubation in a university as a key condition for the formation of innovational micro entrepreneurship in a region

    Directory of Open Access Journals (Sweden)

    Anatoliy Viktorovich Grebenkin

    2012-09-01

    Full Text Available This paper substantiates the hypothesis of the special role of universities in creating an environment of innovational micro entrepreneurship in a region. The role of business incubators is allocated; the algorithm for selecting projects is described. The results of a three-year organizational and economic experiment (with the changing conditions on the functioning of the student business incubator in the Ural State University are shown. Various models of the selection of ideas and projects for different cycles of incubation areimplemented. A decision on theestablishment of the Entrepreneurship Center in theInstitute of Management and Entrepreneurship is made. The Center’s main task is to form a series of events to support continuous generation of students’ business ideas, finding resonant response with the University experts and representatives of business environment in the region. A student in the business incubation system plays a new role for a Russian university — a role of a catalyst, i.e., directly acts as an element of positive feedback in the innovational system. It is shown that the catalytic path of the establishment and development of small high-tech business — Science to Business (StB — leads to the phenomenon of resonance, i.e., sustainable innovation flow generated by the business incubator of the University. The poll of the USU students in 2009-2011 (a sample from 660 to 854 respondents confirmed their positive attitude towards entrepreneurship and allowed to estimate the structure of the factors that hamper to increase student participation in the innovational business. Three blocks of factors were identified: the reluctance to take risks, inaccessibility of material and financial resources and the turbulence of the environment. A system of monitoring students' attitudes towards entrepreneurship, which allows adjusting the curriculum and creating institutional conditions for activation of innovative entrepreneurship

  14. Transuranics and fission products release from PWR fuels in severe accident conditions. Lessons learnt from VERCORS RT3 and RT4 tests

    International Nuclear Information System (INIS)

    Pontillon, Y.; Ducros, G.; Van Winckel, S.; Christiansen, B.; Kissane, M.P.; Dubourg, R.; Dutheillet, Y.; Andreo, F.

    2006-01-01

    Over the last decades, several experimental programs devoted to the source term of fission products (FP) and actinides released from PWR fuel samples in severe accident (SA) conditions have been initiated throughout the world. In France, in this context, the Institute for Radiological Protection and Safety (IRSN) and Electricite de France (EDF) have supported the analytical VERCORS program which was performed by the Commissariat a l'Energie Atomique (CEA). The VERCORS facility at the LAMA-laboratory (CEA-Grenoble, France) was designed to heat up an irradiated fuel sample - taken from EDF's nuclear power reactors - to fuel relocation, and to capture the fission products released from the fuel and deposited downstream on a series of specific filters (impactors, bead-bed filter). On-line gamma detectors aimed at the fuel position, filters and gas capacity monitored the progress of FP release from the fuel, FP deposition on the filters and the fission gases emitted by the fuel (xenon and krypton). Before and after the test, a longitudinal gamma-scan of the fuel was conducted to measure the initial and final FP inventory in order to evaluate the quantitative fractions of FP emitted by the fuel during the test. All the components of the loop were then gamma-scanned to measure and locate the FPs released during the test and to draw up a mass balance of these FP. 25 annealing tests were performed between 1983 and 2002 on irradiated PWR fuels under various conditions of temperature and atmospheres (oxidising or reducing conditions). The influence of the nature of the fuel (UO 2 versus MOX, burn up) and the fuel morphology (initially intact or fragmented fuel) have also been investigated. This led to an extended data base allowing on the one hand to study mechanisms which promote FP release in SA conditions, and on the other hand to enhance models implemented in SA codes. Because gamma spectrometry is well suited to FP measurement and not to actinides (except neptunium

  15. Accident: Reminder

    CERN Multimedia

    2003-01-01

    There is no left turn to Point 1 from the customs, direction CERN. A terrible accident happened last week on the Route de Meyrin just outside Entrance B because traffic regulations were not respected. You are reminded that when travelling from the customs, direction CERN, turning left to Point 1 is forbidden. Access to Point 1 from the customs is only via entering CERN, going down to the roundabout and coming back up to the traffic lights at Entrance B

  16. Expert software for accident identification

    International Nuclear Information System (INIS)

    Dobnikar, M.; Nemec, T.; Muehleisen, A.

