Investigation of Countercurrent Helium-Air Flows in Air-ingress Accidents for VHTRs

Energy Technology Data Exchange (ETDEWEB)

Sun, Xiaodong; Christensen, Richard; Oh, Chang

2013-10-03

The primary objective of this research is to develop an extensive experimental database for the air- ingress phenomenon for the validation of computational fluid dynamics (CFD) analyses. This research is intended to be a separate-effects experimental study. However, the project team will perform a careful scaling analysis prior to designing a scaled-down test facility in order to closely tie this research with the real application. As a reference design in this study, the team will use the 600 MWth gas turbine modular helium reactor (GT-MHR) developed by General Atomic. In the test matrix of the experiments, researchers will vary the temperature and pressure of the helium— along with break size, location, shape, and orientation—to simulate deferent scenarios and to identify potential mitigation strategies. Under support of the Department of Energy, a high-temperature helium test facility has been designed and is currently being constructed at Ohio State University, primarily for high- temperature compact heat exchanger testing for the VHTR program. Once the facility is in operation (expected April 2009), this study will utilize high-temperature helium up to 900°C and 3 MPa for loss-of-coolant accident (LOCA) depressurization and air-ingress experiments. The project team will first conduct a scaling study and then design an air-ingress test facility. The major parameter to be measured in the experiments is oxygen (or nitrogen) concentration history at various locations following a LOCA scenario. The team will use two measurement techniques: 1) oxygen (or similar type) sensors employed in the flow field, which will introduce some undesirable intrusiveness, disturbing the flow, and 2) a planar laser-induced fluorescence (PLIF) imaging technique, which has no physical intrusiveness to the flow but requires a transparent window or test section that the laser beam can penetrate. The team will construct two test facilities, one for high-temperature helium tests with

FINAL REPORT on Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

Energy Technology Data Exchange (ETDEWEB)

Chang H. Oh; Eung S. Kim; Hee C. NO; Nam Z. Cho

2011-01-01

The U.S. Department of Energy is performing research and development that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Generation IV very high temperature reactor (VHTR). Phenomena Identification and Ranking studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important. Consequently, the development of advanced air ingress-related models and verification & validation are of very high priority for the NGNP Project. Following a loss of coolant and system depressurization incident, air ingress will occur through the break, leading to oxidation of the in-core graphite structure and fuel. This study indicates that depending on the location and the size of the pipe break, the air ingress phenomena are different. In an effort to estimate the proper safety margin, experimental data and tools, including accurate multidimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model are required. It will also require effective strategies to mitigate the effects of oxidation, eventually. This 3-year project (FY 2008–FY 2010) is focused on various issues related to the VHTR air-ingress accident, including (a) analytical and experimental study of air ingress caused by density-driven, stratified, countercurrent flow, (b) advanced graphite oxidation experiments, (c) experimental study of burn-off in the core bottom structures, (d) structural tests of the oxidized core bottom structures, (e) implementation of advanced models developed during the previous tasks into the GAMMA code, (f) full air ingress and oxidation mitigation analyses, (g) development of core neutronic models, (h) coupling of the core neutronic and thermal hydraulic models, and (i) verification and validation of the coupled models.

Cladding oxidation during air ingress. Part II: Synthesis of modelling results

International Nuclear Information System (INIS)

Beuzet, E.; Haurais, F.; Bals, C.; Coindreau, O.; Fernandez-Moguel, L.; Vasiliev, A.; Park, S.

2016-01-01

Highlights: • A state-of-the-art for air oxidation modelling in the frame of severe accident is done. • Air oxidation models from main severe accident codes are detailed. • Simulations from main severe accident codes are compared against experimental results. • Perspectives in terms of need for further model development and experiments are given. - Abstract: Air ingress is a potential risk in some low probable situations of severe accidents in a nuclear power plant. Air is a highly oxidizing atmosphere that can lead to an enhanced Zr-based cladding oxidation and core degradation affecting the release of fission products. This is particularly true speaking about ruthenium release, due to its high radiotoxicity and its ability to form highly volatile oxides in a significant manner in presence of air. The oxygen affinity is decreasing from the Zircaloy cladding, fuel and ruthenium inclusions. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues in such scenarios. In the past years, many works have been done on cladding oxidation by air under severe accident conditions. This paper with in addition the paper “Cladding oxidation during air ingress – Part I: Synthesis of experimental results” of this journal issue aim at assessing the state of the art on this phenomenon. In this paper, the modelling of air ingress phenomena in the main severe accident codes (ASTEC, ATHLET-CD, MAAP, MELCOR, RELAP/SCDAPSIM, SOCRAT) is described in details, as well as the validation against the integral experiments QUENCH-10, QUENCH-16 and PARAMETER-SF4. A full review of cladding oxidation by air is thus established.

Studies on air ingress for pebble bed reactors

International Nuclear Information System (INIS)

Moore, R.L.; Oh, C.H.; Merrill, B.J.; Petti, D.A.

2002-01-01

A loss-of-coolant accident (LOCA) has been considered a critical event for helium-cooled pebbled bed reactors. Following helium depressurization, it is anticipated that unless countermeasures are taken air will enter the core through the break and then by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure and graphite pebbles. Thus, without any mitigating features a LOCA will lead to an air ingress event. The INEEL is studying such an event with two well-respected light water reactor transient response codes: RELAP5/ATHENA and MELCOR. To study the degree of graphite oxidation occurring due to an air ingress event, a MELCOR model of a reference pebble bed design was constructed. A modified version of MELCOR developed at INEEL, which includes graphite oxidation capabilities, and molecular diffusion of air into helium was used for these calculations. Results show that the lower reflector graphite consumes all of the oxygen before reaching the core. The results also show a long time delay between the time that the depressurization phase of the accident is over and the time that natural circulation air through the core occurs. (author)

FY-09 Report: Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

Energy Technology Data Exchange (ETDEWEB)

Chang H. Oh; Eung S. Kim

2009-12-01

The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is performing research and development that focuses on key phenomena important during potential scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Gen-IV very high temperature reactor (VHTR). Phenomena Identification and Ranking Studies to date have identified that an air ingress event following on the heels of a VHTR depressurization is a very important incident. Consequently, the development of advanced air ingress-related models and verification and validation data are a very high priority for the NGNP Project. Following a loss of coolant and system depressurization incident, air will enter the core through the break, leading to oxidation of the in-core graphite structure and fuel. If this accident occurs, the oxidation will accelerate heat-up of the bottom reflector and the reactor core and will eventually cause the release of fission products. The potential collapse of the core bottom structures causing the release of CO and fission products is one of the concerns. Therefore, experimental validation with the analytical model and computational fluid dynamic (CFD) model developed in this study is very important. Estimating the proper safety margin will require experimental data and tools, including accurate multidimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model. It will also require effective strategies to mitigate the effects of oxidation. The results from this research will provide crucial inputs to the INL NGNP/VHTR Methods Research and Development project. The second year of this three-year project (FY-08 to FY-10) was focused on (a) the analytical, CFD, and experimental study of air ingress caused by density-driven, stratified, countercurrent flow; (b) advanced graphite oxidation experiments and modeling; (c) experimental study of burn-off in the core bottom structures, (d) implementation of advanced

Oxidation kinetics of innovative carbon materials with respect to severe air ingress accidents in HTRs and graphite disposal or processing

International Nuclear Information System (INIS)

Schloegel, Baerbel

2010-01-01

Currently future nuclear reactor concepts of the Fourth Generation (Gen IV) are under development. To some extend they apply with new, innovative materials developed just for this purpose. This thesis work aims at a concept of Generation IV Very High Temperature Reactors (VHTR) in the framework of the European project RAPHAEL (ReActor for Process heat, Hydrogen And ELectricity generation). The concept named ANTARES (AREVA New Technology based on advanced gas-cooled Reactors for Energy Supply) was developed by AEVA NP. It is a helium cooled, graphite moderated modular reactor for electricity and hydrogen production, by providing the necessary process heat due to its high working temperature. Particular attention is given here to oxidation kinetics of newly developed carbon materials (NBG-17) with still unknown but needed information in context of severe air ingress accident in VHTR's. Special interest is paid to the Boudouard reaction, the oxidation of carbon by CO 2 . In case of an air ingress accident, carbon dioxide is produced in the primary reaction of atmospheric oxygen with reflector graphite. From there CO 2 could flow into the reactor core causing further damage by conversion into CO. The purpose of this thesis is to ascertain if and to what degree this could happen. First of all oxidation kinetic data of the Boudouard reaction with NBG-17 is determined by experiments in a thermo gravimetric facility. The measurements are evaluated and converted into a common formula and a Langmuir-Hinshelwood similar oxidation kinetic equation, as input for the computer code REACT/THERMIX. This code is then applied to analyse severe air ingress accidents for several air flow rates. The results are discussed for two accident situations, in which a certain graphite burn off is achieved. All cases show much more damage to the graphite bottom reflector than to the reactor core. Thus the bottom reflector will lose its structural integrity much earlier than the core itself will

Study air ingress into the reactor vessel using ICARE/CATHARE V2.0 in case of severe accident

International Nuclear Information System (INIS)

Gwenaelle Le Dantec; Fichot, F.

2005-01-01

Full text of publication follows: Safety analyses show that core degradation during a severe reactor accident would not be uniform. This was confirmed by TMI2 examinations. In fact, a central region of the core may overheat, melt and flow down to the lower plenum of the reactor while peripheral regions of the core would remain almost intact. Following rupture of the vessel by molten debris, air may be drawn from the containment by natural convection into the reactor coolant system, and react with the intact rods. Studying air ingress into the reactor vessel is of interest because the interaction of air with Zircaloy cladding can strongly affect the evolution of severe accident scenarios. The main effects are heat generation, increasing clad degradation, fission product release and nitriding. In case of air/steam recirculation in the vessel, significant nitriding of cladding can occur. The resulting ZrN phase is characterized by its brittleness and instability under oxidizing conditions, Oxidation of pre-existing ZrN phase layers has been observed to result in violent oxidation and heat release. Therefore, the first consequence for safety is a risk of strong deflagration in the vessel if a large number of rods on which a substantial layer of ZrN has grown are suddenly in contact with oxygen or steam. The second consequence is a late melting of core materials due to the very exothermic oxidation, leading to a late release of materials out of the reactor pressure vessel (RPV). In this paper we present an ICARE/CATHARE V2.0 calculation simulating air ingress into the vessel and in particular to describe the nitriding due to natural convection in the reactor vessel. The basic modeling and the necessary extensions of both ICARE and CATHARE are explained. The natural circulation is calculated to predict the regions of oxygen starvation where nitriding takes place. Key words: air ingress, nitriding, ICARE/CATHARE V2.0. (authors)

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.

Modeling of the consequences of an air ingress in the vessel on the source term calculation in a severe accident

International Nuclear Information System (INIS)

Beuzet, E.

2010-01-01

This work deals with severe accidents in nuclear power plants with an air ingress in the vessel. Air, which is very oxidizing, enhances noticeably ruthenium release, due to a high fuel oxidation. Oxygen affinity is decreasing between zircaloy cladding, fuel and ruthenium which is released from fuel as volatile and radio-toxic oxides. The main objective was to develop models for cladding and fuel oxidation and ruthenium release and implement them in the severe accident code MAAP. For each phenomenon, the literature study contributes to establish numerical models, validated against specific experiments. Cladding oxidation by air leads to a dense zirconia layer and, after cracking, a porous zirconia. Its modelling was validated against QUENCH-10 (KIT) and PARAMETER-SF4 (LUCH) experiments and 'NUREG' oxide mass gain law was selected as the best current compromise. Ruthenium release is modelled as an instantaneous oxide volatilisation at the fuel surface, controlled by fuel oxidation, which is itself based on a surface reaction assumption. These models were validated against a selected set of VERCORS (CEA) experiments. Ruthenium release is well-reproduced, particularly for air and steam atmospheres. To finish, an accidental scenario with air ingress in the vessel was simulated: a first value was given for the time needed to completely release ruthenium still present in core and was compared to the other studies. This simulation underlines the high ruthenium release under air conditions. (author)

Analysis of helium purification system capability during water ingress accident in RDE

Science.gov (United States)

Sriyono; Kusmastuti, Rahayu; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

The water ingress accident caused by steam generator tube rupture (SGTR) in RDE (Experimental Power Reactor) must be anticipated. During the accident, steam from secondary system diffused and mixed with helium gas in the primary coolant. To avoid graphite corrosion in the core, steam will be removed by Helium purification system (HPS). There are two trains in HPS, first train for normal operation and the second for the regeneration and accident. The second train is responsible to clean the coolant during accident condition. The second train is equipped with additional component, i.e. water cooler, post accident blower, and water separator to remove this mixture gas. During water ingress, the water release from rupture tube is mixed with helium gas. The water cooler acts as a steam condenser, where the steam will be separated by water separator from the helium gas. This paper analyses capability of HPS during water ingress accident. The goal of the research is to determine the time consumed by HPS to remove the total amount of water ingress. The method used is modelling and simulation of the HPS by using ChemCAD software. The BDBA and DBA scenarios will be simulated. In BDBA scenario, up to 110 kg of water is assumed to infiltrate to primary coolant while DBA is up to 35 kg. By using ChemCAD simulation, the second train will purify steam ingress maximum in 0.5 hours. The HPS of RDE has a capability to anticipate the water ingress accident.

Experimental study on air ingress during a primary pipe rupture accident with a graphite reactor core simulator

International Nuclear Information System (INIS)

Takeda, Tetsuaki; Hishida, Makoto; Baba, Shinichi

1991-11-01

When a primary coolant pipe of a High Temperature Gas Cooled Reactor (HTGR) ruptures, helium gas in the reactor core blows out into the container, and the primary cooling system reduces the pressure. After the pressures are balanced between the reactor and the container, air is expected to enter into the reactor core from the breach. It seems to be probable that the graphite structures is oxidized by air. Hence, it is necessary to investigate the air ingress process and the behavior of the generating gases by the oxidation reactions. The previous experimental study is performed on the molecular diffusion and natural convection of the two component gas mixtures using a test model simulating simply the reactor. Objective of the study was to investigate the air ingress process during the early stage of the primary pipe rupture accident. However, since the model did not have any kind of graphite components, the reaction between graphite and oxygen was not simulated. The present model includes the reactor core and the high temperature plenum simulators made of graphite. The major results obtained in the present study are summarized in the followings: (1) The air ingress process with graphite oxidation reaction is similar to that without the reaction qualitatively. (2) When the reactor core simulator is maintained at low temperatures (lower than 450degC), the initiation time of the natural circulation of air is almost equal to that of the natural circulation of nitrogen. On the other hand, when the temperature of the reactor core simulator is high (more than 500degC), the initiation time of the natural circulation of air is earlier than that of nitrogen. (3) When the temperature of the reactor core simulator is higher than 600degC, oxygen is almost dissipated by the graphite structures. When the temperature of the reactor core simulator is below 700degC, carbon dioxide mainly is generated by the oxidation reactions. (author)

Air ingress and graphite burning in HTRs: A survey on analytical examinations performed with the code React/Thermix

International Nuclear Information System (INIS)

Moormann, R.

1995-05-01

Analyses on air ingress in the pebble bed reactors PNP-500, THTR-300, HTR-Modul, AVR-II and AVR process heat plant are outlined; in addition, some results for the VHTR with block type fuel are given. Air ingress requires primary circuit depressurization and large leak(s) to reactor buildings and environment and belongs therefore to highly hypothetical events in the sense of classical safety analysis. One accident class examined is air ingress with forced flow by emergency cooling: For this case, the range of mass flow/air content in cooling gas has been evaluated, in which safe core cool down is possible resp. long term core burning occurs; for highest available emergency cooling flow, a safe cool down of the THTR-300, which has no reactor building, is possible for up to 20 vol-% of air in the cooling gas, wheras low flow allows only for about 5 vol-%. If the amount of available air is restricted to the content of a reactor building, as is examined for the PNP-500, relevant consequences have not to be expected; this remains also true for forced convection flow, if burning of CO, formed by graphite oxidation, within the building is considered. For the second accident class examined, air ingress with natural convection flow by chimney draught as studied for the HTR-Modul and some other concepts, the time span until significant fission product release begins has been determined; in case, that the bottom reflector is hot at accident start (> 600 C) and therefore consumes most of the ingressing oxygen, this time span is at least several hours and leak tightening counter measures may be possible. (orig./HP)

Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

Energy Technology Data Exchange (ETDEWEB)

Chang Ho Oh; Eung Soo Kim; Hee Cheon No; Nam Zin Cho

2008-12-01

The US Department of Energy is performing research and development (R&D) that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP) Program / GEN-IV Very High Temperature Reactor (VHTR). Phenomena identification and ranking studies (PIRT) to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Schultz et al., 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) are very high priority for the NGNP program. Following a loss of coolant and system depressurization, air will enter the core through the break. Air ingress leads to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heat-up of the bottom reflector and the reactor core and will cause the release of fission products eventually. The potential collapse of the bottom reflector because of burn-off and the release of CO lead to serious safety problems. For estimation of the proper safety margin we need experimental data and tools, including accurate multi-dimensional thermal-hydraulic and reactor physics models, a burn-off model, and a fracture model. We also need to develop effective strategies to mitigate the effects of oxidation. The results from this research will provide crucial inputs to the INL NGNP/VHTR Methods R&D project. This project is focused on (a) analytical and experimental study of air ingress caused by density-driven, stratified, countercurrent flow, (b) advanced graphite oxidation experiments, (c) experimental study of burn-off in the bottom reflector, (d) structural tests of the burnt-off bottom reflector, (e) implementation of advanced models developed during the previous tasks into the GAMMA code, (f) full air ingress and oxidation mitigation analyses, (g) development of core neutronic models, (h) coupling of the core neutronic and thermal hydraulic models, and (i

ISOTHERMAL AIR INGRESS VALIDATION EXPERIMENTS AT IDAHO NATIONAL LABORATORY: DESCRIPTION AND SUMMARY OF DATA

International Nuclear Information System (INIS)

Oh, Chang H.; Kim, Eung S.

2010-01-01

Idaho National Laboratory performed air ingress experiments as part of validating computational fluid dynamics code (CFD). An isothermal stratified flow experiment was designed and set to understand stratified flow phenomena in the very high temperature gas cooled reactor (VHTR) and to provide experimental data for validating computer codes. The isothermal experiment focused on three flow characteristics unique in the VHTR air-ingress accident: stratified flow in the horizontal pipe, stratified flow expansion at the pipe and vessel junction, and stratified flow around supporting structures. Brine and sucrose were used as heavy fluids and water was used as light fluids. The density ratios were changed between 0.87 and 0.98. This experiment clearly showed that a stratified flow between heavy and light fluids is generated even for very small density differences. The code was validated by conducting blind CFD simulations and comparing the results to the experimental data. A grid sensitivity study was also performed based on the Richardson extrapolation and the grid convergence index method for modeling confidence. As a result, the calculated current speed showed very good agreement with the experimental data, indicating that the current CFD methods are suitable for predicting density gradient stratified flow phenomena in the air-ingress accident.

Analysis of diffusion process and influence factors in the air ingress accident of the HTR-PM

Energy Technology Data Exchange (ETDEWEB)

Yanhua, Zheng, E-mail: zhengyh@mail.tsinghua.edu.cn; Fubing, Chen; Lei, Shi

2014-05-01

Air ingress, one of the beyond design basis accidents for high temperature gas-cooled reactors, receives high attention during the design of the 250 MW pebble-bed modular high temperature gas-cooled reactor (HTR-PM), because it may result in severe consequence including the corrosion of the fuel element and graphite reflector. The diffusion process and the set-up time of the stable natural convection after the double-ended guillotine break of the hot-gas duct are studied in the paper. On the basis of the preliminary design of the HTR-PM and its DLOCA analysis results, the diffusion process, as well as the influence of the core temperature distribution and the length of the hot-gas duct, is studied with the DIFFLOW code, which adopts a one-dimension variable cross-section diffusion model with fixed wall temperature. To preliminarily estimate the influence of chemical reaction between oxygen and graphite, which will change the gas component of the mixture, the diffusion processes between the He/N{sub 2}, He/O{sub 2}, He/CO and He/CO{sub 2} are calculated, respectively. Furthermore, the code has been improved and the varying wall temperature can be simulated. The more accurate analysis is carried out with the changing temperature distribution from the DLOCA calculation. The analysis shows that there is enough time to adopt appropriate mitigation measures to stop the air ingress and the severe consequence of fuel element damage and large release of fission product can be avoided.

The effects of applying silicon carbide coating on core reactivity of pebble-bed HTR in water ingress accident

Energy Technology Data Exchange (ETDEWEB)

Zuhair, S.; Setiadipura, Topan [National Nuclear Energy Agency of Indonesia, Serpong Tagerang Selatan (Indonesia). Center for Nuclear Reactor Technology and Safety; Su' ud, Zaki [Bandung Institute of Technology (Indonesia). Dept. of Physics

2017-03-15

Graphite is used as the moderator, fuel barrier material, and core structure in High Temperature Reactors (HTRs). However, despite its good thermal and mechanical properties below the radiation and high temperatures, it cannot avoid corrosion as a consequence of an accident of water/air ingress. Degradation of graphite as a main HTR material and the formation of dangerous CO gas is a serious problem in HTR safety. One of the several steps that can be adopted to avoid or prevent the corrosion of graphite by the water/air ingress is the application of a thin layer of silicon carbide (SiC) on the surface of the fuel element. This study investigates the effect of applying SiC coating on the fuel surfaces of pebble-bed HTR in water ingress accident from the reactivity points of view. A series of reactivity calculations were done with the Monte Carlo transport code MCNPX and continuous energy nuclear data library ENDF/B-VII at temperature of 1200 K. Three options of UO{sub 2}, PuO{sub 2}, and ThO{sub 2}/UO{sub 2} fuel kernel were considered to obtain the inter comparison of the core reactivity of pebble-bed HTR in conditions of water/air ingress accident. The calculation results indicated that the UO{sub 2}-fueled pebble-bed HTR reactivity was slightly reduced and relatively more decreased when the thickness of the SiC coating increased. The reactivity characteristic of ThO{sub 2}/UO{sub 2}-fueled pebble-bed HTR showed a similar trend to that of UO{sub 2}, but did not show reactivity peak caused by water ingress. In contrast with UO{sub 2}- and ThO{sub 2}-fueled pebble-bed HTR, although the reactivity of PuO{sub 2}-fueled pebble-bed HTR was the lowest, its characteristics showed a very high reactivity peak (0.33 Δk/k) and this introduction of positive reactivity is difficult to control. SiC coating on the surface of the plutonium fuel pebble has no significant impact. From the comparison between reactivity characteristics of uranium, thorium and plutonium cores with 0

SARNET2 benchmark on air ingress experiments QUENCH-10, -16

International Nuclear Information System (INIS)

Fernandez-Moguel, Leticia; Bals, Christine; Beuzet, Emilie; Bratfisch, Christian; Coindreau, Olivia; Hózer, Zoltan; Stuckert, Juri; Vasiliev, Alexander; Vryashkova, Petya

2014-01-01

Highlights: • Two similar QUENCH air ingress experiments were analysed with eight different codes. • Eight institutions have participated in the study. • Differences in the code were mostly small to moderate during the pre-oxidation. • Differences in the code were larger during the air phase. • Study has proven that there are physical processes that should be further studied. - Abstract: The QUENCH-10 (Q-10) and QUENCH-16 (Q-16) experiments were chosen as a SARNET2 code benchmark (SARNET2-COOL-D5.4) exercise to assess the status of modelling air ingress sequences and to compare the capabilities of the various codes used for accident analyses, specifically ATHLET-CD (GRS and RUB), ICARE-CATHARE (IRSN), MAAP (EDF), MELCOR (INRNE and PSI), SOCRAT (IBRAE), and RELAP/SCDAPSim (PSI). Both experiments addressed air ingress into an overheated core following earlier partial oxidation in steam. Q-10 was performed with extensive preoxidation, moderate/high air flow rate and high temperatures at onset of reflood (max T pct = 2200 K), while Q-16 was performed with limited preoxidation, low air flow rate and relative low temperatures at reflood initiation (max T pct = 1870 K). Variables relating to the major signatures (thermal response, hydrogen generation, oxide layer development, oxygen and nitrogen consumption and reflood behaviour) were compared globally and/or at selected locations. In each simulation, the same input models and assumptions are used for both experiments, differing only in respect of the boundary conditions. However, some slight idealisations were made to the assumed boundary conditions in order to avoid ambiguities in the code-to-code comparisons; in this way, it was possible to focus more easily on the key phenomena and hence make the results of the exercise more transparent. Remarks are made concerning the capability of physical modelling within the codes, description of the experiment facility and test conduct as specified in the code input

Separate-effect tests on zirconium cladding degradation in air ingress situations

Energy Technology Data Exchange (ETDEWEB)

Duriez, C. [Institut de Radioprotection et de Surete Nucleaire, IRSN, Direction de Prevention des Accidents Majeurs, Centre de Cadarache, 13115 St Paul Lez Durance (France)], E-mail: christian.duriez@irsn.fr; Steinbrueck, M. [Forschungszentrum Karlsruhe, FZK, Institut fuer Materialforschung, Postfach 3640, 76021 Karlsruhe (Germany); Ohai, D.; Meleg, T. [Institute for Nuclear Research, INR, Nuclear Material and Corrosion Department, Pitesti, 115400 Mioveni Arges (Romania); Birchley, J.; Haste, T. [Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)

2009-02-15

In the event of air ingress during a reactor or spent fuel pond low probability accident, the fuel rods will be exposed to air-containing atmospheres at high temperatures. In comparison with steam, the presence of air is expected to result in a more rapid escalation of the accident. A state-of-the-art review performed before SARNET started showed that the existing data on zirconium alloy oxidation in air were scarce. Moreover, the exact role of zirconium nitride on the cladding degradation process was poorly understood. Regarding the cladding behaviour in air + steam or nitrogen-enriched atmospheres (encountered in oxygen-starved conditions), almost no data were available. New experimental programmes comprising small-scale tests have therefore been launched at FZK, IRSN (MOZART programme in the frame of the International Source Term Program-ISTP) and INR. Zircaloy-4 cladding in PWR (FZK, IRSN) and in CANDU (INR) geometry are investigated. On-line kinetic data are obtained on centimetre size tube segments, by thermogravimetry (FZK, IRSN and INR) or by mass spectrometry (FZK). Plugged tubes 15 cm long (FZK) are also investigated. The samples are air-oxidised either in the 'as-received' state, or after pre-oxidation in steam. 'Analytical' tests at constant temperature and gas composition provide basic kinetic data, while more prototypical temperature transients and sequential gas compositions are also investigated. The temperature domains extend from 600 deg. C up to 1500 deg. C. Systematic post-test metallographic inspections are performed. The paper gives a synthesis of the results obtained, comparing them in terms of kinetics and oxide scale structure and composition. A comparative analysis is performed with results of the QUENCH-10 (Q-10) bundle test, which included an air ingress phase. It is shown how the data contribute to a better understanding of the cladding degradation process, especially regarding the role of nitrogen. For modelling of

Separate-effect tests on zirconium cladding degradation in air ingress situations

International Nuclear Information System (INIS)

Duriez, C.; Steinbrueck, M.; Ohai, D.; Meleg, T.; Birchley, J.; Haste, T.

2009-01-01

In the event of air ingress during a reactor or spent fuel pond low probability accident, the fuel rods will be exposed to air-containing atmospheres at high temperatures. In comparison with steam, the presence of air is expected to result in a more rapid escalation of the accident. A state-of-the-art review performed before SARNET started showed that the existing data on zirconium alloy oxidation in air were scarce. Moreover, the exact role of zirconium nitride on the cladding degradation process was poorly understood. Regarding the cladding behaviour in air + steam or nitrogen-enriched atmospheres (encountered in oxygen-starved conditions), almost no data were available. New experimental programmes comprising small-scale tests have therefore been launched at FZK, IRSN (MOZART programme in the frame of the International Source Term Program-ISTP) and INR. Zircaloy-4 cladding in PWR (FZK, IRSN) and in CANDU (INR) geometry are investigated. On-line kinetic data are obtained on centimetre size tube segments, by thermogravimetry (FZK, IRSN and INR) or by mass spectrometry (FZK). Plugged tubes 15 cm long (FZK) are also investigated. The samples are air-oxidised either in the 'as-received' state, or after pre-oxidation in steam. 'Analytical' tests at constant temperature and gas composition provide basic kinetic data, while more prototypical temperature transients and sequential gas compositions are also investigated. The temperature domains extend from 600 deg. C up to 1500 deg. C. Systematic post-test metallographic inspections are performed. The paper gives a synthesis of the results obtained, comparing them in terms of kinetics and oxide scale structure and composition. A comparative analysis is performed with results of the QUENCH-10 (Q-10) bundle test, which included an air ingress phase. It is shown how the data contribute to a better understanding of the cladding degradation process, especially regarding the role of nitrogen. For modelling of the oxide scale

Analytical support for the preparation of bundle test QUENCH-10 on air ingress

International Nuclear Information System (INIS)

Birchley, J.; Haste, T.; Homann, C.; Hering, W.

2005-07-01

Bundle test QUENCH-10 is dedicated to study air ingress with subsequent water quench during a supposed accident in a spent fuel storage tank. It was proposed by AEKI, Budapest, Hungary and was performed on 21 July 2004 in the QUENCH facility at Forschungszentrum Karlsruhe. Preparation of the test is based on common analytical work at Forschungszentrum Karlsruhe and Paul Scherrer Institut, Villigen, Switzerland, mainly with the severe accident codes SCDAP/RELAP5 and MELCOR, to derive the protocol for the essential test phases, namely pre-oxidation, air ingress and quench phase. For issues that could not be tackled by this computational work, suggestions for the test conduct were made and applied during the test. Improvements of the experimental set-up and the test conduct were suggested and largely applied. In SCDAP/RELAP5, an error was found: for thick oxide scales, the output value of the oxide scale is sensibly underestimated. For the aims of the test preparation, its consequences could be taken into account. Together with the related computational and other analytical support by the engaged institutions the test is co-financed as test QUENCH-L1 by the European Community under the Euratom Fifth Framework Programme on Nuclear Fission Safety 1998 - 2002 (LACOMERA Project, contract No. FIR1-CT2002-40158). (orig.)

Oxidation kinetics of innovative carbon materials with respect to severe air ingress accidents in HTRs and graphite disposal or processing; Oxidationskinetik innovativer Kohlenstoffmaterialien hinsichtlich schwerer Lufteinbruchstoerfaelle in HTR's und Graphitentsorgung oder Aufbereitung

Energy Technology Data Exchange (ETDEWEB)

Schloegel, Baerbel

2010-07-01

Currently future nuclear reactor concepts of the Fourth Generation (Gen IV) are under development. To some extend they apply with new, innovative materials developed just for this purpose. This thesis work aims at a concept of Generation IV Very High Temperature Reactors (VHTR) in the framework of the European project RAPHAEL (ReActor for Process heat, Hydrogen And ELectricity generation). The concept named ANTARES (AREVA New Technology based on advanced gas-cooled Reactors for Energy Supply) was developed by AEVA NP. It is a helium cooled, graphite moderated modular reactor for electricity and hydrogen production, by providing the necessary process heat due to its high working temperature. Particular attention is given here to oxidation kinetics of newly developed carbon materials (NBG-17) with still unknown but needed information in context of severe air ingress accident in VHTR's. Special interest is paid to the Boudouard reaction, the oxidation of carbon by CO{sub 2}. In case of an air ingress accident, carbon dioxide is produced in the primary reaction of atmospheric oxygen with reflector graphite. From there CO{sub 2} could flow into the reactor core causing further damage by conversion into CO. The purpose of this thesis is to ascertain if and to what degree this could happen. First of all oxidation kinetic data of the Boudouard reaction with NBG-17 is determined by experiments in a thermo gravimetric facility. The measurements are evaluated and converted into a common formula and a Langmuir-Hinshelwood similar oxidation kinetic equation, as input for the computer code REACT/THERMIX. This code is then applied to analyse severe air ingress accidents for several air flow rates. The results are discussed for two accident situations, in which a certain graphite burn off is achieved. All cases show much more damage to the graphite bottom reflector than to the reactor core. Thus the bottom reflector will lose its structural integrity much earlier than the

Discussion on amount of water ingress mass in steam generator heat-exchange tube rupture accident of high- temperature gas-cooled reactor

International Nuclear Information System (INIS)

Wang Yan; Zheng Yanhua; Shi Lei; Li Fu; Sun Ximing

2009-01-01

The steam generator heat-exchange tube rupture (SGTR) accident which will result in the water ingress to the primary circuit of reactor is an important and particular accident for high-temperature gas-cooled reactor (HTGR). The analysis of the water ingress accident is significant for verifying the inherent safety characteristics of HTGR. The amount of water ingress mass is one of the decisive factors for the seriousness of the accident consequence. The 250 MW Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) designed by Institute of Nuclear and New Energy Technology of Tsinghua University was selected as an example of analysis. The analysis results show that the amount of water ingress mass is not only affected directly with the broken position and the broken area of the tubes, but also related with the diameter of draining piping and restrictor, draining control valve, action setting of emptier system. With reasonable parameters chosen, the water in steam generator could be drained effectively, so it will prevent the primary circuit of reactor from water ingress in large quantity and reduce the radioactive isotopes ingress to the secondary circuit. (authors)

The study on water ingress mass in the steam generator heat-exchange tube rupture accident of modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Wang Yan; Shi Lei; Li Fu; Zheng Yanhua

2012-01-01

The steam generator heat-exchange tube rupture (SGTR) accident is an important and particular accident which will result in water ingress to the primary loop of reactor. Water ingress will result in chemical reaction of graphite fuel and structure with water, which may cause overpressure due to generation of explosive gaseous in large quantity. The study on the water ingress accident is significant for the verification of the inherent characteristics of high temperature gas-cooled reactor. The previous research shows that the amount of water ingress mass is the dominant key factor on the severity of the accident consequence. The 200 MWe high temperature gas-cooled reactor (HTR-PM), which is the first modular pebble-bed high temperature gas-cooled reactor in China designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is selected to be analyzed in this paper. The different DBA accident scenarios of double-ended break of single heat-exchange tube are simulated respectively by the thermal-hydraulic analysis code RETRAN-02. The results show the water ingress mass through the broken heat-exchange tube is related to the break location. The amount of water ingress mass is affected obviously by the capacity of the emptier system. With the balance of safety and economical efficiency, the amount of water ingress mass from the secondary side of steam generator into the primary coolant loop will be reduced by increasing properly the diameter of the draining lines. (authors)

AIR INGRESS ANALYSIS: COMPUTATIONAL FLUID DYNAMIC MODELS

Energy Technology Data Exchange (ETDEWEB)

Chang H. Oh; Eung S. Kim; Richard Schultz; Hans Gougar; David Petti; Hyung S. Kang

2010-08-01

The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is performing research and development that focuses on key phenomena important during potential scenarios that may occur in very high temperature reactors (VHTRs). Phenomena Identification and Ranking Studies to date have ranked an air ingress event, following on the heels of a VHTR depressurization, as important with regard to core safety. Consequently, the development of advanced air ingress-related models and verification and validation data are a very high priority. Following a loss of coolant and system depressurization incident, air will enter the core of the High Temperature Gas Cooled Reactor through the break, possibly causing oxidation of the in-the core and reflector graphite structure. Simple core and plant models indicate that, under certain circumstances, the oxidation may proceed at an elevated rate with additional heat generated from the oxidation reaction itself. Under postulated conditions of fluid flow and temperature, excessive degradation of the lower plenum graphite can lead to a loss of structural support. Excessive oxidation of core graphite can also lead to the release of fission products into the confinement, which could be detrimental to a reactor safety. Computational fluid dynamic model developed in this study will improve our understanding of this phenomenon. This paper presents two-dimensional and three-dimensional CFD results for the quantitative assessment of the air ingress phenomena. A portion of results of the density-driven stratified flow in the inlet pipe will be compared with results of the experimental results.

Investigations on accidents with massive water ingress exemplified by the pebble bed reactor PNP-500

International Nuclear Information System (INIS)

Moormann, R.

1986-01-01

A computer code is used for analyses of massive water ingress accidents in the High-Temperature Gas Cooled Reactor concept PNP-500 with pebble bed core. The analyses are mainly focussed on graphite corrosion processes. For the investigated accidents a correct reactor shut down in assumed. The mass of water ingressing into the primary circuit is varied between 1000 and 7500 kg (i.e., up to hypothetical values). The dependence of accident consequences on parameters such as intensity and starting time of the afterheat removal system or kinetic values of the chemical processes is examined. The results show that even under pessimistic assumptions the extent of the graphite corrosion is relatively low; significant damaging of fuel elements or graphite components does not occur. A primary circuit depressurization, combined with local burning of water gas, would probably not affect the fission product retention potential of the (gastight) containment. Summing up, the risk caused by these accidents remains small. (orig.) [de

Analysis on blow-down transient in water ingress accident of high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Wang, Yan; Zheng, Yanhua; Li, Fu; Shi, Lei

2014-01-01

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor, which will cause a positive reactivity introduction with the increase of steam density in reactor core to enhance neutron slowing-down, also the chemical corrosion of graphite fuel elements and the damage of reflector structure material. The increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The research on water ingress transient is significant for the verification of inherent safety characteristics of high temperature gas-cooled reactor. The 200 MWe high temperature gas-cooled reactor (HTR-PM), designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is exampled to be analyzed in this paper. The design basis accident (DBA) scenarios of double-ended guillotine break of single heat-exchange tube (steam generator heat-exchange tube rupture) are simulated by the thermal-hydraulic analysis code, and some key concerns which are relative to the amount of water into the reactor core during the blow-down transient are analyzed in detail. The results show that both of water mass and steam ratio of the fluid spouting from the broken heat-exchange tube are affected by break location, which will increase obviously with the broken location closing to the outlet of the heat-exchange tube. The double-ended guillotine rupture at the outlet of the heat-exchange will result more steam penetrates into the reactor core in the design basis accident of water ingress. The mass of water ingress will also be affected by the draining system. It is concluded that, with reasonable optimization on design to balance safety and economy, the total mass of water ingress into the primary circuit of reactor could be limited effectively to meet the safety requirements, and the pollution of

ANALISIS EFEK KECELAKAAN WATER INGRESS TERHADAP REAKTIVITAS DOPPLER TERAS RGTT200K

Directory of Open Access Journals (Sweden)

Zuhair Zuhair

2015-03-01

Full Text Available Dalam high temperature reactor, koefisien reaktivitas temperatur yang didesain negatif menjamin reaksi fisi dalam teras tetap berada di bawah kendali dan panas peluruhan tidak akan pernah melelehkan bahan bakar yang menyebabkan terlepasnya zat radioaktif ke lingkungan. Namun masuknya air (water ingress ke dalam teras reaktor akibat pecahnya tabung penukar panas generator uap, yang dikenal sebagai salah satu kecelakaan dasar desain, dapat mengintroduksi reaktivitas positif dengan potensi bahaya lainnya seperti korosi grafit dan kerusakan material struktur reflektor. Makalah ini akan menganalisis efek kecelakaan water ingress terhadap reaktivitas Doppler teras RGTT200K. Kapabilitas koefisien reaktivitas Doppler untuk mengkompensasi reaktivitas positif yang timbul selama kecelakaan water ingress akan diuji melalui serangkaian perhitungan dengan program MCNPX dan pustaka ENDF/B-VII untuk perubahan temperatur bahan bakar dari 800K hingga 1800K. Tiga opsi kernel bahan bakar UO2, ThO2/UO2 dan PuO2 dengan tiga model kisi bahan bakar pebble di teras reaktor diterapkan untuk kondisi water ingress dengan densitas air dari 0 hingga 1.000 kg/m3. Hasil perhitungan memperlihatkan koefisien reaktivitas Doppler tetap negatif untuk seluruh opsi bahan bakar yang dipertimbangkan bahkan untuk posibilitas water ingress yang besar. Efek water ingress lebih kuat pada model kisi dengan fraksi packing lebih rendah karena lebih banyak volume yang tersedia untuk air yang memasuki teras reaktor. Efek water ingress juga lebih kuat di teras uranium dibandingkan teras thorium dan plutonium sebagai konsekuensi dari fenomena Doppler dimana absorpsi neutron di daerah resonansi 238U lebih besar daripada 232Th dan 240Pu. Secara keseluruhan dapat disimpulkan bahwa, koefisien Doppler teras RGTT200K mampu mengkompensasi insersi reaktivitas yang diintroduksi oleh kecelakaan water ingress. Teras RGTT200K dengan bahan bakar UO2, ThO2/UO2 dan PuO2 dapat mempertahankan fitur keselamatan

Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Dong Yujie; Scherer, Winfried

2005-01-01

Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was 3 s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to ∼3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged

Separate-effects experiments on the hydrodynamics of air ingress phenomena for the very high temperature reactor

International Nuclear Information System (INIS)

Kim, S.; Talley, J.; Yadav, M.; Ireland, A.; Bajorek, S.