    2003-01-01

    Each type of an accident in a Nuclear Power Plant (NPP) causes immediately after the start of the accident variations of physical parameters that are typical for that type of the accident thus enabling its identification. Examples of these parameter are: decrease of reactor coolant system pressure, increase of radiation level in the containment, increase of pressure in the containment. An expert software enabling a fast preliminary identification of the type of the accident in Krsko NPP has been developed. As input data selected typical parameters from Emergency Response Data System (ERDS) of the Krsko NPP are used. Based on these parameters the expert software identifies the type of the accident and also provides the user with appropriate references (past analyses and other documentation of such an accident). The expert software is to be used as a support tool by an expert team that forms in case of an emergency at Slovenian Nuclear Safety Administration (SNSA) with the task to determine the cause of the accident, its most probable scenario and the source term. The expert software should provide initial identification of the event, while the final one is still to be made after appropriate assessment of the event by the expert group considering possibility of non-typical events, multiple causes, initial conditions, influences of operators' actions etc. The expert software can be also used as an educational/training tool and even as a simple database of available accident analyses. (author)

  17. Thiazolidine-2,4-dione derivatives: programmed chemical weapons for key protein targets of various pathological conditions.

    Science.gov (United States)

    Chadha, Navriti; Bahia, Malkeet Singh; Kaur, Maninder; Silakari, Om

    2015-07-01

    Thiazolidine-2,4-dione is an extensively explored heterocyclic nucleus for designing of novel agents implicated for a wide variety of pathophysiological conditions, that is, diabetes, diabetic complications, cancer, arthritis, inflammation, microbial infection, and melanoma, etc. The current paradigm of drug development has shifted to the structure-based drug design, since high-throughput screenings have continued to generate disappointing results. The gap between hit generation and drug establishment can be narrowed down by investigation of ligand interactions with its receptor protein. Therefore, it would always be highly beneficial to gain knowledge of molecular level interactions between specific protein target and developed ligands; since this information can be maneuvered to design new molecules with improved protein fitting. Thus, considering this aspect, we have corroborated the information about molecular (target) level implementations of thiazolidine-2,4-diones (TZD) derivatives having therapeutic implementations such as, but not limited to, anti-diabetic (glitazones), anti-cancer, anti-arthritic, anti-inflammatory, anti-oxidant and anti-microbial, etc. The structure based SAR of TZD derivatives for various protein targets would serve as a benchmark for the alteration of existing ligands to design new ones with better binding interactions. Copyright © 2015 Elsevier Ltd. All rights reserved.

  18. The Key to Acetate: Metabolic Fluxes of Acetic Acid Bacteria under Cocoa Pulp Fermentation-Simulating Conditions

    Science.gov (United States)

    Adler, Philipp; Frey, Lasse Jannis; Berger, Antje; Bolten, Christoph Josef; Hansen, Carl Erik

    2014-01-01

    Acetic acid bacteria (AAB) play an important role during cocoa fermentation, as their main product, acetate, is a major driver for the development of the desired cocoa flavors. Here, we investigated the specialized metabolism of these bacteria under cocoa pulp fermentation-simulating conditions. A carefully designed combination of parallel 13C isotope labeling experiments allowed the elucidation of intracellular fluxes in the complex environment of cocoa pulp, when lactate and ethanol were included as primary substrates among undefined ingredients. We demonstrate that AAB exhibit a functionally separated metabolism during coconsumption of two-carbon and three-carbon substrates. Acetate is almost exclusively derived from ethanol, while lactate serves for the formation of acetoin and biomass building blocks. Although this is suboptimal for cellular energetics, this allows maximized growth and conversion rates. The functional separation results from a lack of phosphoenolpyruvate carboxykinase and malic enzymes, typically present in bacteria to interconnect metabolism. In fact, gluconeogenesis is driven by pyruvate phosphate dikinase. Consequently, a balanced ratio of lactate and ethanol is important for the optimum performance of AAB. As lactate and ethanol are individually supplied by lactic acid bacteria and yeasts during the initial phase of cocoa fermentation, respectively, this underlines the importance of a well-balanced microbial consortium for a successful fermentation process. Indeed, AAB performed the best and produced the largest amounts of acetate in mixed culture experiments when lactic acid bacteria and yeasts were both present. PMID:24837393