2011-01-01

The present study performs scaled separate-effects experiments to investigate the hydrodynamics in the air-ingress phenomena following a Depressurized Condition Cooldown in the Very High Temperature Gas-Cooled Reactor. First, a scoping experiment using water and brine is performed. The volumetric exchange rate is measured using a hydrometer, and flow visualizations are performed. Next, Helium-air experiments are performed to obtain three-dimensional oxygen concentration transient data using an oxygen analyzer. It is found that there exists a critical density difference ratio, before which the ingress rate increases linearly with time and after which the ingress rate slows down significantly. In both the water-brine and Helium-air experiments, this critical ratio is found to be approximately 0.7. (author)

Separate-effects experiments on the hydrodynamics of air ingress phenomena for the very high temperature reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, S.; Talley, J.; Yadav, M., E-mail: skim@psu.edu [The Pennsylvania State Univ., University Park, Pennsylvania (United States); Ireland, A.; Bajorek, S. [The United States Nuclear Regulatory Commission, Washington DC (United States)

2011-07-01

The present study performs scaled separate-effects experiments to investigate the hydrodynamics in the air-ingress phenomena following a Depressurized Condition Cooldown in the Very High Temperature Gas-Cooled Reactor. First, a scoping experiment using water and brine is performed. The volumetric exchange rate is measured using a hydrometer, and flow visualizations are performed. Next, Helium-air experiments are performed to obtain three-dimensional oxygen concentration transient data using an oxygen analyzer. It is found that there exists a critical density difference ratio, before which the ingress rate increases linearly with time and after which the ingress rate slows down significantly. In both the water-brine and Helium-air experiments, this critical ratio is found to be approximately 0.7. (author)

Oxidation parameters of nuclear graphite for HTGR air-ingress

International Nuclear Information System (INIS)

Kim, E.S.; No, H.C.

2004-01-01

In order to investigate chemical behaviors of the graphite during an air-ingress accident in HTGR, the kinetic tests on nuclear graphite IG-110 were performed in chemical reaction dominant regime. In the present experiment, inlet gas flow rate ranged between 8 and 18 SLPM, graphite temperatures and oxygen mole fraction ranged from 540 to 630degC and from 3 to 30% respectively. The test section was made of a quartz tube having 75 mm diameter and 750 mm length and the test specimen machined to the size of 21 mm diameter and 30 mm length was supported at the center of it by the alumina rod. The 15 kW induction heater was installed around the outside of test section to heat the specimen and its temperature was measured by 2 infrared thermometers. The oxidation rate was calculated from the gas concentration analysis between inlet and outlet using NDIR (non-dispersive infrared) gas analyzer. As a result the activation energy (Ea) and the order of reaction (n) were determined within 95% confidence level and the qualitative characteristics of the two parameters were also widely investigated by experimental and analytical methods. (author)

Severe water ingress accident analysis for a Modular High Temperature Gas Cooled Reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Scherer, Winfried

1997-01-01

This paper analyzes the severe water ingress accidents in the SIEMENS 200MW Modular High Temperature Gas Cooled Reactor (HTR-Module) under the assumption of no active safety protection systems in order to find the safety margin of the current HTR-Module design. A water, steam and helium multi-phase cavity model is originally developed and implemented in the DSNP simulation system. The developed DSNP system is used to simulate the primary circuit of HTR-Module power plant. The comparisons of the models with the TINTE calculations validate the current simulation. After analyzing the effects of blower separation on water droplets, the wall heat storage, etc., it is found that the maximum H 2 O density increase rate in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduces the impulse of the H 2 O in the reactor core. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600degC was not reached in any case. (author)

Experimental and analytical investigations to air and steam ingress into the vacuum vessel of fusion reactors

International Nuclear Information System (INIS)

Kruessenberg, A.K.

1996-12-01

The basic fusion safety objective is the development of fusion power plants with features that protect individuals, society and the environment by establishing and maintaining an effective defence against radiological and other hazards. The most important specific principle is the establishment of three sequential levels of defence, characterized in priority order by prevention, protection and mitigation. The safety conscious selection of materials as one prevention feature gives the basis for the work described in this report. In order to protect the metallic first wall of fusion reactors from direct interaction with the plasma an extra armour is foreseen. Carbon offers the features low atomic number, high melting point, high thermal conductivity and good mechanical stability up to high temperatures making it to a favourite armour material. Looking on the safety behaviour of fusion reactors it has to be noted that carbon is unstable against oxidizing media like oxygen and steam at high temperatures und carbon has a high sorption capacity for radiologically important tritium. And tritium used as intermediate fuel in the actual reactor concepts is the one form radioactivity is present in fusion reactors. Accidents like loss of vacuum (LOVA) will lead to an air ingress into the vacuum vessel, oxidation of the hot carbon and a partial mobilization of the sorbed tritium. In a similar manner loss of coolant into vacuum (LOCIV) will lead to a water/steam ingress into the vacuum vessel, also accompanied by carbon oxidation and tritium release. (orig.)

Autonomous Landing and Ingress of Micro-Air-Vehicles in Urban Environments Based on Monocular Vision

Science.gov (United States)

Brockers, Roland; Bouffard, Patrick; Ma, Jeremy; Matthies, Larry; Tomlin, Claire

2011-01-01

Unmanned micro air vehicles (MAVs) will play an important role in future reconnaissance and search and rescue applications. In order to conduct persistent surveillance and to conserve energy, MAVs need the ability to land, and they need the ability to enter (ingress) buildings and other structures to conduct reconnaissance. To be safe and practical under a wide range of environmental conditions, landing and ingress maneuvers must be autonomous, using real-time, onboard sensor feedback. To address these key behaviors, we present a novel method for vision-based autonomous MAV landing and ingress using a single camera for two urban scenarios: landing on an elevated surface, representative of a rooftop, and ingress through a rectangular opening, representative of a door or window. Real-world scenarios will not include special navigation markers, so we rely on tracking arbitrary scene features; however, we do currently exploit planarity of the scene. Our vision system uses a planar homography decomposition to detect navigation targets and to produce approach waypoints as inputs to the vehicle control algorithm. Scene perception, planning, and control run onboard in real-time; at present we obtain aircraft position knowledge from an external motion capture system, but we expect to replace this in the near future with a fully self-contained, onboard, vision-aided state estimation algorithm. We demonstrate autonomous vision-based landing and ingress target detection with two different quadrotor MAV platforms. To our knowledge, this is the first demonstration of onboard, vision-based autonomous landing and ingress algorithms that do not use special purpose scene markers to identify the destination.

Improvement design study on steam generator of MHR-50/100 aiming higher safety level after water ingress accident

International Nuclear Information System (INIS)

Oyama, S.; Minatsuki, I.; Shimizu, K.

2012-01-01

Mitsubishi Heavy Industries, Ltd. (MHI) has been studying on MHI original High Temperature Gas cooled Reactor (HTGR), namely MHR-50/100, for commercialization with supported by JAEA. In the heat transfer system, steam generator (SG) is one of the most important components because it should be imposed a function of heat transfer from reactor power to steam turbine system and maintaining a nuclear grade boundary. Then we especially focused an effort of a design study on the SG having robustness against water ingress accident based on our design experience of PWR, FBR and HTGR. In this study, we carried out a sensitivity analysis from the view point of economic and plant efficiency. As a result, the SG design parameter of helium inlet/outlet temperature of 750 deg. C/300 deg. C, a side-by-side layout and one unit of SG attached to a reactor were selected. In the next, a design improvement of SG was carried out from the view point of securing the level of inherent safety without reliance on active steam dump system during water ingress accident considering the situation of the Fukushima nuclear power plant disaster on March 11, 2011. Finally, according to above basic design requirement to SG, we performed a conceptual design on adapting themes of SG structure improvement. (authors)

Experiments on the behaviour of ruthenium in air ingress accidents

International Nuclear Information System (INIS)

Kaerkelae, T.; Backman, Ul; Auvinen, A.; Zilliacus, R.; Lipponen, M.; Kekki, T.; Tapper, U.; Jokiniemi, J.

2007-03-01

During routine nuclear reactor operation, ruthenium will accumulate in the fuel in relatively high concentrations. In a severe accident in a nuclear power plant it is possible that air gets into contact with the reactor core. In this case ruthenium may oxidise and form volatile ruthenium species, RuO3 and RuO4, which can be transported into the containment. In order to estimate the amount of gaseous ruthenium species, it is of interest to know, how they are formed and how they behave. In our experiments the formation and transport of volatile ruthenium oxides was studied by exposing RuO2 powder to diverse oxidising atmospheres at a relatively high temperature. Transport of gaseous RuO4 was further investigated by injecting it into the facility in similar conditions. Upon cooling of the gas flow RuO2 aerosol particles were formed in the system. They were removed from the gas stream with plane filters. Gaseous ruthenium species were trapped in 1M NaOH-water solution, which is capable of trapping RuO4 totally. Ruthenium in the solution was filtered for analysis. The determination of ruthenium both in aerosol and in liquid filters was made using instrumental neutron activation analysis (INAA). In order to close the mass balance and to achieve better time resolution seven experiment were carried out using radioactive tracer. In this report, the facility for the ruthenium behaviour study and results from experiments are presented. Preliminary conclusions from the experiments are reported as well. Final conclusions will be made after modelling of the facility is completed in a continuation work of this study. (au)

Experiments on the behaviour of ruthenium in air ingress accidents

Energy Technology Data Exchange (ETDEWEB)

Kaerkelae, T.; Backman, Ul; Auvinen, A.; Zilliacus, R.; Lipponen, M.; Kekki, T.; Tapper, U.; Jokiniemi, J. [Technical Research Centre of Finland VTT (Finland)

2007-03-15

During routine nuclear reactor operation, ruthenium will accumulate in the fuel in relatively high concentrations. In a severe accident in a nuclear power plant it is possible that air gets into contact with the reactor core. In this case ruthenium may oxidise and form volatile ruthenium species, RuO3 and RuO4, which can be transported into the containment. In order to estimate the amount of gaseous ruthenium species, it is of interest to know, how they are formed and how they behave. In our experiments the formation and transport of volatile ruthenium oxides was studied by exposing RuO2 powder to diverse oxidising atmospheres at a relatively high temperature. Transport of gaseous RuO4 was further investigated by injecting it into the facility in similar conditions. Upon cooling of the gas flow RuO2 aerosol particles were formed in the system. They were removed from the gas stream with plane filters. Gaseous ruthenium species were trapped in 1M NaOH-water solution, which is capable of trapping RuO4 totally. Ruthenium in the solution was filtered for analysis. The determination of ruthenium both in aerosol and in liquid filters was made using instrumental neutron activation analysis (INAA). In order to close the mass balance and to achieve better time resolution seven experiment were carried out using radioactive tracer. In this report, the facility for the ruthenium behaviour study and results from experiments are presented. Preliminary conclusions from the experiments are reported as well. Final conclusions will be made after modelling of the facility is completed in a continuation work of this study. (au)

Ingress of radioactive material into dwellings

International Nuclear Information System (INIS)

Roed, J.; Goddard, A.J.H.

1991-01-01

People inside buildings will normally obtain some protection against airborne pollution originating in the outdoor air. This protection will be due to deposition indoors on floors, walls, ceilings and furniture, and to filtration during ingress, in cracks, crevices and pores in the building structure. A building will also give some protection against direct gamma radiation from outdoor deposits on the ground and external building surfaces. However, indoor deposits, while reducing air exposure, may add to the dose commitment from direct gamma radiation. Other subsidiary factors such as mechanical transport may need to be considered. We used test houses for finding factors by which outdoor air concentrations exceed those found indoors. Be-7 of cosmogenic origin and associated with natural aerosol has been used as a tracer, together with radionuclides dispersed in the atmosphere as a result of the Chernobyl accident. In addition, SF 6 tracer measurements have been used to measure air exchange rates. Novel experimental arrangements, to isolate the deposition process, were used. Attention has also been given to developing suitable tracers for monodisperse artificial aerosols and to means of detecting such aerosols. The use of these techniques in studying mechanical transport of particulate material is described. The indoor to outdoor airborne concentration ratios in a test house during the Chernobyl accident were found to be 0.25 for caesium, 0.4 for particulate iodine and 0.5 for beryllium-7. Indoor deposition velocities, for a range of radionuclides, ranged from 7 10-5 to 2 10-4 m s-1. Subsequently, aerosol deposition measurements were conducted in unfurnished, partly furnished and fully furnished houses, using Be-7 as a tracer. Measured deposition velocities ranged from 10-5 m s-1 for the unfurnished room to 3 10-5 m s-1 for the fully furnished room

Physics of the pebble-bed high temperature reactor in massive water ingress accidents

International Nuclear Information System (INIS)

Scherer, W.

1989-10-01

A point-kinetics model was developed to describe qualitatively hypothetical water ingress transients in the primary loop of High Temperature Reactors. Neutron kinetics, heat-flow balance and the chemical reaction of graphite corrosion together with their mutual influence are included. The qualitative behaviour of the transients is calculated and discussed for two fictitious examples, namely the long-term water ingress into a medium sized HTR (HTR-500) and the 'startup' of a small HTR after an intensive water flooding of the core. The model developed and the computer code KINKOR are thought to be tools for the general understanding of the water ingress phenomena and should be looked at as basis for more elaborated systems. (orig./HP) [de

Numerical simulation of severe water ingress accidents in a modular high temperature gas cooled reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Scherer, W.

1996-01-01

This report analyzes reverse water ingress accidents in the SIEMENS 200 MW Modular Pebble-Bed High Temperature Gas Cooled Reactor (HTR-MODULE) under the assumption of no active safety protection systems in order to find the safety margins of the current HTR-MODULE design and to realize a catastrophe-free nuclear technology. A water, steam and helium multi-phase cavity model is developed and implemented in the DSNP simulation system. The DSNP system is then used to simulate the primary and secondary circuit of a HTR-MODULE power plant. Comparisons of the model with experiments and with TINTE calculations serve as validation of the simulation. The analysis of the primary circuit tries to answer the question how fast the water enters the reactor core. It was found that the maximum H 2 O concentration increase in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduce the ingress velocity of the H 2 O into the reactor core. In order to answer the question how much water enters the primary circuit, the full cavitation of the feed water pumps is analyzed. It is found that if the secondary circuit is depressurized enough, the feed water pumps will be inherently stopped by the full cavitation. This limits the water to be pumped from the deaerator to the steam generator. A comprehensive simulation of the MODUL-HTR power plant then shows that the H 2 O inventory in the primary circuit can be limited to about 3000 kg. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600 C was not reached in any case. (orig.) [de

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

International Nuclear Information System (INIS)

Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

1993-08-01

The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

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

International Nuclear Information System (INIS)

Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

1993-08-01

The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

Experiments on the behaviour of ruthenium in air ingress accidents - Progress report

International Nuclear Information System (INIS)

Kaerkelae, T.; Backman, U.; Auvinen, A.; Ziliacus, R.; Lipponen, M.; Kekki, T.; Tapper, U.; Jokiniemi, J.

2006-02-01

During routine nuclear reactor operation, ruthenium will accumulate in the fuel in relatively high concentrations. In an accident in a nuclear power plant it is possible that air gets into contact with the reactor core. In this case ruthenium can oxidise and form volatile ruthenium species, RuO3 and RuO4, which can be transported into the containment. In order to estimate the amount of gaseous ruthenium species it is of interest to know, how it is formed and how it behaves. In our experiments RuO2 is exposed to diverse oxidising atmospheres at a relatively high temperature. In this report, the experimental system for the ruthenium behaviour study is presented. Also preliminary results from experiments carried out during year 2005 are reported. In the experiments gaseous ruthenium oxides were produced in a furnace. Upon cooling RuO2 aerosol particles were formed in the system. They were removed with plane filters from the gas stream. Gaseous ruthenium species were trapped in 1M NaOH-water solution, which is capable of trapping RuO4 totally. Ruthenium in the solution was filtered for analysis. The determination of ruthenium both in aerosol and in liquid filters was made using instrumental neutron activation analysis (INAA). In order to close mass balance and achieve better time resolution three experiment using radioactive tracer were carried out. (au)

Experiments on the behaviour of ruthenium in air ingress accidents - Progress report

Energy Technology Data Exchange (ETDEWEB)

Kaerkelae, T.; Backman, U.; Auvinen, A.; Ziliacus, R.; Lipponen, M.; Kekki, T.; Tapper, U.; Jokiniemi, J. [VTT Technical Research Centre of Finland (Finland)

2006-02-15

During routine nuclear reactor operation, ruthenium will accumulate in the fuel in relatively high concentrations. In an accident in a nuclear power plant it is possible that air gets into contact with the reactor core. In this case ruthenium can oxidise and form volatile ruthenium species, RuO3 and RuO4, which can be transported into the containment. In order to estimate the amount of gaseous ruthenium species it is of interest to know, how it is formed and how it behaves. In our experiments RuO2 is exposed to diverse oxidising atmospheres at a relatively high temperature. In this report, the experimental system for the ruthenium behaviour study is presented. Also preliminary results from experiments carried out during year 2005 are reported. In the experiments gaseous ruthenium oxides were produced in a furnace. Upon cooling RuO2 aerosol particles were formed in the system. They were removed with plane filters from the gas stream. Gaseous ruthenium species were trapped in 1M NaOH-water solution, which is capable of trapping RuO4 totally. Ruthenium in the solution was filtered for analysis. The determination of ruthenium both in aerosol and in liquid filters was made using instrumental neutron activation analysis (INAA). In order to close mass balance and achieve better time resolution three experiment using radioactive tracer were carried out. (au)

On high-temperature reactor accident topology

International Nuclear Information System (INIS)

Fassbender, J.; Kroeger, W.; Wolters, J.

1981-01-01

American and German risk studies for an HTGR and independent investigations of hypothetical accident sequences led to a fundamental understanding of the topology of HTGR accident sequences. The dominating importance of core heat-up accidents was confirmed and the initiating events were identified. Complications of core heat-up accidents by air or water ingress are of minor importance for the risk, whereas the long-term development of accidents during days and weeks plays an important role for the environmental impact. The risk caused by an HTGR at a German site cannot yet be determined exactly, because no modern German HTGR design has passed a licensing procedure. Cautious estimates show that risk will appear to be substantially smaller than the LWR risk. The main reasons are the considerably reduced release of fission procucts and the slow development of core heat-up accidents leaving much time for measures which reduce the risk. (orig.) [de

Effect of air ingress on the energy performance of coal fired thermal power plants

International Nuclear Information System (INIS)

Siddhartha Bhatt, M.

2007-01-01

Ingress of air in boilers leads to drops in energy efficiency. This paper presents the effects of air ingress in the combustion zone, post-combustion zone and air pre-heater (APH) on the energy efficiency and loading capacity of a coal fired thermal power plant operating on fuel with high ash (35-45%). The optimal O 2 in the flue gas for a pulverized coal fired system is 3.5% (corresponding to 20% excess air). The operating values are in the range of 4.2-6.0% in membrane type boilers and up to 10% in refractory type boilers (after sustained periods of operation). The leakage rate of boilers (up to the entrance of the APH) is designed at 0.2% while the average operating values are 7.25% for membrane type enclosures and 33.61% for refractory enclosures. The leakage rate of the APH is designed at 5.0% while the operating values range from 13.66% to 20.13% for rotary and tubular APHs. When the O 2 in the combustion zone varies from 3.5% to 8.0%, efficiency drops of 2.0% points are experienced in the boiler and turbine separately, and the gross overall efficiency drop is ∼3.0% points. The units do not experience any capacity drop up to an O 2 in the flue gas of 6.0% before the APH. At an O 2 in the flue gas (before APH) of 7.2%, a mild limitation on the unit capacity of around 2-3% is experienced. When O 2 in the flue gas (before APH) reaches a level of 9.0%, 20% capacity drop of the unit is experienced due to which the plant load cannot be raised higher than 80%. Beyond the level of 9.0% (rare occurrence), the unit is quite difficult to operate and has to be taken off for overhaul

78 FR 77027 - Overhead Clearance (Air-Draft) Accidents

Science.gov (United States)

2013-12-20

... No. USCG-2013-0466] Overhead Clearance (Air-Draft) Accidents AGENCY: Coast Guard, DHS. ACTION... clearance (air-draft) accidents. In its petition, which calls for vessel masters to be provided with... accidents that it says were avoidable and that resulted in damage to or destruction of waterway...

Hypothetical air ingress scenarios in advanced modular high temperature gas cooled reactors

International Nuclear Information System (INIS)

Kroeger, P.G.

1988-01-01

Considering an extremely hypothetical scenario of complete cross duct failure and unlimited air supply into the reactor vessel of a modular high temperature gas cooled ractor, it is found that the potential air inflow remains limited due to the high friction pressure drop through the active core. All incoming air will be oxidized to CO and some local external burning would be temporarily possible in such a scenario. The accident would have to continue with unlimited air supply for hundreds of hours before the core structural integrity would be jeopardized

Response of fuel, fuel elements and gas cooled reactor cores under accidental air or water ingress conditions. Proceedings of a technical committee meeting held in Beijing, China, 25-27 October 1993

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-01-01

The meeting was convened by the IAEA on the recommendation of the IAEA`s International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Netherlands, Switzerland, the Russian Federation, the United Kingdom and the United States of America. The meeting covered the following topics: experimental investigations of the effects of air and water ingress; predicted response of fuel, graphite and other reactor components; options for minimizing or mitigating the effects of air or water ingress. 19 papers were presented at the meeting. A separate abstract was prepared for each of these papers. Refs, figs and tabs.

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

Credible investigation of air accidents

International Nuclear Information System (INIS)

Smart, K.

2004-01-01

Within the United Kingdom the Air Accidents Investigation Branch (AAIB) has been used as a model for the other transport modes accident investigation bodies. Government Ministers considered that the AAIB's approach had established the trust of the public and the aviation industry in its ability to conduct independent and objective investigations. The paper will examine the factors that are involved in establishing this trust. They include: the investigation framework; the actual and perceived independence of the accident investigating body; the aviation industry's safety culture; the qualities of the investigators and the quality of their liaison with bereaved families those directly affected by the accidents they investigate

Study on the experimental VHTR safety with analysis for a hypothetical rapid depressurization accident

International Nuclear Information System (INIS)

Mitake, S.; Suzuki, K.; Ohno, T.; Okada, T.

1982-01-01

A hypothetical rapid depressurization accident of the experimental VHTR has been analyzed, including all phenomena in the accident, from its initiating depressurization of the coolant to consequential radiological hazard. Based on reliability analysis of the engineered safety features, all possible sequences, in which the safety systems are in success or in failure, have been investigated with event tree analysis. The result shows the inherent safety characteristics of the reactor and the effectiveness of the engineered safety features. And through the analysis, it has been indicated that further investigations on some phenomena in the accident, e.g., air ingress by natural circulation flow and fission product transport in the plant, will bring forth more reasonable and sufficient safety of the reactor

Application of transient ignition model to multi-canister (MCO) accident analysis

International Nuclear Information System (INIS)

Kummerer, M.

1996-01-01

The potential for ignition of spent nuclear fuel in a Multi-Canister Overpack (MCO) is examined. A transient model is applied to calculate the highest ambient gas temperature outside an MCO wall tube or shipping cask for which a stable temperature condition exists. This integral analysis couples reaction kinetics with a description of the MCO configuration, heat and mass transfer, and fission product phenomena. It thereby allows ignition theory to be applied to various complex scenarios, including MCO water loss accidents and dry MCO air ingression

An analysis of severe air transport accidents

International Nuclear Information System (INIS)

McClure, J.D.; Luna, R.E.

1989-01-01

The objective of this paper is to analyze the severity of aircraft accidents that may involve the air transport of radioactive materials (RAM). One of the basic aims of this paper is to provide a numerical description of the severity of aircraft transport accidents so that the accident severity can be compared with the accident performance standards that are specified in IAEA Safety Series 6, the international packaging standards for the safe movement of RAM. The existing packaging regulations in most countries embrace the packaging standards developed by the IAEA. Historically, the packaging standards for Type B packages have been independent of the transport mode. That is, if the shipment occurs in a certified packaging, then the shipment can take place by any transport mode. In 1975, a legislative action occurred in the US Congress which led to the development of a package designed specifically for the air transport of plutonium. Changes were subsequently made to the US packaging regulations in 10CFR71 to incorporate the plutonium air transport performance standards. These standards were used to certify the air transport package for plutonium which is commonly referred to as PAT-1 (US NRC). The PAT-1 was certified by the US Nuclear Regulatory Commission in September 1978

Googles Augmented-Reality-Game "Ingress"

Directory of Open Access Journals (Sweden)

Karl H. Stingeder

2013-12-01

Full Text Available Das an das "Geocoaching"-Spielprinzip erinnernde Augmented-­Reality-Game "Ingress" ist auf allen mit Android-Betriebssystem ausgestatteten Smartphones spielbar. Das Alleinstellungsmerkmal des Spiels ist gleich­zeitig auch das beste Patent-Rezept für bzw. gegen notorisches "Couch-Potato-ing": Das Game-Design des "Draußenspiels" erfordert jedoch stets die physische Anwesenheit der SpielerInnen am realen Ort des Geschehens, meist touristisch attraktive Sehenswürdigkeiten und architektonisch markante Plätze. Sobald "Ingress" per App am Smart­phone gestartet wird, stehen Interaktionen und Machtverhältnisse auf dem Handy-Display im Dreh- und Angelpunkt des Geschehens. Die Spielgrafik ist auf das Wesentliche reduziert und erinnert an die Optik von Videospielen der 80er Jahre. "Ingress" im Allgemeinen sowie "Magnus13", die erste von Google organisierte Fan-Veranstaltung Österreichs mit mehreren Hundert TeilnehmerInnen und ist mit einer "digitalen Schnitzeljagd" vergleichbar: Gemeinsam ist der klassischen "Schnitzeljagd" und dem digitalen "Schere-Schein-Papier" Prinzip von "Ingress" die tragende Rolle des im Spiel entstehenden Gemeinschaftsgefühls, welche in einer hohen Langzeitmotivation mündet. Trotz aller Euphorie, ob des unkonventionellen Augmented-Reality-Spielkonzepts und des in Folge der sozialen Dynamik außergewöhnlichen Sucht­potentials, ist angesichts von Edward Snowden und vielfältiger NSA-Abhörmaßnahmen, eine gesunde Portion Skepsis angebracht: Welchen Zweck erfüllt "Ingress"? Gibt es ein verhülltes "Mittel zum Zweck"? Welche Gefahren bestehen mit einer systematischen Auswertung des umfangreichen, ortsrelevanten Daten-Sammelsuriums? "Ingress" verdeutlicht trotz aller Risiken den soziokulturellen Bedeutungswandel des digitalen Spiels: Spiele per se und das Spiel als soziale Interaktion sind aktuell dabei, alle Bereiche unseres Lebens zu erfassen.

Large eddy simulation of Loss of Vacuum Accident in STARDUST facility

International Nuclear Information System (INIS)

Benedetti, Miriam; Gaudio, Pasquale; Lupelli, Ivan; Malizia, Andrea; Porfiri, Maria Teresa; Richetta, Maria

2013-01-01

Highlights: ► Fusion safety, plasma material interaction. ► Numerical and experimental data comparison to analyze the consequences of Loss of Vacuum Accident that can provoke dust mobilization inside the Vacuum Vessel of the Nuclear Fusion Reactor ITER-like. -- Abstract: The development of computational fluid dynamic (CFD) models of air ingress into the vacuum vessel (VV) represents an important issue concerning the safety analysis of nuclear fusion devices, in particular in the field of dust mobilization. The present work deals with the large eddy simulations (LES) of fluid dynamic fields during a vessel filling at near vacuum conditions to support the safety study of Loss of Vacuum Accidents (LOVA) events triggered by air income. The model's results are compared to the experimental data provided by STARDUST facility at different pressurization rates (100 Pa/s, 300 Pa/s and 500 Pa/s). Simulation's results compare favorably with experimental data, demonstrating the possibility of implementing LES in large vacuum systems as tokamaks

Oxygen ingress : a practical look at typical ingress mechanisms and the consequences

Energy Technology Data Exchange (ETDEWEB)

Lerbscher, J. [Baker Hughes Canada, Calgary, AB (Canada); Marlowe, D. [ChevronTexaco, Kenai, AK (United States); Thomas, J. [Baker Hughes, Edmonton, AB (Canada)

2008-07-01

This paper discussed methods of identifying and treating oxygen corrosion in hydrocarbon processing facilities. Oxygen corrosion is often misdiagnosed during the course of corrosion failure analyses. Measures to find the source of ingress are only typically initiated when test results demonstrate significant sources of oxygen within processing systems. The iron oxides produced as byproducts from oxygen reactions increase pitting and corrosion rates, and most of the chemical inhibitors used in oil and gas processing are not designed to work in the presence of oxygen. Oxygen reacts with hydrogen sulfide (H{sub 2}S) to form elemental sulfur. The high pressures used in processing facilities enhance the thermodynamic and kinetic tendencies of the chemical reactions with oxygen. Sulfur particles are known to enhance corrosion rates by an order of magnitude, and can also cause fouling and flow restrictions. Oxygen ingress can occur via vapor recovery unit, vacuum excursions, and liquid storage tanks. Symptoms that indicate oxygen ingress can include the presence of iron compounds in solid samples; the presence of sulfur; fouling of wet gas transmission lines; the presence of ionic polysulfides in the aqueous phase; higher corrosion rates than predicted; and the degradation of glycols in dehydration units. Portable gas chromatography, oxygen detection vials, and X-ray diffraction analysis techniques are used to detect oxygen ingress. Real time oxygen monitors are also connected to SCADA systems. It was concluded that oxygen testing should be conducted periodically in order to identify and eliminate its source of entry. A technical summary of corrosive species was included. 1 tab., 15 figs.

Pulmonic Ingressive Speech in Shetland English

Science.gov (United States)

Sundkvist, Peter

2012-01-01

This paper presents a study of pulmonic ingressive speech, a severely understudied phenomenon within varieties of English. While ingressive speech has been reported for several parts of the British Isles, New England, and eastern Canada, thus far Newfoundland appears to be the only locality where researchers have managed to provide substantial…

Modeling moisture ingress through simplified concrete crack geometries

DEFF Research Database (Denmark)

Pease, Bradley Justin; Michel, Alexander; Geiker, Mette Rica

2011-01-01

, considered to have two parts; 1) a coalesced crack length which behaves as a free-surface for moisture ingress, and 2) an isolated microcracking length which resists ingress similarly to the bulk material. Transport model results are compared to experimental results from steel fibre reinforced concrete wedge......This paper introduces a numerical model for ingress in cracked steel fibre reinforced concrete. Details of a simplified crack are preset in the model’s geometry using the cracked hinge model (CHM). The total crack length estimated using the CHM was, based on earlier work on conventional concrete...... on moisture ingress. Results from the transport model indicate the length of the isolated microcracks was approximately 19 mm for the investigated concrete composition....

Chloride ingress in cement paste and mortar

DEFF Research Database (Denmark)

Jensen, Ole Mejlhede; Hansen, Per Freiesleben; Coats, Alison M.

1999-01-01

In this paper chloride ingress in cement paste and mortar is followed by electron probe microanalysis. The influence of several paste and exposure parameters on chloride ingress are examined (e.g., water-cement ratio, silica fume addition, exposure time, and temperature), The measurements...

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

Cerebral magnetic resonance imaging of compressed air divers in diving accidents.

Science.gov (United States)

Gao, G K; Wu, D; Yang, Y; Yu, T; Xue, J; Wang, X; Jiang, Y P

2009-01-01

To investigate the characteristics of the cerebral magnetic resonance imaging (MRI) of compressed air divers in diving accidents, we conducted an observational case series study. MRI of brain were examined and analysed on seven cases compressed air divers complicated with cerebral arterial gas embolism CAGE. There were some characteristics of cerebral injury: (1) Multiple lesions; (2) larger size; (3) Susceptible to parietal and frontal lobe; (4) Both cortical grey matter and subcortical white matter can be affected; (5) Cerebellum is also the target of air embolism. The MRI of brain is an sensitive method for detecting cerebral lesions in compressed air divers in diving accidents. The MRI should be finished on divers in diving accidents within 5 days.

Containment Performance Evaluation of a Sodium Fire Event Due to Air Ingress into the Cover Gas Region of the Reactor Vessel in the PGSFR

International Nuclear Information System (INIS)

Ahn, Sang June; Chang, Won-Pyo; Kang, Seok Hun; Choi, Chi-Woong; Yoo, Jin; Lee, Kwi Lim; Jeong, Jae-Ho; Lee, Seung Won; Jeong, Taekyeong; Ha, Kwi-Seok

2015-01-01

Comparing with the light water reactor, sodium as a reactor coolant violently reacts with oxygen in the containment atmosphere. Due to this chemical reaction, heat generated from the combustion heat increases the temperature and pressure in the containment atmosphere. The structural integrity of the containment building which is a final radiological defense barrier is threaten. A sodium fire event in the containment due to air ingress into the cover gas region in the reactor vessel is classified as one of the design basis events in the PGSFR. This event comes from a leak or crack on the reactor upper closure header surface. It accompanys an event of the radiological fission products release to the inside the containment. In this paper, evaluation for the sodium fire and radiological influence due to air ingress into the cover gas region of the reactor vessel is described. To evaluate this event, the CONTAIN-LMR, MACCS-II and OR-IGEN-II codes are used. For the sodium pool fire event in the containment, the performance evaluation and radiological influence are carried out. In the thermal hydraulic aspects, the 1 cell containment yields the most conservative result. In this event, the maximum temperature and pressure in the containment are calculated 0.185 MPa, 280.0 .deg. C, respectively. The radiological dose at the EAB and LPZ are below the acceptance criteria specified in the 10CFR100

Containment Performance Evaluation of a Sodium Fire Event Due to Air Ingress into the Cover Gas Region of the Reactor Vessel in the PGSFR

Energy Technology Data Exchange (ETDEWEB)

Ahn, Sang June; Chang, Won-Pyo; Kang, Seok Hun; Choi, Chi-Woong; Yoo, Jin; Lee, Kwi Lim; Jeong, Jae-Ho; Lee, Seung Won; Jeong, Taekyeong; Ha, Kwi-Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Comparing with the light water reactor, sodium as a reactor coolant violently reacts with oxygen in the containment atmosphere. Due to this chemical reaction, heat generated from the combustion heat increases the temperature and pressure in the containment atmosphere. The structural integrity of the containment building which is a final radiological defense barrier is threaten. A sodium fire event in the containment due to air ingress into the cover gas region in the reactor vessel is classified as one of the design basis events in the PGSFR. This event comes from a leak or crack on the reactor upper closure header surface. It accompanys an event of the radiological fission products release to the inside the containment. In this paper, evaluation for the sodium fire and radiological influence due to air ingress into the cover gas region of the reactor vessel is described. To evaluate this event, the CONTAIN-LMR, MACCS-II and OR-IGEN-II codes are used. For the sodium pool fire event in the containment, the performance evaluation and radiological influence are carried out. In the thermal hydraulic aspects, the 1 cell containment yields the most conservative result. In this event, the maximum temperature and pressure in the containment are calculated 0.185 MPa, 280.0 .deg. C, respectively. The radiological dose at the EAB and LPZ are below the acceptance criteria specified in the 10CFR100.

Chloride Ingress in Concrete Cracks under Cyclic Loading

DEFF Research Database (Denmark)

Küter, André; Geiker, Mette Rica; Olesen, John Forbes

2005-01-01

was similar for both sets and the maximum crack width was kept constant throughout the exposure period by means of precracking and an external prestressed reinforcement. Chloride profiles after 40 days revealed a considerable increase in ingress towards the crack tip in contrast to data from the literature....... Preliminary investigations have been undertaken to quantify the effect of dynamic load application on the chloride ingress into concrete cracks. Specimens were designed allowing ingress of a chloride solution into a single crack of a saturated unreinforced mortar beam. One set of specimens was subjected...... to a load frequency of ten applications per minute and a second set to one application per hour simulating static cracks, however limiting the ingress hampering effects of autogenous healing and a possible dense precipitation on the crack faces. The averaged chloride exposure interval of the crack faces...

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...

Chloride ingress profiles measured by electron probe micro analysis

DEFF Research Database (Denmark)

Jensen, Ole mejlhede; Coats, Alison M.; Glasser, Fred P.