  19. The role of radiation protection in the rehabilitation of living conditions in contaminated territories after a nuclear accident or a radiological event

    International Nuclear Information System (INIS)

    Lochard, J.

    2006-01-01

    The implementation of radiation protection programmes in long -term contaminated territories is raising difficult technical, methodological, but also ethical questions, which commit the responsibility of the radiation protection professionals and the way they are exercising their expertise in such situation. The presence of contamination in the environment is a new reality which is profoundly affecting the living conditions of the population in all aspects: health, environment but also the psychological, social and economical cultural dimensions of daily life. In such a context, the classical radiation protection approach is inoperative to restore the normality of the ante accident situation On the contrary, the scientific and technological expertise and the prescriptive administrative arrangements, including the radiation protection ones, which are heavily mobilized to respond to the many challenges facing the society induce a progressive de -qualification of the environment and goods and a loss of control of the situation by the population which turn into a social depression. The paper illustrates how radiation protection programmes for managing long term contaminated territories have an ambivalent impact: on the one hand they ensure the protection of the population against the potential health impacts of radiations, on the other hand they re -enforce the loss of control, de-qualification and exclusion processes which results in a weakening of the resilience capacity of the population and the territories. The recent experience of co-expertise and co-operation implemented in the contaminated territories of Belarus have demonstrated that the direct engagement of local professionals and the population in the daily management of the radiological situation is not only feasible but also necessary to stop this vicious circle of loss of control and exclusion. To be effective and sustainable this engagement must rely on the diffusion among the population of a 'practical

  20. Prevention of pedestrian accidents.

    OpenAIRE

    Kendrick, D

    1993-01-01

    Child pedestrian accidents are the most common road traffic accident resulting in injury. Much of the existing work on road traffic accidents is based on analysing clusters of accidents despite evidence that child pedestrian accidents tend to be more dispersed than this. This paper analyses pedestrian accidents in 573 children aged 0-11 years by a locally derived deprivation score for the years 1988-90. The analysis shows a significantly higher accident rate in deprived areas and a dose respo...

  1. [Drugs and occupational accident].

    Science.gov (United States)

    Bratzke, H; Albers, C

    1996-02-01

    In a case of a fatal occupational accident (construction worker, fall from roof, urine test positive for cocaine and THC, e.g. cannabis) the question arised to what extent those drug-related occupational accidents occur. In the literature only few cases, mainly dealing with cannabis influence, have been reported, however, a higher number is suspected. Cocaine and other stimulating drugs (amphetamine) are more often used to increase physical fitness. By direct or indirect interference with vigilance these compounds may provoke accidents. Due to the lack of a legal basis proving of the influence of drugs at the working place is still very limited, although highly sensitive chemical-toxicological assay procedures are available to detect even the chronic abuse (in hair). In the general conditions of accident insurances a compensation is excluded when alcohol is involved, but drugs are not mentioned. It is indeed difficult to establish a concentration limit for drugs like that existing for alcohol (1.1%). In each case the assay of the drug involved and exact knowledge of its specific effects is in an essential prerequisite to prove the causal relationship.