1996-01-01

Traditional techniques for measuring chloride ingress profiles do not apply well to high performance cement paste systems; the geometric resolution of the traditional measuring techniques is too low. In this paper measurements by Electron Probe Micro Analysis (EPMA) are presented. EPMA is demonst......Traditional techniques for measuring chloride ingress profiles do not apply well to high performance cement paste systems; the geometric resolution of the traditional measuring techniques is too low. In this paper measurements by Electron Probe Micro Analysis (EPMA) are presented. EPMA...... is demonstated to determine chloride ingress in cement paste on a micrometer scale. Potential chloride ingress routes such as cracks or the paste-aggregate interface may also be characterized by EPMA. Copyright (C) 1996 Elsevier Science Ltd...

Methods for air cleaning system design and accident analysis

International Nuclear Information System (INIS)

Gregory, W.S.; Nichols, B.D.

1987-01-01

This paper describes methods, in the form of a handbook and five computer codes, that can be used for nuclear facility air cleaning system design and accident analysis. Four of the codes were developed primarily at the Los Alamos National Laboratory, and one was developed in France. Tools such as these are used to design ventilation systems in the mining industry but do not seem to be commonly used in the nuclear industry. For example, the Nuclear Air Cleaning Handbook is an excellent design reference, but it fails to include information on computer codes that can be used to aid in the design process. These computer codes allow the analyst to use the handbook information to form all the elements of a complete system design. Because these analysis methods are in the form of computer codes they allow the analyst to investigate many alternative designs. In addition, the effects of many accident scenarios on the operation of the air cleaning system can be evaluated. These tools originally were intended for accident analysis, but they have been used mostly as design tools by several architect-engineering firms. The Cray, VAX, and personal computer versions of the codes, an accident analysis handbook, and the codes availability will be discussed. The application of these codes to several design operations of nuclear facilities will be illustrated, and their use to analyze the effect of several accident scenarios also will be described

Safety implications of a graphite oxidation accident in the compact ignition tokamak device

International Nuclear Information System (INIS)

Merrill, B.J.; O'Brien, M.H.

1989-01-01

This paper addresses the possible safety consequences of an air ingress accident for the Compact Ignition Tokamak (CIT) device. An experimental program was undertaken to determine oxidation rates of four nuclear grade graphites in air at temperatures ranging from 800 to 1800 C and flow velocities from 3 to 7 m/s. On the basis of these test results, an analytic model was developed to assess the extent of first wall/divertor protective tile oxidation and the amount of energy released from this oxidation. For CIT, a significant restriction to vacuum vessel air inflow will be provided by the air seals and walls of the surrounding test cells. Under these conditions, the graphite oxidation reaction inside the vacuum vessel will become oxygen starved within minutes of the onset of this event. Since significant oxidation rates were not achieved, the heat release did not elevate structural temperatures to levels of concern with regard to activated material release. 7 refs., 9 figs

Numerical analysis on ingress-of-coolant events in fusion reactors with TRAC-PF1 code

International Nuclear Information System (INIS)

Ose, Yasuo; Takase, Kazuyuki; Akimoto, Hajime

2000-01-01

As for accident events related with thermal-hydraulics, in a fusion experimental reactor an ingress-of-coolant event (ICE) and a loss-of-vacuum-accident event (LOVA) should be considered. An integrated ICE/LOVA test apparatus is under planning in order to estimate quantitatively heat transfer and fluid flow characteristics under ICE and LOVA events. This study was carried out to predict numerically the thermal-hydraulic characteristics in fusion reactors at the ICE events before construction of the integrated ICE/LOVA test apparatus. The TRAC-PF1 code, which was originally developed for the thermal-hydraulic safety analysis in light water reactors, was used. The numerical analyses were performed for two kinds of system configuration with/without a pressure-suppression tank:the former for is investigation of the pressure rise characteristics and two-phase flow behavior; the latter for estimation of an effect of the pressure reduction due to the pressure-suppression tank. From the present analytical results, effects of the ingress water flow rate and vessel temperatures on the pressure rise ware clarified quantitatively. Furthermore, the pressure-rise suppression effect due to the vapor condensation in the pressure-suppression tank was predicted numerically. In addition, the useful information regarding to the design of the integrated ICE/LOVA test apparatus and the knowledge with respect to the effective usage of the TRAC-PF1 code were obtained through the present numerical study. (author)

Prime Example Ingress Reframing the Pervasive Game Design Framework (PGDF

Directory of Open Access Journals (Sweden)

Heinrich Söbke

2017-06-01

Full Text Available The growing availability of mobile communication infrastructure over the last decade has contributed significantly to the maturity of Pervasive Gaming. The massive success of games such as Ingress and Pokémon Go made pervasive gaming a viable option for transforming learning. By its adaptability to location and context, pervasive technology is a valuable support for the design of engaging learning experiences. Despite profound examples of pervasive gaming as learning tool, there is still a lack of reliable methodologies to construct purposeful pervasive learning experiences. The Pervasive Game Design Framework (PGDF is intended to fill this gap. In this article, we present the PGDF using the example of Ingress. Ingress is a prominent pervasive game, as it has received huge attention since its appearance in 2012. A large community of players and third-party-tool suppliers has created a rich set of experiences since then. In this research, we examine Ingress according to PGDF’s categories based on a survey among long-term Ingress players (N=133. Founded on this analysis we identify three main benefits for Ingress players. Furthermore, we discuss the consequences of these findings on the PGDF. Summarizing, this work strengthens the applicability of the PGDF, in order to enable the construction of enriched pervasive learning experiences.

Air oxidation of Zircaloy-4 in the 600-1000 °C temperature range: Modeling for ASTEC code application

Science.gov (United States)

Coindreau, O.; Duriez, C.; Ederli, S.

2010-10-01

Progress in the treatment of air oxidation of zirconium in severe accident (SA) codes are required for a reliable analysis of severe accidents involving air ingress. Air oxidation of zirconium can actually lead to accelerated core degradation and increased fission product release, especially for the highly-radiotoxic ruthenium. This paper presents a model to simulate air oxidation kinetics of Zircaloy-4 in the 600-1000 °C temperature range. It is based on available experimental data, including separate-effect experiments performed at IRSN and at Forschungszentrum Karlsruhe. The kinetic transition, named "breakaway", from a diffusion-controlled regime to an accelerated oxidation is taken into account in the modeling via a critical mass gain parameter. The progressive propagation of the locally initiated breakaway is modeled by a linear increase in oxidation rate with time. Finally, when breakaway propagation is completed, the oxidation rate stabilizes and the kinetics is modeled by a linear law. This new modeling is integrated in the severe accident code ASTEC, jointly developed by IRSN and GRS. Model predictions and experimental data from thermogravimetric results show good agreement for different air flow rates and for slow temperature transient conditions.

Supply Ventilation and Prevention of Carbon Monoxide (II) Ingress into Building Premises

Science.gov (United States)

Litvinova, N. A.

2017-11-01

The article contains the relationships of carbon monoxide (II) concentration versus height-above-ground near buildings derived based on results of studies. The results of studies are crucial in preventing external pollutants ingress into a ventilation system. Being generated by external emission sources, such as motor vehicles and city heating plants, carbon monoxide (II) enters the premises during operation of a supply ventilation system. Fresh air nomographic charts were drawn to select the height of a fresh air intake into the ventilation system. Nomographic charts take into account external sources. The selected emission sources are located at various levels above ground relative to the building. The recommendations allow designing supply ventilation taking into account the quality of ambient air through the whole building height.

Source term estimation for small sized HTRs

International Nuclear Information System (INIS)

Moormann, R.

1992-08-01

Accidents which have to be considered are core heat-up, reactivity transients, water of air ingress and primary circuit depressurization. The main effort of this paper belongs to water/air ingress and depressurization, which requires consideration of fission product plateout under normal operation conditions; for the latter it is clearly shown, that absorption (penetration) mechanisms are much less important than assumed sometimes in the past. Source term estimation procedures for core heat-up events are shortly reviewed; reactivity transients are apparently covered by them. Besides a general literature survey including identification of areas with insufficient knowledge this paper contains some estimations on the thermomechanical behaviour of fission products in water in air ingress accidents. Typical source term examples are also presented. In an appendix, evaluations of the AVR experiments VAMPYR-I and -II with respect to plateout and fission product filter efficiency are outlined and used for a validation step of the new plateout code SPATRA. (orig.)

Water Ingress Testing of the Turbula Jar and U-233 Lead Pig Containers

Energy Technology Data Exchange (ETDEWEB)

Reeves, Kirk Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Karns, Tristan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Smith, Paul Herrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-11-02

Understanding the water ingress behavior of containers used at the TA-55 Plutonium Facility has significant implications for criticality safety. The purpose of this report is to document the water ingress behavior of the Turbula Jar with Bakelite lid and Viton gaskets (Turbula Jar) used in oxide blending operations and the U-233 lead pig container used to store and transport U-233 material. The technical basis for water resistant containers at TA-55 is described in LA-UR-15-22781, “Water Resistant Container Technical Basis Document for the TA-55 Criticality Safety Program.” Testing of the water ingress behavior of various containers is described in LA-CP-13-00695, “Water Penetration Tests on the Filters of Hagan and SAVY Containers,” LA-UR-15-23121, “Water Ingress into Crimped Convenience Containers under Flooding Conditions,” and in LA-UR- 16-2411, “Water Ingress Testing for TA-55 Containers.” Water ingress criteria are defined in TA55-AP-522 “TA-55 Criticality Safety Program”, and in PA-RD-01009 “TA55 Criticality Safety Requirements.” The water ingress criteria for submersion is no more than 50 ml of water ingress at a 6” water column height for a period of 2 hours.

Aerosol challenges to air cleaning systems during severe accidents in nuclear plants

International Nuclear Information System (INIS)

Gieseke, J.A.

1985-01-01

A variety of air cleaning systems may be operating in nuclear power plants and under severe accident conditions, these systems may be treating airborne concentrations of aerosols which are very high. Predictions of airborne aerosol concentrations in nuclear power plant containments under severe accident conditions are reviewed to provide a basis for evaluating the potential effects on the air cleaning systems. The air cleaning systems include filters, absorber beds, sprays, water pools, ice beds, and condensers. Not all of these were intended to operate as air cleaners but will in fact be good aerosol collectors. Knowledge of expected airborne concentrations will allow better evaluation of system performances

An overview of the severe accident research activities within the LACOMERA platform at the Forschungszentrum Karlsruhe

International Nuclear Information System (INIS)

Miassoedov, A.; Alsmeyer, H.; Meyer, L.; Steinbrueck, M.; Tromm, W.

2006-01-01

The LACOMERA project at the Forschungszentrum Karlsruhe, Germany, is a 4 year action within the 5th Framework Programme of the EU which started in September 2002. Overall objective of the project is to offer research institutions from the EU member countries and associated states access to four large-scale experimental facilities QUENCH, LIVE, DISCO, and COMET. These facilities can be used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. The paper summarises the main results obtained in the following experiments performed up to now. QUENCH-L1: Impact of air ingression on core degradation. The test provides unique data for the investigation of air ingress phenomenology in conditions as representative of a spent fuel pool accident as possible; QUENCH-L2: Boil-off of a flooded bundle. The test is of a generic interest for all reactor types, provided a link between the severe accident and design basis areas, and would deliver oxidation and thermal hydraulic data at high temperatures. DISCO-L1: Thermal hydraulic behaviour of the corium melt dispersion neglecting the chemical effects such as hydrogen generation and combustion. COMET-L1: Long-term 2D concrete ablation in a siliceous concrete cavity at intermediate decay heat power level with a top flooding phase after a phase of dry concrete erosion. COMET-L2: Investigation of long-term melt-concrete interaction of metallic corium in a cylindrical siliceous concrete cavity under dry conditions with decay heat simulation of intermediate power during the first test phase, and subsequently at reduced power during the second test phase. (author)

Methods and data for HTGR fuel performance and radionuclide release modeling during normal operation and accidents for safety analysis

International Nuclear Information System (INIS)

Verfondern, K.; Martin, R.C.; Moormann, R.

1993-01-01

The previous status report released in 1987 on reference data and calculation models for fission product transport in High-Temperature, Gas-Cooled Reactor (HTGR) safety analyses has been updated to reflect the current state of knowledge in the German HTGR program. The content of the status report has been expanded to include information from other national programs in HTGRs to provide comparative information on methods of analysis and the underlying database for fuel performance and fission product transport. The release and transport of fission products during normal operating conditions and during the accident scenarios of core heatup, water and air ingress, and depressurization are discussed. (orig.) [de

Chloride Ingress into Concrete under Water Pressure

DEFF Research Database (Denmark)

Lund, Mia Schou; Sander, Lotte Braad; Grelk, Bent

2011-01-01

The chloride ingress into concrete under water pressures of 100 kPa and 800 kPa have been investigated by experiments. The specimens were exposed to a 10% NaCl solution and water mixture. For the concrete having w/c = 0.35 the experimental results show the chloride diffusion coefficient at 800 k......Pa (～8 atm.) is 12 times greater than at 100 kPa (～1 atm.). For w/c = 0.45 and w/c = 0.55 the chloride diffusion coefficients are 7 and 3 times greater. This means that a change in pressure highly influences the chloride ingress into the concrete and thereby the life length models for concrete structures....

HTGR [High Temperature Gas-Cooled Reactor] ingress analysis using MINET

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Yang, J.W.; Kroeger, P.G.; Mallen, A.N.; Aronson, A.L.

1989-04-01

Modeling of water/steam ingress into the primary (helium) cooling circuit of a High Temperature Gas-Cooled Reactor (HTGR) is described. This modeling was implemented in the MINET Code, which is a program for analyzing transients in intricate fluid flow and heat transfer networks. Results from the simulation of a water ingress event postulated for the Modular HTGR are discussed. 27 refs., 6 figs., 6 tabs

Screening of Low Clinker Binders, Compressive Strength and Chloride Ingress

DEFF Research Database (Denmark)

Geiker, Mette Rica; De Weerdt, Klaartje; Garzón, Sergio Ferreiro

2017-01-01

This paper reports an initial screening of potential new binders for concrete with reduced CO2-emission. Mortars cured saturated for 90 days are compared with regard to a) compressive strength of mortars with similar water-to-binder ratio, and b) chloride ingress in similar design strength mortar...... compromising the 90 days compressive strength and resistance to chloride ingress in marine exposure by using selected alternative binders....

Degradation of fly ash concrete under the coupled effect of carbonation and chloride aerosol ingress

International Nuclear Information System (INIS)

Liu, Jun; Qiu, Qiwen; Chen, Xiaochi; Wang, Xiaodong; Xing, Feng; Han, Ningxu; He, Yijian

2016-01-01

Highlights: • Carbonation affects the chloride profile in concrete under chloride aerosol attack. • The chloride binding capacity can be reduced by the presence of carbonation. • Carbonation increases the rate of chloride diffusion for chloride aerosol ingress. • Chloride aerosol ingress reduces the carbonation depth and increases the pH value. • The use of fly ash in concrete enhances the resistance of chloride aerosol ingress. - Abstract: This paper presents an experimental investigation regarding the coupled effect of carbonation and chloride aerosol ingress on the durability performance of fly ash concrete. Test results demonstrate that carbonation significantly affects the chloride ingress profile, reduces the chloride binding capacity, and accelerates the rate of chloride ion diffusion. On the other hand, the carbonation rate of fly ash concrete is reduced by the presence of chlorides aerosol. The interaction nature between concrete carbonation and chloride aerosol ingress is also demonstrated by the microscopic analysis results obtained from scanning electron microscope and mercury intrusion porosimetry.

Analysis of emergency response procedures and air traffic accidents ...

African Journals Online (AJOL)

Incessant air transport accidents have been a source of concern to stakeholders and aviation experts in Nigeria, yet the response and process has not been adequately appraised. This study attempts an evaluation of the emergency response procedures in the aviation industry with particular focus on Murtala Muhammed ...

Literature investigation of air/steam ingress through small cracks in concrete wall under pressure differences

International Nuclear Information System (INIS)

Jiang, J.T.

2008-01-01

Traditionally within CANDU safety analysis, a loss coefficient of ∼2.8 is used to characterize turbulent flow leakage through narrow, sharp-edged cracks into, and out of Steam Protected Rooms (SPRs). In the event of main steam line break (MSLB), the pressure differences observed between SPRs and the surrounding area of the powerhouse range from 0.01kPa to 0.1 kPa. The relatively low pressure differences, coupled with narrow crack sizes, for instance, below 1 mm, may result in laminar flow leakage pathways as opposed to the turbulent variety assumed in analysis. The main purpose of this paper is thus (a) to calculate the loss coefficient for laminar flow through small cracks; and (b) to assess the effect of steam ingress to SPRs when the flow through some or all of the room leakage area is assumed to be laminar. Based on the literature review, the loss coefficient for laminar flow, through 1 mm crack size at 0.1 kPa pressure difference, ranges from 10 to about 65. This value represents an increase in loss coefficient of 3 ∼ 22 times the loss coefficient used for SPR safety analysis. The actual volumetric leakage rate is therefore 3 ∼ 8 times smaller than the amount previously applied. This paper demonstrates how the traditional loss coefficient used in safety analysis is extremely conservative in the analysis of the SPRs steam ingress phenomenon. (author)

Literature investigation of air/steam ingress through small cracks in concrete wall under pressure differences

Energy Technology Data Exchange (ETDEWEB)

Jiang, J.T. [McMaster Univ., Engineering Physics Dept., Hamilton, Ontario (Canada)], E-mail: jiangj3@mcmaster.ca

2008-07-01

Traditionally within CANDU safety analysis, a loss coefficient of {approx}2.8 is used to characterize turbulent flow leakage through narrow, sharp-edged cracks into, and out of Steam Protected Rooms (SPRs). In the event of main steam line break (MSLB), the pressure differences observed between SPRs and the surrounding area of the powerhouse range from 0.01kPa to 0.1 kPa. The relatively low pressure differences, coupled with narrow crack sizes, for instance, below 1 mm, may result in laminar flow leakage pathways as opposed to the turbulent variety assumed in analysis. The main purpose of this paper is thus (a) to calculate the loss coefficient for laminar flow through small cracks; and (b) to assess the effect of steam ingress to SPRs when the flow through some or all of the room leakage area is assumed to be laminar. Based on the literature review, the loss coefficient for laminar flow, through 1 mm crack size at 0.1 kPa pressure difference, ranges from 10 to about 65. This value represents an increase in loss coefficient of 3 {approx} 22 times the loss coefficient used for SPR safety analysis. The actual volumetric leakage rate is therefore 3 {approx} 8 times smaller than the amount previously applied. This paper demonstrates how the traditional loss coefficient used in safety analysis is extremely conservative in the analysis of the SPRs steam ingress phenomenon. (author)

Chloride ingress in cement paste and mortar measured by Electron Probe Micro Analysis

DEFF Research Database (Denmark)

Jensen, Ole Mejlhede

1998-01-01

is an important basis for designing the durability of concrete structures. As an example the Danish Great Belt link is designed to have 100 years durability based on calculation of chloride ingress.During the last 15 years the types of concrete used in practice have changed substantially. Due to plasticizers......Chloride ingress is a common cause of deterioration of reinforced concrete structures. Concrete may be exposed to chloride by sea water or deicing salts. The chloride initiates corrosion of the reinforcement which through expansion disrupts the concrete. Modelling the chloride ingress...... and mineral additives concretes with higher strengths and reduced permeability are produced. Recently it has become clear that traditional chloride ingress models do not apply to modern concretes. Actually, the life time model used for the Danish Great Belt link has shown up to be based on wrong assumptions...

Plating end fittings to reduce hydrogen ingress at rolled joints in CANDU reactors

International Nuclear Information System (INIS)

White, A.J.; Urbanic, V.F.; Bahurmuz, A.A.; Clendening, W.R.; Joynes, R.; McDougall, G.M.; Skinner, B.C.; Venkatapathi, S.

1993-10-01

Zr-2.5Nb pressure tubes in CANDU nuclear reactors absorb hydrogen at a low rate from the primary heat transport water circulated through the tubes. Extra hydrogen is picked up at the rolled joints that connect the pressure tubes to out-of-core steel piping. This enhanced ingress may contribute to pressure-tube cracking at incorrectly assembled joints. The risk of pressure-tube failure has been decreased by ensuring correct joint assembly, and could be further decreased by reducing hydrogen ingress at rolled joints. This paper reviews progress toward using plated end fittings to reduce rolled-joint hydrogen ingress

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

Air Oxidation Behaviors of Zircaloy-4 Cladding During a LOCA In Spent Fuel Pool

International Nuclear Information System (INIS)

Bang, Je Geon; Chun, Tae Hyun; Kim, Sun Ki; Koo, Yang Hyun

2014-01-01

It is well known that air oxidation induces a serious degradation of the Zircaloy cladding material, compared with steam oxidation. From the oxidant point of view, in comparison with steam, chemical heat release during oxidation in air is higher by 80%, which may lead to a more rapid degradation of the Zircaloy cladding, and further evolution of the accident.. Additionally, the oxidation kinetics in air is much faster than in steam due to the formation of non-protective oxide layer. From the safety point of view, the barrier effect of the cladding against release of fission products is lost much earlier in air compared to steam. The objective of this study is to investigate the oxidation behaviors of fuel cladding in two different conditions such as isothermal and transient condition and to generate its kinetic data under an accident condition in the spent fuel pool. In this study, the oxidation behaviors and its kinetics of the Zircaloy-4 were investigated in air environment for various air ingress scenarios in the temperature range 600 .deg. C-1,400 .deg. C by thermo-gravimetric analysis. In this study, the oxidation behaviors of the Zircaloy-4 for both isothermal condition and transient condition were investigated in air environment. In comparison with isothermal condition, the retardation of oxidation rate in transient condition was observed at both 1,200 .deg. C and 1,400 .deg. C. This seems to be ascribed to the effect of thin oxide formed during a heating

Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Zheng Yanhua; Shi Lei; Wang Yan

2010-01-01

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants. Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.

The risk of an air accident as a result of a serious incident of the hybrid type

International Nuclear Information System (INIS)

Skorupski, Jacek

2015-01-01

Safety in air traffic is a multilayered concept and consists of many safety barriers. The practical side of increasing safety is mainly based on analysing the causes of accidents and incidents. This analysis leads to finding gaps in the safety structure and to developing corrective recommendations in order to eliminate them. In this paper we indicate that this practice is insufficient. Most incidents could transform into accidents with fatalities. The standard method of investigating incidents does not answer the question as to whether safety barrier is permanent or whether it was activated accidentally. This paper proposes a new method for analysing incidents aimed at finding their consequences rather than their causes. This makes it possible to find areas that need improvement. Stochastic, timed, coloured Petri nets were used for the analysis. There are three types of air traffic incidents, distinguished according to events that lead to a transformation of an incident into an accident: causal and temporal. The hybrid case, in which both types are important, has been discussed in detail. The method is useful in evaluating the current level of safety in air traffic. Applicability of this method has been shown on the example of the runway incursion problem. - Highlights: • Current accident investigation procedure is not sufficient. • New method aimed to study incident’s conversion into accident was proposed. • The Petri net model of air traffic accident was developed. • Method allows calculating accident probability. • The hybrid case in which both causal and temporal events are important is discussed

Chloride ingress prediction

DEFF Research Database (Denmark)

Frederiksen, Jens Mejer; Geiker, Mette Rica

2008-01-01

Prediction of chloride ingress into concrete is an important part of durability design of reinforced concrete structures exposed to chloride containing environment. This paper presents experimentally based design parameters for Portland cement concretes with and without silica fume and fly ash...... in marine atmospheric and submersed South Scandinavian environment. The design parameters are based on sequential measurements of 86 chloride profiles taken over ten years from 13 different types of concrete. The design parameters provide the input for an analytical model for chloride profiles as function...... of depth and time, when both the surface chloride concentration and the diffusion coefficient are allowed to vary in time. The model is presented in a companion paper....

Witwatersrand Water Ingress Project - Information Management System (WWIPIMS), South Africa

Energy Technology Data Exchange (ETDEWEB)

Hild, S.; Sieste, M.; Barth, A.; Rudinskaya, J. [Beak Consultants GmbH, Freiberg (Germany); Croukamp, L.; Roos, M. [Council for Geoscience (CGS), Pretoria (South Africa)

2006-07-01

The Witwatersrand Water Ingress Project at the Council for Geoscience, South Africa (CGS) deals with an inventory, a risk assessment and the development of rehabilitation strategies for abandoned mining sites in the Witwatersrand Mining Basin. The main focus is the prevention of water ingress and to understand the future decanting scenario. An Information Management System consisting of both a relational database and an application for the Witwatersrand Water Ingress Project is established for accessing and managing all project-related data. This easy to use application makes the data available to all staff at the CGS via several modules as well as a GIS-component for accessing and querying spatial data. This will enable the scientists to derive further knowledge of the water flowing processes by directly using all of the existing up-to-date data. Many additional functions, such as the support for map printing on demand, extensive possibilities for inquiries, data import and export, diagrams and a GIS-viewer for spatial inquiries do complete the system. (orig.)

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.

The Safety of Hospital Beds: Ingress, Egress, and In-Bed Mobility.

Science.gov (United States)

Morse, Janice M; Gervais, Pierre; Pooler, Charlotte; Merryweather, Andrew; Doig, Alexa K; Bloswick, Donald

2015-01-01

To explore the safety of the standard and the low hospital bed, we report on a microanalysis of 15 patients' ability to ingress, move about the bed, and egress. The 15 participants were purposefully selected with various disabilities. Bed conditions were randomized with side rails up or down and one low bed with side rails down. We explored the patients' use of the side rails, bed height, ability to lift their legs onto the mattress, and ability to turn, egress, and walk back to the chair. The standard bed was too high for some participants, both for ingress and egress. Side rails were used by most participants when entering, turning in bed, and exiting. We recommend that side rails be reconsidered as a means to facilitate in-bed movement, ingress, and egress. Furthermore, single deck height settings for all patients are not optimal. Low beds as a safety measure must be re-evaluated.

Safety assessment for electricity generation failure accident of gas cooled nuclear power plant using system dynamics (SD) method

Energy Technology Data Exchange (ETDEWEB)

Woo, Tae Ho [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering

2013-04-15

The power production failure happens in the loss of coolant of the nuclear power plants (NPPs). The air ingress is a serious accident in gas cooled NPPs. The quantification of the study performed by the system dynamics (SD) method which is processed by the feedback algorithms. The Vensim software package is used for the simulation, which is performed by the Monte-Carlo method. Two kinds of considerations as the economic and safety properties are important in NPPs. The result shows the stability of the operation when the power can be decided. The maximum value of risk is the 11.77 in 43rd and the minimum value is 0.0 in several years. So, the success of the circulation of coolant is simulated by the dynamical values. (orig.)

Persistence of airline accidents.

Science.gov (United States)

Barros, Carlos Pestana; Faria, Joao Ricardo; Gil-Alana, Luis Alberiko

2010-10-01

This paper expands on air travel accident research by examining the relationship between air travel accidents and airline traffic or volume in the period from 1927-2006. The theoretical model is based on a representative airline company that aims to maximise its profits, and it utilises a fractional integration approach in order to determine whether there is a persistent pattern over time with respect to air accidents and air traffic. Furthermore, the paper analyses how airline accidents are related to traffic using a fractional cointegration approach. It finds that airline accidents are persistent and that a (non-stationary) fractional cointegration relationship exists between total airline accidents and airline passengers, airline miles and airline revenues, with shocks that affect the long-run equilibrium disappearing in the very long term. Moreover, this relation is negative, which might be due to the fact that air travel is becoming safer and there is greater competition in the airline industry. Policy implications are derived for countering accident events, based on competition and regulation. © 2010 The Author(s). Journal compilation © Overseas Development Institute, 2010.

Ingress in Geography: Portals to Academic Success?

Science.gov (United States)

Davis, Michael

2017-01-01

Niantic Labs has developed an augmented virtual reality mobile app game called Ingress in which agents must seek out and control locations for their designated factions. The app uses the Google Maps interface along with GPS to enhance a geocaching-like experience with elements of other classical games such as capture-the-flag. This study aims to…

Study on heat transfer and fluid flow in the stand pipe rupture accident

International Nuclear Information System (INIS)

Fumizawa, Motoo; Hishida, Makoto

1991-09-01

This paper deals with an experimental investigation of the buoyancy driven exchange flow which takes place through a narrow cylindrical channel, during the stand pipe rupture accident in a high temperature gas-cooled reactor (HTGR). The velocity distribution through the cylindrical channel is measured by a laser Doppler velocimeter, in order to evaluate the air ingress flow rate. The experiments are performed under atmospheric pressure with nitrogen as a working fluid. Rayleigh number ranges from 1.3 x 10 7 to 7.0 x 10 7 . The following conclusions were obtained: (1) The laser Doppler velocimeter was found a good method for the measurement of the velocity of the exchange flow. (2) When the temperature of the hemisphere and the bottom heated plate, which simulate the top cover of the reactor, was kept uniform, the volumetric exchange flow rate agreed well with Epstein's result. (3) The exchange flow through a narrow cylindrical channel fluctuated irregularly with time and space. (author)

Ingression-type cell migration drives vegetal endoderm internalisation in the Xenopus gastrula.

Science.gov (United States)

Wen, Jason Wh; Winklbauer, Rudolf

2017-08-10

During amphibian gastrulation, presumptive endoderm is internalised as part of vegetal rotation, a large-scale movement that encompasses the whole vegetal half of the embryo. It has been considered a gastrulation process unique to amphibians, but we show that at the cell level, endoderm internalisation exhibits characteristics reminiscent of bottle cell formation and ingression, known mechanisms of germ layer internalisation. During ingression proper, cells leave a single-layered epithelium. In vegetal rotation, the process occurs in a multilayered cell mass; we refer to it as ingression-type cell migration. Endoderm cells move by amoeboid shape changes, but in contrast to other instances of amoeboid migration, trailing edge retraction involves ephrinB1-dependent macropinocytosis and trans -endocytosis. Moreover, although cells are separated by wide gaps, they are connected by filiform protrusions, and their migration depends on C-cadherin and the matrix protein fibronectin. Cells move in the same direction but at different velocities, to rearrange by differential migration.

Passive decay heat removal by natural air convection after severe accidents

Energy Technology Data Exchange (ETDEWEB)

Erbacher, F.J.; Neitzel, H.J. [Forschungszentrum Karlsruhe Institut fur Angewandte Thermo- und Fluiddynamik, Karlsruhe (Germany); Cheng, X. [Technische Universitaet Karlsruhe Institut fur Stroemungslehre und Stroemungsmaschinen, Karlsruhe (Germany)

1995-09-01

The composite containment proposed by the Research Center Karlsruhe and the Technical University Karlsruhe is to cope with severe accidents. It pursues the goal to restrict the consequences of core meltdown accidents to the reactor plant. One essential of this new containment concept is its potential to remove the decay heat by natural air convection and thermal radiation in a passive way. To investigate the coolability of such a passive cooling system and the physical phenomena involved, experimental investigations are carried out at the PASCO test facility. Additionally, numerical calculations are performed by using different codes. A satisfying agreement between experimental data and numerical results is obtained.

Debris interactions in reactor vessel lower plena during a severe accident. II. Integral analysis

International Nuclear Information System (INIS)

Suh, K.Y.; Henry, R.E.

1996-01-01

For pt.I see ibid., p.147-63, 1996. The integral physico-numerical model for the reactor vessel lower head response has been exercised for the TMI-2 accident and possible severe accident scenarios in PWR and BWR designs. The proposed inherent cooling mechanism of the reactor material creep and subsequent water ingression implemented in this predictive model provides a consistent representation of how the debris was finally cooled in the TMI-2 accident and how the reactor lower head integrity was maintained during the course of the incident. It should be recalled that in order for this strain to occur, the vessel lower head had to achieve temperatures in excess of 1000 C. This is certainly in agreement with the temperatures determined by metallographic examinations during the TMI-2 vessel inspection program. The integral model was also applied to typical PWR and BWR lower plena with and without structures under pressurized conditions spanning the first relocation of core material to the reactor vessel failure due to creep without recovery actions. The design application results are presented with particular attention being focused on water ingression into the debris bed through the gap formed between the debris and the vessel wall. As an illustration of the accident management application, the lower plenum with structures was recovered after an extensive amount of creep had damaged the vessel wall. The computed lower head temperatures were found to be significantly lower (by more than 300 K in this particular example) with recovery relative to the case without recovery. This clearly demonstrates the potential for in-vessel cooling of the reactor vessel without a need to externally submerge the lower head should such a severe accident occur as core melting and relocation. (orig.)

Chloride Ingress in Chemically Activated Calcined Clay-Based Cement

Directory of Open Access Journals (Sweden)

Joseph Mwiti Marangu

2018-01-01

Full Text Available Chloride-laden environments pose serious durability concerns in cement based materials. This paper presents the findings of chloride ingress in chemically activated calcined Clay-Ordinary Portland Cement blended mortars. Results are also presented for compressive strength development and porosity tests. Sampled clays were incinerated at a temperature of 800°C for 4 hours. The resultant calcined clay was blended with Ordinary Portland Cement (OPC at replacement level of 35% by mass of OPC to make test cement labeled PCC35. Mortar prisms measuring 40 mm × 40 mm × 160 mm were cast using PCC35 with 0.5 M Na2SO4 solution as a chemical activator instead of water. Compressive strength was determined at 28th day of curing. As a control, OPC, Portland Pozzolana Cement (PPC, and PCC35 were similarly investigated without use of activator. After the 28th day of curing, mortar specimens were subjected to accelerated chloride ingress, porosity, compressive strength tests, and chloride profiling. Subsequently, apparent diffusion coefficients (Dapp were estimated from solutions to Fick’s second law of diffusion. Compressive strength increased after exposure to the chloride rich media in all cement categories. Chemically activated PCC35 exhibited higher compressive strength compared to nonactivated PCC35. However, chemically activated PCC35 had the least gain in compressive strength, lower porosity, and lower chloride ingress in terms of Dapp, compared to OPC, PPC, and nonactivated PCC35.

Prediction of chloride ingress and binding in cement paste

DEFF Research Database (Denmark)

Geiker, Mette Rica; Nielsen, Erik Pram; Herforth, Duncan

2007-01-01

This paper summarizes recent work on an analytical model for predicting the ingress rate of chlorides in cement-based materials. An integral part of this is a thermodynamic model for predicting the phase equilibria in hydrated Portland cement. The model’s ability to predict chloride binding...... in Portland cement pastes at any content of chloride, alkalis, sulfates and carbonate was verified experimentally and found to be equally valid when applied to other data in the literature. The thermodynamic model for predicting the phase equilibria in hydrated Portland cement was introduced into an existing...... Finite Difference Model for the ingress of chlorides into concrete which takes into account its multi-component nature. The “composite theory” was then used to predict the diffusivity of each ion based on the phase assemblage present in the hydrated Portland cement paste. Agreement was found between...

Supply air filters after the nuclear reactor accident at Chernobyl

International Nuclear Information System (INIS)

Bonka, H.

1989-01-01

In the case of increased activity concentration in the air supply air filters are the facility compounds where enhanced activity is collected. Therefore, it was understandable that the people put questions about the doses caused by supply air filters after the nuclear reactor accident at Chernobyl. When comparing the local dose rate at a distance of 1 m in front of filters with the outdoor local dose rate due to dry deposited radionuclides, nearly the same local dose rate results assuming an air flow rate of approx. 60 m 3 /h. Supposing a 10 hours stay at a distance of 1 m in front of filters and an air flow rate of approx. 5000 m 3 /h the same dose is obtained after a 10 days delay as the dose due to outdoor inhalation. At Aachen, the local dose rate near to filters increased up to approx. 10 μSv/h. After a suitable time delay of one month filters could be rejected like normal waste. A review is given on individual measured values

Actinide nuclides in environmental air and precipitation samples after the Chernobyl accident

International Nuclear Information System (INIS)

Rosner, G.; Hoetzl, H.; Winkler, R.

1988-01-01

The present paper describes the analysis of isotopes of uranium, neptunium, plutonium, americium and curium, in air and deposition samples taken at our laboratory site 10 km north of Munich, subsequent to the Chernobyl accident. Uranium-234, 237 U, 238 U, 239 Np, 238 Pu, 239+240 Pu and 242 Cm have been identified and upper limits of detection have been established for 241 Am and 244 Cm. Deposition and air concentration values are discussed. 12 refs., 1 fig., 2 tabs

French gas cooled reactor experience with moisture ingress

International Nuclear Information System (INIS)

Bastien, D.; Brie, M.

1995-01-01

During the history of operation of six gas cooled reactors in France, some experience has been gained with accidental water ingress into the primary system. This occurred as a result of leaks in steam generators. This paper describes the cause of the leaks, and the resulting consequences. (author). 2 refs, 8 figs

Oxidation and Volatilization from Tantalum Alloy T-222 During Air Exposure

Energy Technology Data Exchange (ETDEWEB)

Smolik, Galen Richard; Petti, David Andrew; Sharpe, John Phillip; Schuetz, Stanley Thomas

2000-10-01

Tantalum alloys are one of the refractory metals with renewed consideration for high temperatures in fusion reactor applications. Tantalum alloys perform well in protective environments but oxidized readily in gases containing higher oxygen levels. In addition, the radioactive isotope Ta-182 would be produced in tantalum and could be a significant contributor to dose if mobilized. Other isotopes of importance are produced from tungsten and hafnium. Mobilization of activated products during an accident with air ingress is therefore a safety issue. In this study, we measured the extent of oxidation and mobilization from tantalum alloy T-222 oxidized in flowing air between 500 and 1200°C. This alloy nominally contains 10 wt% tungsten, 2.5 wt% hafnium and 0.01 wt% carbon. We found that the mobilization of Ta and Hf was closely linked to the occurrence of oxide spalling. These elements showed no migration from the test chamber. Some W was mobilized by volatilization as evidenced by transport from the chamber. Tungsten volatilization could occur primarily during initial stages of oxidation before an oxide scale forms and impedes the process. The mobilization of Ta and W are presented in terms of the mass flux (g/m 2 -h) as a function of test temperature. These measurements along with specific designs, activation calculations, and accident scenarios provide information useful for dose calculations of future fusion devices

Oxidation and Volatilization from Tantalum Alloy T-222 During Air Exposure

Energy Technology Data Exchange (ETDEWEB)

Smolik, G.R.; Petti, D.A.; Sharpe, J.P.; Schuetz, S.T.