  2. Output signal analysis for a variation of the R-C passive elements in a 4-2 mA R-L-C equivalent circuit modeling under a high temperature accident condition in NPPs

    International Nuclear Information System (INIS)

    Kil-Mo, Koo; Sang-Baik, Kim; Hee-Dong, Kim; Gyu-Tae, Kim

    2007-01-01

    An electrical signal should be checked to see whether it lies within its expected electrical range when there is a doubtful condition. The normal signal level for pressure, flow, level and resistance temperature detector sensors is 4-20 mA in most industrial process controls. In the case of an abnormal signal level from an instrument under a severe accident condition, it is necessary to obtain a more accurate signal validation to operate a system in a control room in NPPs. Diagnostics and analysis for some abnormal signals have been performed through an important equivalent circuits modeling for passive elements under severe accident conditions. Unlike the design basis accidents, there are some inherent uncertainties for the instrumentation capabilities under severe accident conditions. In this paper, to implement a diagnostic analysis for an equivalent circuits modeling, a kind of linked LabVIEW program for each PSpice and MULTI-SIM code is introduced as a one body order system, which can obtain some abnormal signal patterns by a special function such as an advanced simulation tool for each PSpice and Multi-SIM code as a means of a function for a PC based ASSA (Abnormal Signal Simulation Analyzer) module. The output signal can be analyzed by a comparative analysis of each PSpice and Multi-SIM code for a 4-20 mA circuit modeling which is by a composition of an R-L-C passive circuit as an alternating range of the elements for the temperature accident condition. In this simulation, a new simulator through an analysis of the important equivalent circuits modeling has been designed, the designed simulator is composed of the LabVIEW code as a main tool and the out-put file of each PSpice code and a Multi-SIM engine code as an engine tool is exported to the in-put file of the LabVIEW code. There are 3 main function units of the ASSA module, the first one is individual PSpice and Multi-SIM engine code units to comprise an equivalent circuit element, the second one is the

  3. Investigation on accident management measures for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Tusheva, P.; Schaefer, F.; Rohde, U.; Reinke, N.

    2009-01-01

    A consequence of a total loss of AC power supply (station blackout) leading to unavailability of major active safety systems which could not perform their safety functions is that the safety criteria ensuring a secure operation of the nuclear power plant would be violated and a consequent core heat-up with possible core degradation would occur. Currently, a study which examines the thermal-hydraulic behaviour of the plant during the early phase of the scenario is being performed. This paper focuses on the possibilities for delay or mitigation of the accident sequence to progress into a severe one by applying Accident Management Measures (AMM). The strategy 'Primary circuit depressurization' as a basic strategy, which is realized in the management of severe accidents is being investigated. By reducing the load over the vessel under severe accident conditions, prerequisites for maintaining the integrity of the primary circuit are being created. The time-margins for operators' intervention as key issues are being also assessed. The task is accomplished by applying the GRS thermal-hydraulic system code ATHLET. In addition, a comparative analysis of the accident progression for a station blackout event for both a reference German PWR and a reference VVER-1000, taking into account the plant specifics, is being performed. (authors)

  4. Radiation accidents

    International Nuclear Information System (INIS)

    Saenger, E.L.

    1986-01-01

    It is essential that emergency physicians understand ways to manage patients contaminated by radioactive materials and/or exposed to external radiation sources. Contamination accidents require careful surveys to identify the metabolic pathway of the radionuclides to guide prognosis and treatment. The level of treatment required will depend on careful surveys and meticulous decontamination. There is no specific therapy for the acute radiation syndrome. Prophylactic antibodies are desirable. For severely exposed patients treatment is similar to the supportive care given to patients undergoing organ transplantation. For high-dose extremity injury, no methods have been developed to reverse the fibrosing endarteritis that eventually leads to tissue death so frequently found with this type of injury. Although the Three Mile Island episode of March 1979 created tremendous public concern, there were no radiation injuries. The contamination outside the reactor building and the release of radioiodine were negligible. The accidental fuel element meltdown at Chernobyl, USSR, resulted in many cases of acute radiation syndrome. More than 100,000 people were exposed to high levels of radioactive fallout. The general principles outlined here are applicable to accidents of that degree of severity

  5. Chernobyl accident

    International Nuclear Information System (INIS)

    Bar'yakhtar, V.G.