2000-10-31

Tantalum alloys are one of the refractory metals with renewed consideration for high temperatures in fusion reactor applications. Tantalum alloys perform well in protective environments but oxidized readily in gases containing higher oxygen levels. In addition, the radioactive isotope Ta-182 would be produced in tantalum and could be a significant contributor to dose if mobilized. Other isotopes of importance are produced from tungsten and hafnium. Mobilization of activated products during an accident with air ingress is therefore a safety issue. In this study, we measured the extent of oxidation and mobilization from tantalum alloy T-222 oxidized in flowing air between 500 and 1200 C. This alloy nominally contains 10 wt% tungsten, 2.5 wt% hafnium and 0.01 wt% carbon. We found that the mobilization of Ta and Hf was closely linked to the occurrence of oxide spalling. These elements showed no migration from the test chamber. Some W was mobilized by volatilization as evidenced by transport from the chamber. Tungsten volatilization could occur primarily during initial stages of oxidation before an oxide scale forms and impedes the process. The mobilization of Ta and W are presented in terms of the mass flux (g/m 2 -h) as a function of test temperature. These measurements along with specific designs, activation calculations, and accident scenarios provide information useful for dose calculations of future fusion devices.

Contaminant ingress into multizone buildings: An analytical state-space approach

KAUST Repository

Parker, Simon; Coffey, Chris; Gravesen, Jens; Kirkpatrick, James; Ratcliffe, Keith; Lingard, Bryan; Nally, James

2013-01-01

The ingress of exterior contaminants into buildings is often assessed by treating the building interior as a single well-mixed space. Multizone modelling provides an alternative way of representing buildings that can estimate concentration time

Oxidation Resistant Graphite Studies

Energy Technology Data Exchange (ETDEWEB)

W. Windes; R. Smith

2014-07-01

The Very High Temperature Reactor (VHTR) Graphite Research and Development Program is investigating doped nuclear graphite grades exhibiting oxidation resistance. During a oxygen ingress accident the oxidation rates of the high temperature graphite core region would be extremely high resulting in significant structural damage to the core. Reducing the oxidation rate of the graphite core material would reduce the structural effects and keep the core integrity intact during any air-ingress accident. Oxidation testing of graphite doped with oxidation resistant material is being conducted to determine the extent of oxidation rate reduction. Nuclear grade graphite doped with varying levels of Boron-Carbide (B4C) was oxidized in air at nominal 740°C at 10/90% (air/He) and 100% air. The oxidation rates of the boronated and unboronated graphite grade were compared. With increasing boron-carbide content (up to 6 vol%) the oxidation rate was observed to have a 20 fold reduction from unboronated graphite. Visual inspection and uniformity of oxidation across the surface of the specimens were conducted. Future work to determine the remaining mechanical strength as well as graphite grades with SiC doped material are discussed.

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

Actinide nuclides in environmental air and precipitation samples after the Chernobyl accident

Energy Technology Data Exchange (ETDEWEB)

Rosner, G.; Hoetzl, H.; Winkler, R. (Gesellschaft fuer Strahlen- und Umweltforschung mbH Muenchen (West Germany))

1988-01-01

The present paper describes the analysis of isotopes of uranium, neptunium, plutonium, americium and curium, in air and deposition samples taken at our laboratory site 10 km north of Munich, subsequent to the Chernobyl accident. Uranium-234, {sup 237}U, {sup 238}U, {sup 239}Np, {sup 238}Pu, {sup 239+240}Pu and {sup 242}Cm have been identified and upper limits of detection have been established for {sup 241}Am and {sup 244}Cm. Deposition and air concentration values are discussed. 12 refs., 1 fig., 2 tabs.

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.)

A review of human factors causations in commercial air transport accidents and incidents: From to 2000-2016

Science.gov (United States)

Kharoufah, Husam; Murray, John; Baxter, Glenn; Wild, Graham

2018-05-01

Human factors have been defined by the International Civil Aviation Organization (ICAO) as "about people in their living and working situations; about their relationship with machines, with procedures and with the environment about them; and about their relationships with other people (at work)". Human factors contribute to approximately 75% of aircraft accidents and incidents. As such, understanding their influence is essential to improve safety in the aviation industry. This study examined the different human factors causations in a random sample of over 200 commercial air transport accidents and incidents from 2000 to 2016. The main objective of this study was to identify the principal human factor contributions to aviation accidents and incidents. An exploratory research design was utilised. The qualitative data were recorded in a database, and were coded into categories about the flights (including date, manufacturer, carrier, state of occurrence, etc). These categories were then analysed using Chi-Squared tests to determine which were statistically significant in terms of having an influence on the accidents/incidents. The most significant human factor was found to be situational awareness followed by non-adherence to procedures. In addition, charter operations proved to have a significantly higher rate of human factor related occurrence as compared to other type of operations. A significant finding was that Africa has a high rate of accidents/incidents relative to the amount of traffic and aircraft movements. These findings reflect some of the more noteworthy incidents that have received significant media attention, including Air Asia 8501 on the 28th of December 2014, TransAsia Airways 235 on the 4th of February 2015, and Air France 447 on the 1st of June 2009; these accidents resulted in a significant loss of lives where situational awareness and non-adherence to procedures were significant contributing factors.

Fission product release by fuel oxidation after water ingress

International Nuclear Information System (INIS)

Schreiber.

1990-01-01

On the basis of data obtained by a literature search, a computer code has been established for the calculation of the degree of oxidation of the fuel in the damaged fuel particles, and hence of the fission product release as a function of the time period of steam ingress. (orig.) [de

Technical Issues Associated with Air Ingression During Core Degradation

International Nuclear Information System (INIS)

Powers, Dana A.

2000-01-01

This paper has shown that it is possible to get significant air intrusion into a ruptured reactor vessel even from a reactor cavity with restricted access. This suggests that there is some importance to considering the consequences of air intrusion following vessel penetration by core debris. The consequences will depend on the nature of core degradation in air and other oxidizing gases. If, indeed, fuel becomes exposed to strongly oxidizing gases, significant releases of ruthenium and hexavalent urania can be expected. Hexavalent urania could alter the nature of cesium release and cesium revaporization from the reactor coolant system. Hexavalent urania could destabilize CSI and enhance the formation of gaseous iodine unless there are other materials that will react readily with atomic iodine along the flow path to the reactor containment

Portable Filtered Air Suction System for Released Radioactive Gases Prevention under a Severe Accident of NPPs

International Nuclear Information System (INIS)

Gu, Beom W.; Choi, Su Y.; Rim, Chun T.

2013-01-01

In this paper, the portable filtered air suction system (PoFASS) for released radioactive gases prevention under a severe accident of NPP is proposed. This technology can prevent the release of the radioactive gases to the atmosphere and it can be more economical than FVCS because PoFASS can cover many NPPs with its high mobility. The conceptual design of PoFASS, which has the highest cost effectiveness and robustness to the environment condition such as wind velocity and precipitation, is suggested and the related previous research is introduced in this paper. The portable filtered air suction system (PoFASS) for released radioactive gases prevention can play a key role to mitigate the severe accident of NPP with its high cost effectiveness and robustness to the environment conditions. As further works, the detail design of PoFASS to fabricate a prototype for a demonstration will be proceeded. When released radioactive gases from the broken containment building in the severe accident of nuclear power plants (NPPs) such as the Chernobyl and Fukushima accidents occur, there are no ways to prevent the released radioactive gases spreading in the air. In order to solve this problem, several European NPPs have adopted the filtered vented containment system (FVCS), which can avoid the containment failure through a pressure relief capability to protect the containment building against overpressure. However, the installation cost of FVCS for a NPP is more than $10 million and this system has not been widely welcomed by NPP operating companies due to its high cost

Intensity of the absorbing dose of the gamma rays in the air of Krusevac during and after nuclear accident in Chernobil

International Nuclear Information System (INIS)

Fortuna, D.

1997-01-01

In this paper are compared overage daily values of the intensity of the absorbing doses of the gamma rays in the air of Krusevac, during and after nuclear accident in Cernobil. Average daily values of intensity of the absorbing doses of gamma rays in the air of Krusevac, immediately after nuclear accident in Cernobil were, three to seven time higher than of the average daily values of the natural rays. (author)

Effects of Oxidation on Oxidation-Resistant Graphite

Energy Technology Data Exchange (ETDEWEB)

Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Rebecca [Idaho National Lab. (INL), Idaho Falls, ID (United States); Carroll, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-05-01

The Advanced Reactor Technology (ART) Graphite Research and Development Program is investigating doped nuclear graphite grades that exhibit oxidation resistance through the formation of protective oxides on the surface of the graphite material. In the unlikely event of an oxygen ingress accident, graphite components within the VHTR core region are anticipated to oxidize so long as the oxygen continues to enter the hot core region and the core temperatures remain above 400°C. For the most serious air-ingress accident which persists over several hours or days the continued oxidation can result in significant structural damage to the core. Reducing the oxidation rate of the graphite core material during any air-ingress accident would mitigate the structural effects and keep the core intact. Previous air oxidation testing of nuclear-grade graphite doped with varying levels of boron-carbide (B4C) at a nominal 739°C was conducted for a limited number of doped specimens demonstrating a dramatic reduction in oxidation rate for the boronated graphite grade. This report summarizes the conclusions from this small scoping study by determining the effects of oxidation on the mechanical strength resulting from oxidation of boronated and unboronated graphite to a 10% mass loss level. While the B4C additive did reduce mechanical strength loss during oxidation, adding B4C dopants to a level of 3.5% or more reduced the as-fabricated compressive strength nearly 50%. This effectively minimized any benefits realized from the protective film formed on the boronated grades. Future work to infuse different graphite grades with silicon- and boron-doped material as a post-machining conditioning step for nuclear components is discussed as a potential solution for these challenges in this report.

Probabilistic Models and Computational Methods for Chloride Ingress in Concrete

DEFF Research Database (Denmark)

Engelund, S.

Within the last decades it has been recognized that reinforced concrete structures are subject to a number of destructive mechanisms which might affect the structure in such a way that it is not able to fulfil its purpose efficiently. The present report focuses on chloride ingress and chloride...

Monitoring of the vegetable garden open air spinach and soil radioactivity in Beijing during the Japan Fukushima nuclear accident

International Nuclear Information System (INIS)

Zhou Qiang; Xu Cuihua; Tuo Fei; Zhang Jing; Li Wenhong; Zhang Qing; Zhang Weijun; Zhang Jianfeng; Su Xu

2012-01-01

Objective: To detect artificial radionuclide content in the spinach and soil in open air vegetable garden in Beijing during Fukushima nuclear accident and to study the radioactive contamination characteristics of the samples. Methods: 6 spinach samples and 3 soil samples in open air vegetable garden were obtained through continuously sampling. High-purity germanium (HPGe) γ spectrometry was used to analyze activity concentrations of artificial radionuclide 131 I etc in these samples. Results: Artificial radionuclide 131 I was detected in the 6 open air spinach samples. Artificial radionuclide 137 Cs was detected in 3 vegetable garden soil samples, trace amount of 131 I was detected in 1 open-air surface soil sample alone. Conclusions: Following the Fukushima nuclear accident, spinach in Beijing's open-air vegetable garden was slightly polluted by artificial radionuclide 131 I, with the highest values of pollution appeared around April 4, 2011, but it could not cause harm to human health. (authors)

Characteristic behaviour of Pebble Bed High Temperature Gas-cooled Reactors during water ingress events

International Nuclear Information System (INIS)

Khoza, Samukelisiwe N.; Serfontein, Dawid E.; Reitsma, Frederik

2014-01-01

The presence of water on the tube-side of the steam generators in high temperature gas-cooled reactors (HTGRs) with indirect cycle layouts presents a possibility for a penetration of neutron moderating steam into the core, which may cause a power excursion. This article presents results on the effect of water ingress into the core of the two South African Pebble Bed Modular Reactor design concepts, i.e. the PBMR-200 MW th and the PBMR-400 MW th developed by PBMR SOC Ltd. The VSOP 99/05 suite of codes was used for the simulation of this event. Partial steam vapour pressures were added in stages into the primary circuit in order to investigate the effect of water ingress on reactivity, power profiles and thermal neutron flux profiles. The effects of water ingress into the core are explained by increased neutron moderation, due to the addition of 1 H, which leads to a decrease in resonance capture by 238 U and therefore an increase in the multiplication factor. The more effective moderation of neutrons by definition reduces the fast neutron flux and increases the thermal flux in the core, i.e. leads to a softer spectrum. The more effective moderation also increases the average increase in lethargy between collisions of a neutron with successive fuel kernels, which reduces the probability for neutron capture in the radiative capture resonances of 238 U. The resulting higher resonance escape probability also increases the thermal flux in the core. The softening of the neutron spectrum leads to an increased effective microscopic fission cross section in the fissile isotopes and thus to increased neutron absorption for fission, which reduces the remaining number of neutrons that can diffuse into the reflectors. Therefore water ingress into the core leads to a reduced thermal neutron flux in the reflectors. The power density spatial distribution behaved similarly to the thermal neutron flux in the core. Analysis of possible mechanisms was conducted. The results show that

Radionuclides from the Fukushima accident in the air over Lithuania: measurement and modelling approaches.

Science.gov (United States)

Lujanienė, G; Byčenkienė, S; Povinec, P P; Gera, M

2012-12-01

Analyses of (131)I, (137)Cs and (134)Cs in airborne aerosols were carried out in daily samples in Vilnius, Lithuania after the Fukushima accident during the period of March-April, 2011. The activity concentrations of (131)I and (137)Cs ranged from 12 μBq/m(3) and 1.4 μBq/m(3) to 3700 μBq/m(3) and 1040 μBq/m(3), respectively. The activity concentration of (239,240)Pu in one aerosol sample collected from 23 March to 15 April, 2011 was found to be 44.5 nBq/m(3). The two maxima found in radionuclide concentrations were related to complicated long-range air mass transport from Japan across the Pacific, the North America and the Atlantic Ocean to Central Europe as indicated by modelling. HYSPLIT backward trajectories and meteorological data were applied for interpretation of activity variations of measured radionuclides observed at the site of investigation. (7)Be and (212)Pb activity concentrations and their ratios were used as tracers of vertical transport of air masses. Fukushima data were compared with the data obtained during the Chernobyl accident and in the post Chernobyl period. The activity concentrations of (131)I and (137)Cs were found to be by 4 orders of magnitude lower as compared to the Chernobyl accident. The activity ratio of (134)Cs/(137)Cs was around 1 with small variations only. The activity ratio of (238)Pu/(239,240)Pu in the aerosol sample was 1.2, indicating a presence of the spent fuel of different origin than that of the Chernobyl accident. Copyright © 2011 Elsevier Ltd. All rights reserved.

Durability testing of low clinker blends - chloride ingress in similar strength mortar exposed to seawater

DEFF Research Database (Denmark)

Geiker, Mette Rica; De Weerdt, Klaartje; Garzón, Sergio Ferreiro

2017-01-01

Resistance to chloride ingress of ten different binders was investigated. Most of the binders were prepared with 35% substitution of a new clinker by limestone filler, calcined clay, burnt shale and/or siliceous fly ash. Mortar samples with similar design compressive strength after 90 days were e...... exposed to artificial sea-water for 270 days. The results indicate that the use of alternative binders may lead to up to around 15% reduction in CO2 emission without compromising 90 days compressive strength and resistance to chloride ingress in marine exposure....

SARNET integrated European Severe Accident Research-Conclusions in the source term area

Energy Technology Data Exchange (ETDEWEB)

Haste, T., E-mail: tim.haste@irsn.f [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Giordano, P. [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France); Herranz, L. [Centro de Investigaciones Energeticas Medio Ambientales y Tecnologica, CIEMAT, Avda. Complutense 22, E-28040 Madrid (Spain); Girault, N.; Dubourg, R. [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France); Sabroux, J.-C. [Institut de Radioprotection et de Surete Nucleaire, IRSN, Saclay Research Centre, BP 68, F-91192 Gif-sur-Yvette Cedex (France); Cantrel, L. [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France); Bottomley, D. [European Commission Joint Research Centre, Transuranium Institute, P.O. Box 2340, D-76125 Karlsruhe (Germany); Parozzi, F. [ENEA - Ricerca sul Sistema Elettrico (ERSE) SpA., Via Rubattino 54, I-20134 Milano (Italy); Auvinen, A. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT Espoo (Finland); Dickinson, S. [National Nuclear Laboratory, Harwell Business Centre, Didcot, OX11 0QJ (United Kingdom); Lamy, J.-C. [Electricite de France, 12-14 avenue Dutrievoz, F-69100 Villeurbanne (France); Weber, G. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsgelaende, D-85748 Garching (Germany); Albiol, T. [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France)

2009-12-15

The overall aim of the SARNET (Severe Accident Research NETwork), in the EU 6th Framework programme was to integrate in a sustainable manner the research capabilities of fifty-one European organisations from eighteen member states of the European Union (EU) plus the Joint Research Centres, with one Canadian company, to resolve important remaining uncertainties and safety issues concerning existing and future nuclear plant, especially water-cooled reactors, under hypothetical severe accident conditions. It emphasised integrating activities, spreading of excellence (including knowledge transfer) and jointly executed research, with the knowledge gained being encapsulated in the European severe accident modelling code ASTEC. This paper summarises the achievements over the whole project in the Source Term Topic, which dealt with potential radioactive release to the environment, covering release of fission products and structural materials from the core, their transport in the primary circuit, and their behaviour in the containment. The main technical areas covered, as emphasised by the earlier EURSAFE project, were the effect of oxidative conditions on fission product release and transport (especially the behaviour of the highly radiotoxic ruthenium under air ingress conditions), iodine volatility in the primary circuit, control rod aerosol release (Ag-In-Cd) that affects iodine transport, containment by-pass in the case of steam generator tube rupture, aerosol retention in containment cracks, aerosol remobilisation in the circuit, and iodine/ruthenium behaviour in the containment especially concerning the volatile fraction in the atmosphere. The studies also covered performance of new experiments, analysis of existing data, and formulation and improvement of theoretical models. Significant progress was made in each area. Looking to the future, the 7th Framework successor project SARNET2 covers the remaining issues concerning iodine and ruthenium, including practical

Oxygen ingress study of 3D printed gaseous radiation detector enclosures

Energy Technology Data Exchange (ETDEWEB)

Steer, Christopher A.; Durose, Aaron [AWE, Alderrnaston, Reading, Berkshire, RG7 4PR (United Kingdom)

2015-07-01

As part of our ongoing studies into the potential application of 3D printing techniques to gaseous radiation detectors, we have studied the ability of 3D printed enclosures to resist environmental oxygen ingress. A set of cuboid and hexagonal prism shaped enclosures with wall thicknesses of 4 mm, 6 mm, 8 mm and 10 mm were designed and printed in nylon using a EOSINT P 730 Selective Laser Sintering 3D printer system These test enclosures provide a comparison of different environmental gas ingress for different 3D printing techniques. The rate of change of oxygen concentration was found to be linear, decreasing as the wall thickness increases. It was also found that the hexagonal prism geometry produced a lower rate of change of oxygen concentration compared with the cuboid shaped enclosures. Possible reasons as to why these results were obtained are discussed The implications for the this study for deployable systems are also discussed (authors)

Oxygen ingress study of 3D printed gaseous radiation detector enclosures

International Nuclear Information System (INIS)

Steer, Christopher A.; Durose, Aaron

2015-01-01

As part of our ongoing studies into the potential application of 3D printing techniques to gaseous radiation detectors, we have studied the ability of 3D printed enclosures to resist environmental oxygen ingress. A set of cuboid and hexagonal prism shaped enclosures with wall thicknesses of 4 mm, 6 mm, 8 mm and 10 mm were designed and printed in nylon using a EOSINT P 730 Selective Laser Sintering 3D printer system These test enclosures provide a comparison of different environmental gas ingress for different 3D printing techniques. The rate of change of oxygen concentration was found to be linear, decreasing as the wall thickness increases. It was also found that the hexagonal prism geometry produced a lower rate of change of oxygen concentration compared with the cuboid shaped enclosures. Possible reasons as to why these results were obtained are discussed The implications for the this study for deployable systems are also discussed (authors)

Summary of oxidation driven mobilization data and their use in fusion safety assessments

International Nuclear Information System (INIS)

McCarthy, K.A.; Smolik, G.R.; Hagrman, D.L.; Petti, D.A.