    1995-01-01

    The monograph contains the catastrophe's events chronology, the efficiency assessed of those measures assumed for their localization as well as their environmental and socio-economic impact. Among materials of the monograph the results are presented of research on the radioactive contamination field forming as well as those concerning the investigation of biogeochemical properties of Chernobyl radionuclides and their migration process in the environment of the Ukraine. The data dealing with biological effects of the continued combined internal and external radioactive influence on plants, animals and human health under the circumstances of Chernobyl accident are of the special interest. In order to provide the scientific generalizing information on the medical aspects of Chernobyl catastrophe, the great part of the monograph is allotted to appraise those factors affecting the health of different population groups as well as to depict clinic aspects of Chernobyl events and medico-sanitarian help system. The National Programme of Ukraine for the accident consequences elimination and population social protection assuring for the years 1986-1993 and this Programme concept for the period up to the year 2000 with a special regard of the world community participation there

  6. Guidelines for calculating radiation doses to the public from a release of airborne radioactive material under hypothetical accident conditions in nuclear reactors

    International Nuclear Information System (INIS)

    1991-04-01

    This standard provides guidelines and a methodology for calculating effective doses and thyroid doses to people (either individually or collectively) in the path of airborne radioactive material released from a nuclear facility following a hypothetical accident. The radionuclides considered are those associated with substances having the greatest potential for becoming airborne in reactor accidents: tritium (HTO), noble gases and their daughters, radioiodines, and certain radioactive particulates (Cs, Ru, Sr, Te). The standard focuses on the calculation of radiation doses for external exposures from radioactive material in the cloud; internal exposures for inhalation of radioactive material in the cloud and skin penetration of tritium; and external exposures from radionuclides deposited on the ground. It uses as modified Gaussian plume model to evaluate the time-integrated concentration downwind. (52 refs., 12 tabs., 21 figs.)

  7. Reactor accident in Chernobyl

    International Nuclear Information System (INIS)

    Sokalski, A.; Kowalski, A.

    1990-11-01

    The bibliography contains 1568 descriptions of papers devoted to Chernobylsk accident and recorded in ''INIS Atomindex'' to 30 June 1990. The descriptions were taken from ''INIS Atomindex'' and are presented in accordance with volumes of this journal (chronology of recording). Therefore all descriptions have numbers showing first the number of volume and then the number of record. The bibliography has at the end the detailed subject index consisting of 465 main headings and a lot of qualifiers. Some of them are descriptors taken from ''INIS Atomindex'' and some are key words taken from natural language. The index is in English as descriptions in the bibliography. (author)

  8. The Development of a PiSA Module for a Diagnosis of Instrumental Signals Associated with an ASSA Module for Accident Controls

    International Nuclear Information System (INIS)

    Koo, Kil Mo; Ahn, Kwang Il

    2010-05-01

    A review of a plant's accident management capabilities is one of the key elements in achieving regulatory closure of severe accident issues. During accidents, information and data from plant's instruments, as well as others sources, are essential for assessing the plant's status and response. Unlike for design basis accidents, there are inherently some uncertainties to instrumentation capabilities for severe accident conditions. There are many ways to obtain information during a severe accident. Moreover, precise measurements are not necessary. The redundancy and ruggedness of a plant's instrumentation provides considerable depth in the capability of existing designs. The circuit simulation analysis and diagnosis methods are used to assess instruments in detail when they give apparently abnormal readings. A new simulator, PiSA(Provability Instrument Signal Analysis) associated with ASSA(Abnormal Signal Simulation Analysis), through an analysis of the important circuits modeling under severe accident conditions has been designed. It has three main functions which are a signal processing tool, an accident management tool, and an additional guide from the initial screen. In this report, it can be simulated to the temperature characteristic analysis procedure of the PiSA including ASSA's data comparative methods and using specific signal processing under severe accident condition

  9. A cliff edge evaluation for CANDU-6 beyond design basis accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.M.; Kho, D.W., E-mail: wolsong@khnp.co.kr [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Yi, S.D.; Kang, S.H.; Kim, S.R. [Nuclear Engineering Service and Solution Co., Ltd., Daejeon (Korea, Republic of)

    2015-07-01

    The condition of nuclear power plant in the event of station black out (SBO) accompanying large-scale natural disaster exceeding design basis accident (DBA) was evaluated. Additional scenarios were added to the evaluation to review capability of the plant to endure different conditions with different actions. The analysis resulted that the key action required from the operator was to ensure the opening of main steam safety valves (MSSVs) in the secondary side and of motor-operated valves for high pressure injection of Emergency Core Cooling System (HPECCS) to mitigate accidents or extend the cliff edge. (author)

  10. The Human Aspect of the Fukushima Daiichi Accident

    International Nuclear Information System (INIS)

    Anegawa, T.; Kawano, A.

    2016-01-01

    Recognizing itself as the main party involved in the nuclear accident triggered by the Tohoku-Chihou-Taiheiyo-Oki Earthquake on March 11, 2011, Tokyo Electric Power Company (TEPCO) has performed accident investigation from various aspects. Results of the investigation are reported mainly in two reports; (1) Fukushima Nuclear Accident Analysis Report (June 20, 2012), which identified the timeline and the proximate causes of the accident, and (2) Summary of Fukushima Nuclear Accident and Nuclear Safety Reform Plan (March 29, 2013) to set forth the results of the investigation and provide an analysis of the background factors surrounding the accident and countermeasures taken. This presentation will first provide overview of the accident response at Fukushima Daiichi and Daini Nuclear Power Stations. Voices from the first responders at the sites will be introduced in order to share thoughts of individuals involved in the emergency response. Summary of retrospective study of the accident by one of the shift supervisors at the time of the accident will be presented in order to share the facts that happened at main control rooms. The shift supervisor and his crew had to manage the situation for extended period of time that exceeded the scenarios that they had been trained, in a situation with no lightning and high radiation condition. During the accident response, shift supervisors had to decide to dispatch some of his crew members to the field to open valves, check the status of equipment etc., in the situation where the high radiation exposure is foreseen. The presentation will include conflict of shift supervisors and crew focusing on the human aspects. In addition, actions being taken at the Emergency Response Centers (ERC) set up at the seismic-isolated building on-site and the Headquarters in Tokyo will be shared focusing on the human aspects related to the accident progress. This includes difficult decisions to dispatch first responders to the field, in the

  11. Application of the accident management information needs methodology to a severe accident sequence

    International Nuclear Information System (INIS)

    Ward, L.W.; Hanson, D.J.; Nelson, W.R.; Solberg, D.E.

    1989-01-01

    The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel

  12. Application of the accident management information needs methodology to a severe accident sequence

    Energy Technology Data Exchange (ETDEWEB)

    Ward, L.W.; Hanson, D.J.; Nelson, W.R. (Idaho National Engineering Laboratory, Idaho Falls (USA)); Solberg, D.E. (Nuclear Regulatory Commission, Washington, DC (USA))

    1989-11-01

    The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel.

  13. Use of accident experience in developing criteria for teleoperator equipment

    International Nuclear Information System (INIS)

    Vallario, E.J.; Selby, J.M.

    1985-10-01

    The 1961 SL-1 reactor accident in Idaho and the Recuplex accident at Hanford are reviewed to identify problems common to emergency situations, lessons learned from accidents, criteria for emergency equipment, and recommendations for using robotics to solve problems during emergencies. Teleoperator equipment could be used to assess the extent of the damage and the condition of the reactor, retrieve dosimeters, evacuate and treat accident victims, clean up debris and decontaminate accident areas. 2 refs., 9 figs

  14. Chernobyl accident

    International Nuclear Information System (INIS)

    Capra, D.; Facchini, U.; Gianelle, V.; Ravasini, G.; Bacci, P.

    1988-01-01

    The radioactive cloud released during the Chernobyl accident reached the Padana plain and Lombardy in the night of April 30th 1986; the cloud remained in the northern Italian skies for a few days and then disappeared either dispersed by winds and washed by rains. The evidence in atmosphere of radionuclides as Tellurium, Iodine, Cesium, was promptly observed. The intense rain, in first week of may, washed the radioactivity and fall-out contamined the land, soil, grass. The present work concerns the overall contamination of the Northern Italy territory and in particular the radioactive fall-out in the Lakes region. Samples of soil have been measured at the gamma spectroscope; a correlation is found between the radionuclides concentration in soil samples and the rain intensity, when appropriate deposition models are considered. A number of measurements has been done on the Como'lake ecosystem: sediments, plankton, fishes and the overall fall-out in the area has been investigated

  15. 43 CFR 15.13 - Report of accidents.

    Science.gov (United States)

    2010-10-01

    ... 43 Public Lands: Interior 1 2010-10-01 2010-10-01 false Report of accidents. 15.13 Section 15.13 Public Lands: Interior Office of the Secretary of the Interior KEY LARGO CORAL REEF PRESERVE § 15.13 Report of accidents. Accidents involving injury to life or property shall be reported as soon as possible...

  16. Self-reported accidents

    DEFF Research Database (Denmark)

    Møller, Katrine Meltofte; Andersen, Camilla Sloth

    2016-01-01

    The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals.......The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals....

  17. Accident Assessment

    International Nuclear Information System (INIS)

    Tripputi, Ivo; Lund, Ingemar

    2002-01-01

    There is a general feeling that decommissioning is an activity involving limited risks, compared to NPP operation, and in particular risks involving the general public. This is technically confirmed by licensing analysis and evaluations, where, once the spent fuel has been removed from the plant, the radioactivity inventory available to be released to the environment is very limited. Decommissioning activities performed so far in the world have also confirmed the first assumptions and no specific issue has been identified, in this field, to justify a completely new approach. Commercial interests in international harmonization, which could drive an in-depth discussion about the bases of this approach, are weak at the moment. However, there are several reasons why a discussion in an international framework about the Safety Case for decommissioning (and, in particular, about Accident Assessment) may be considered necessary and important, and why it may show some specific and peculiar aspects. An effort for a comprehensive and systematic D and D accident safety assessment of the decommissioning process is justified. It is necessary also to explore in a holistic way the aspects of industrial safety, and develop tools for the decision-making process optimization. The expected results are the implementation of appropriate and optimized protective measures in any event and of adequate on/off-site emergency plans for optimal public and workers protection. The experience from other decommissioning projects and large-scale industrial activities is essential to balance provisions and an Operating Experience review process (specific for decommissioning) should help to focus on real issues

  18. JAERI's activities in JCO accident

    International Nuclear Information System (INIS)

    2000-09-01

    The Japan Atomic Energy Research Institute (JAERI) was actively involved in a variety of technical supports and cooperative activities, such as advice on terminating the criticality condition, contamination checks of the residents and consultation services for the residents, as emergency response actions to the criticality accident at the uranium processing facility operated by the JCO Co. Ltd., which occurred on September 30, 1999. These activities were carried out in collaborative ways by the JAERI staff from the Tokai Research Establishment, Naka Fusion Research Establishment, Oarai Research Establishment, and Headquarter Office in Tokyo. As well, the JAERI was engaged in the post-accident activities such as identification of accident causes, analyses of the criticality accident, and dose assessment of exposed residents, to support the Headquarter for Accident Countermeasures of the Science and Technology Agency (STA), the Accident Investigation Committee and the Health Control Committee of the Nuclear Safety Commission of Japan (NSC). This report compiles the activitie