1996-01-01

We have carried out experiments to simulate accident conditions with air or steam ingress, and determine the amount of material mobilized from the material. We also perform first principles modeling to understand the mechanisms involved in mobilization, and determine whether volatilization or oxide spalling dominates mobilization. Our results indicate that if long-term accident temperatures are kept below ∝700 C, oxidation-driven mobilization may be less important than resuspension of tokamak dust, release of corrosion products, and release of plasma-vaporized material. (orig.)

Moisture ingress into electronics enclosures under isothermal conditions

DEFF Research Database (Denmark)

Staliulionis, Zygimantas; Jabbaribehnam, Mirmasoud; Hattel, Jesper Henri

2016-01-01

loads are still not understood well by design engineers, therefore this field has become one of the bottlenecks in the electronics system design. The objective of this paper is to model moisture ingress into an electronics enclosure under isothermal conditions. The moisture diffusion model is based......The number of electronics used in outdoor environment is constantly growing. The humidity causes about 19 % of all electronics failures and, especially, moisture increases these problems due to the ongoing process of miniaturization and lower power consumption of electronic components. Moisture...

Chloride ingress in cracked concrete : A laser induced breakdown spectroscopy (LIBS) study

NARCIS (Netherlands)

Savija, B.; Schlangen, E.; Pacheco Farias, J.; Millar, S.; Eichler, T.; Wilsch, G.

2014-01-01

racks are always present in reinforced concrete structures. In the presented research, influence of mechanical cracks on chloride ingress is studied. A compact reinforced concrete specimen was designed, mimicking the cracking behaviour of beam elements. Cracks of different widths were induced by

Simulation of lumbar and neck angle flexion while ingress of paratransit (angkot in Indonesia as a preliminary design study

Directory of Open Access Journals (Sweden)

Yukhi Mustaqim Kusuma Sya’bana

2017-12-01

Full Text Available This is the preliminary finding of a study to simulate lumbar and neck flexion while ingress to the paratransit. The result of simulation will determine design aspect criteria as a preliminary step before ideation and implementation design steps. Biomechanics of Bodies (BoB is software that used to represent passenger task during paratransit ingress simulation, with skeleton model that used is height 165 cm and weight 65 kg. Environment to represent this simulation is measured Suzuki Carry SS 2013 as a private car that has been modified into a public transportation in accordance with the Indonesian government road-worthy test. Due to the low height of the entrance and the high ground clearance, lumbar and neck joint angle was a focus of this ingress simulation. The peak angle at the neck joint is 40° when 2 s skeleton nod in the door limitation ingress and lumbar flexion is 70° when 5 s skeleton is walking while bend over that will increase the load on that area. Based on biomechanical simulation approach, we may suggest the dimension of public transportation design framework developments, especially paratransit.

Effect of steam corrosion on core post strength loss: I. Low, chronic steam ingress rates

International Nuclear Information System (INIS)

Wichner, R.P.

1976-10-01

The purpose of the study was to assess the effect of chronic, low levels of steam ingress into the primary system of the HTGR on the corrosion, and consequent strength loss of the core support posts. The assessment proceeded through the following three steps: (1) The impurity composition in the primary system was estimated as a function of a range of steady ingress rates of from 0.001 to 1.0 g/sec, both by means of an analysis of the Dragon steam ingress experiment and a computer code, TIMOX, which treats the primary system as a well-mixed pot. (2) The core post burnoffs which result from 40-year exposures to these determined impurity atmospheres were then estimated using a corrosion rate expression derived from published ATJ-graphite corrosion rate data. Burnoffs were determined for both the core posts at the nominal and the maximum sustained temperature, estimated to be 90 0 C above nominal. (3) The final step involved assessment of the degree of strength loss resulting from the estimated burnoffs. An empirical equation was developed for this purpose which compares reasonably well with strength loss data for a number of different graphites and specimen geometries

Deuterium ingress at rolled joints in Embalse nuclear power plant

International Nuclear Information System (INIS)

Ramos Nervi, J. E.; Schroeter, F.

2013-01-01

Deuterium ingress model at the Rolled Joint has been extensively used for CANDU Nuclear Power Plants Operators in the Life Management of the Pressure Tubes. The importance of understanding the model is vital to avoid delayed hydride cracking at the Rolled Joint. This work reports the first step on develop the model presented on literature to be used in Argentinean CANDU 6, Embalse Nuclear Power Plant. (author)

Radiation safety of the helicopter air crow-liquidators of the Chernobyl NPP accident

International Nuclear Information System (INIS)

Davydov, B.I.; Ponomarenko, V.A.; Ushakov, I.B.

1992-01-01

Three aspects of the radiation safety of the helicopter air crew who took part in the elimination of the Chernobyl NPP accident after-effects during the period from the 27-th of April to July 1986 were considered: dosimetry, physical protection and radioprotector use. 1125 subjects from the flight and engineer personnel were irradiated: 6%-2.5 mSv and more, 32%-2.1-2.5 mSv and about 60% - up to 2.0 mSv. It is stated that the use of physical protection (local screening of a pilot seat and use of special antiradiation belts) makes irradiation doe 2-3 times less. In the intense conditions of flight activity the pilots preferred the radioprotector adrenaline to cystamin. Alimentary aids in the earlier and later periods after the accident influenced favourably somatic and psychic states. 16 refs.; 1 tab

Landing Distance Minimization to Prevent Overrun Accidents Using Field Theory and Stabilizing Air Traffic - A Novel Approach

Science.gov (United States)

Krishna Kumar, R.; Navaneeth, M.; Shachin Shibi, R.

2017-09-01

Airplane is considered to be the pinnacle of engineering as it has proven that it is possible for a manmade object to fly. Before its invention, flying was just a dream for mankind. In such an esteemed domain, landing is the most challenging part and it is where a large number of accidents occur, especially due to overrun. As the name suggests, overrun accidents occur due to insufficient runway length. In the present study, the concept of planar electromagnetic fields is incorporated to minimize the landing distance of an aircraft, thus preventing the overrun accidents. As a result, unexpected losses can be avoided. In addition to this, the stability of air traffic control can be perpetuated and the fuel consumed during loitering time can be diminished.

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

Vision-Aided Autonomous Landing and Ingress of Micro Aerial Vehicles

Science.gov (United States)

Brockers, Roland; Ma, Jeremy C.; Matthies, Larry H.; Bouffard, Patrick

2012-01-01

Micro aerial vehicles have limited sensor suites and computational power. For reconnaissance tasks and to conserve energy, these systems need the ability to autonomously land at vantage points or enter buildings (ingress). But for autonomous navigation, information is needed to identify and guide the vehicle to the target. Vision algorithms can provide egomotion estimation and target detection using input from cameras that are easy to include in miniature systems.

Monitoring Ingress of Moisture in Structural Concrete Using a Novel Optical-Based Sensor Approach

International Nuclear Information System (INIS)

Yeo, T L; Cox, M A C; Boswell, L F; Sun, T; Grattan, K T V

2006-01-01

The detection of moisture ingress in concrete is important for structural monitoring and in this work is realised by monitoring the shift in the characteristic wavelength of a fibre Bragg grating-based sensor. The sensor relies upon a moisture-sensitive polymer layer deposited on the fibre Bragg grating (FBG) and the strain induced on it as a result of polymer swelling is monitored. Moisture ingress experiments were carried out using two such optical fibre sensors, placed at varying distances from the edge of the face of standard concrete cubes to the inner part of the concrete sample and subjected to water at a constant temperature. Information on the properties of different types of concrete and thus potentially on the migration of dissolved salts and their effect on reinforcement bars within concrete can be obtained

Radioactive contamination from Chernobyl accident over Alexandria city

International Nuclear Information System (INIS)

Ammar, E.A.; El-Khatib, A.M.; Wahba, A.G.; Elraey, M.

1987-01-01

The concentration of radioactive contamination in air resulting from the Chernobyl accident has been followed up. A sudden and sharp increase was detected seven days after the start of the accident. This increase amounted to about 650 times the normal air-borne activity. (author)

A software tool for evaluation of hydrogen ingress in CANDU pressure tubes

International Nuclear Information System (INIS)

Mihalache, Maria; Vasile, Radu; Deaconu, Mariea

2009-01-01

The prediction of hydrogen isotopes concentration into the body and in the rolled joints of operating pressure tubes as a function of reactor hot hours is very important in many fitness-for-service assessments and end of life estimates. The rolled joints are high stress zones with potential for delayed hydride cracking. Predictive models for assessing the long-term deuterium ingress in both body and rolled joint of the pressure tubes have been implemented in a software tool, ROHID, developed in INR-Pitesti. ROHID is a PC-based Windows application with a user-friendly interface that predicts the equivalent hydrogen ingress for Zr-2.5Nb pressure tubes. It uses colour-coded reactor core maps to display the predicted deuterium concentration as a function of time for selected axial locations. Plots of deuterium versus axial location and time for individual pressure tubes are also available. Also, the software tool can predict the exceeding of hydrogen terminal solid solubility (HTSS) from hydrides during precipitation and dissolving processes as a function of time and axial location. (authors)

Evaluation of ATP measurements to detect microbial ingress by wastewater and surface water in drinking water.

Science.gov (United States)

Vang, Óluva K; Corfitzen, Charlotte B; Smith, Christian; Albrechtsen, Hans-Jørgen

2014-11-01

Fast and reliable methods are required for monitoring of microbial drinking water quality in order to protect public health. Adenosine triphosphate (ATP) was investigated as a potential real-time parameter for detecting microbial ingress in drinking water contaminated with wastewater or surface water. To investigate the ability of the ATP assay in detecting different contamination types, the contaminant was diluted with non-chlorinated drinking water. Wastewater, diluted at 10(4) in drinking water, was detected with the ATP assay, as well as 10(2) to 10(3) times diluted surface water. To improve the performance of the ATP assay in detecting microbial ingress in drinking water, different approaches were investigated, i.e. quantifying microbial ATP or applying reagents of different sensitivities to reduce measurement variations; however, none of these approaches contributed significantly in this respect. Compared to traditional microbiological methods, the ATP assay could detect wastewater and surface water in drinking water to a higher degree than total direct counts (TDCs), while both heterotrophic plate counts (HPC 22 °C and HPC 37 °C) and Colilert-18 (Escherichia coli and coliforms) were more sensitive than the ATP measurements, though with much longer response times. Continuous sampling combined with ATP measurements displays definite monitoring potential for microbial drinking water quality, since microbial ingress in drinking water can be detected in real-time with ATP measurements. The ability of the ATP assay to detect microbial ingress is influenced by both the ATP load from the contaminant itself and the ATP concentration in the specific drinking water. Consequently, a low ATP concentration of the specific drinking water facilitates a better detection of a potential contamination of the water supply with the ATP assay. Copyright © 2014 Elsevier Ltd. All rights reserved.

Safety regulations regarding to accident monitoring and accident sampling at Russian NPPs with VVER type reactors

International Nuclear Information System (INIS)

Sharafutdinov, Rachet; Lankin, Michail; Kharitonova, Nataliya

2014-01-01

The paper describes a tendency by development of regulatory document requirements related to accident monitoring and accident sampling at Russia's NPPs. Lessons learned from the Fukushima Daiichi accident pointed at the importance and necessary to carry out an additional safety check at Russia's nuclear power plants in the preparedness for management of severe accidents at NPPs. Planned measures for improvement of severe accidents management include development and implementation of the accident instrumentation systems, providing, monitoring, management and storage of information in a severe accident conditions. The draft of Safety Guidelines <accident monitoring system of nuclear power plants with VVER reactors' prepared by Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) established the main criteria for accident monitoring instrumentation that can monitor relevant plant parameters in the reactor and inside containment during and after a severe accident in nuclear power plants. Development of these safety guidelines is in line with the recommendations of IAEA Action Plan on Nuclear Safety in response to the Fukushima Daiichi event and recommendations of the IAEA Nuclear Energy series Report <<Accident Monitoring Systems for Nuclear Power Plants' (Draft V 2.7). The paper presents the principles, which are used as the basis for selection of plant parameters for accident monitoring and for establishing of accident monitoring instrumentation. The recommendations to the accident sampling system capable to obtain the representative reactor coolant and containment air and fluid samples that support accurate analytical results for the parameters of interest are considered. The radiological and chemistry parameters to be monitored for primary coolant and sump and for containment air are specified. (author)

Reducing the ingress of urban noise through natural ventilation openings.

Science.gov (United States)

Oldham, D J; de Salis, M H; Sharples, S

2004-01-01

For buildings in busy urban areas affected by high levels of road traffic noise the potential to use natural ventilation can be limited by excessive noise entering through ventilation openings. This paper is concerned with techniques to reduce noise ingress into naturally ventilated buildings while minimizing airflow path resistance. A combined experimental and theoretical approach to the interaction of airflow and sound transmission through ventilators for natural ventilation applications is described. A key element of the investigation has been the development of testing facilities capable of measuring the airflow and sound transmission losses for a range of ventilation noise control strategies. It is demonstrated that a combination of sound reduction mechanisms -- one covering low frequency sound and another covering high frequency sound -- is required to attenuate effectively noise from typical urban sources. A method is proposed for quantifying the acoustic performance of different strategies to enable comparisons and informed decisions to be made leading to the possibility of a design methodology for optimizing the ventilation and acoustic performance of different strategies. The need for employing techniques for combating low frequency sound in tandem with techniques for reducing high frequency sound in reducing the ingress of noise from urban sources such as road traffic to acceptable levels is demonstrated. A technique is proposed for enabling the acoustic and airflow performance of apertures for natural ventilation systems to be designed simultaneously.

Influence of temporal resolution and processing of exposure data on modeling of chloride ingress and reinforcement corrosion in concrete

DEFF Research Database (Denmark)

Flint, Madeleine; Michel, Alexander; Billington, Sarah L.

2014-01-01

a numerical heat and mass transport model that includes full coupling of heat, moisture and ion transport. Heat, moisture, and chloride concentration distributions were passed to a simplified reinforcement corrosion initiation and propagation model. The numerical study indicates that processing and temporal...... resolution of the exposure data has a considerable impact on long-term hygrothermal distribution, chloride ingress, and reinforcement section loss results. Use of time-averaged exposure data in the heat and mass transport model reduces the rate of chloride ingress in concrete and affects prediction...

Operating experience of a sodium rig following the ingress of oil

Energy Technology Data Exchange (ETDEWEB)

Bell, A C; Mehew, R D; Robertson, C M [UKAEA, Dounreay, Thurso, Caithness, Scotland (United Kingdom)

1980-05-01

The experience of operating the Small Water Leak Rig at Dounreay in the two years following the ingress into the sodium of silicone oil from the pump is reported. The carbon penetration into the rig pipework has been monitored and has been found to agree with a diffusion model. The predictions based on this model have enabled an operating strategy to be determined to allow continued operation of the rig. (author)

Operating experience of a sodium rig following the ingress of oil

International Nuclear Information System (INIS)

Bell, A.C.; Mehew, R.D.; Robertson, C.M.

1980-01-01

The experience of operating the Small Water Leak Rig at Dounreay in the two years following the ingress into the sodium of silicone oil from the pump is reported. The carbon penetration into the rig pipework has been monitored and has been found to agree with a diffusion model. The predictions based on this model have enabled an operating strategy to be determined to allow continued operation of the rig. (author)

A separate effect study of the influence of metallic fission products on CsI radioactive release from nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Di Lemma, F.G., E-mail: fidelma.dilemma@gmail.com [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Department of Radiation Science and Technology, Faculty of Applied Sciences, Delft University of Technology, Delft, 2629 JB (Netherlands); Colle, J.Y., E-mail: jean-yves.colle@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Beneš, O. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Konings, R.J.M. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Department of Radiation Science and Technology, Faculty of Applied Sciences, Delft University of Technology, Delft, 2629 JB (Netherlands)

2015-10-15

The chemistry of cesium and iodine is of main importance to quantify the radioactive release in case of a nuclear reactor accident, or sabotage involving irradiated nuclear materials. We studied the interaction of CsI with different metallic fission products such as Mo and Ru. These elements can be released from nuclear fuel when exposed to oxidising conditions, as in the case of contact of overheated nuclear fuel with air (e.g. in a spent fuel cask sabotage, uncovering of a spent fuel pond, or air ingress accidents). Experiments were performed by vaporizing mixtures of the compounds in air, and analysing the produced aerosols in view of a possible gas–gas and gas–aerosol reactions between the compounds. These results were compared with the gaseous species predicted by thermochemical equilibrium calculations and experimental equilibrium vaporization tests using Knudsen Effusion Mass Spectrometry.

SR-Site: Oxygen ingress in the rock at Forsmark during a glacial cycle

Energy Technology Data Exchange (ETDEWEB)

Sidborn, Magnus (Kemakta Konsult AB (Sweden)); Sandstroem, Bjoern (WSP Sverige AB (Sweden)); Tullborg, Eva-Lena (Terralogica AB (Sweden)); Salas, Joaquin; Maia, Flavia; Delos, Anne; Molinero, Jorge (Amphos21 (Spain)); Hallbeck, Lotta; Pedersen, Karsten (Microbial Analytics Sweden AB (Sweden))

2010-11-15

The aim of this report is to assess the possibility for oxygen to be transported by glacial melt-water to canister positions in a final repository for spent nuclear fuel at the proposed location in Forsmark. The approach for this assessment is to combine reactive transport modelling with geological observations of present and historical indications of oxygen ingress. For safety assessment purposes a cautious approach in the modelling is required when estimating the extent of oxygen ingress. In this report, a cautious approach has been applied both in the conceptualisation of the problem and in the choice of input parameters used in the models. Oxygen consuming processes are only neglected in the modelling if they are expected to further decrease the extent of oxygen ingress. Several oxygen consuming processes have been identified, each of which may play an important role in the scavenging of oxygen along recharge flow paths in the rock. These processes include biological pathways with degradation of organic material of ground surface origin, and biotically mediated reactions with reduced rock minerals and with various materials expected to be present in the backfilled repository volume. In the absence of microbes most of these reactions may also follow abiotic pathways. Present day observations show that degradation of organic material is the most powerful oxygen scavenging process. At Forsmark, oxygen is generally depleted within a few metres under present day temperate conditions. Although biological activity is likely to exist also during different phases of a glaciation, large uncertainties exist regarding e.g. the population growth dynamics, the biotic reaction rates and the availability of organic material under the highly varying conditions expected. Microbial activity and degradation of organic material is therefore pessimistically neglected in the calculations in this report. In the absence of organic material, ferrous iron present in minerals in the rock

SR-Site: Oxygen ingress in the rock at Forsmark during a glacial cycle

International Nuclear Information System (INIS)

Sidborn, Magnus; Sandstroem, Bjoern; Tullborg, Eva-Lena; Salas, Joaquin; Maia, Flavia; Delos, Anne; Molinero, Jorge; Hallbeck, Lotta; Pedersen, Karsten

2010-11-01

The aim of this report is to assess the possibility for oxygen to be transported by glacial melt-water to canister positions in a final repository for spent nuclear fuel at the proposed location in Forsmark. The approach for this assessment is to combine reactive transport modelling with geological observations of present and historical indications of oxygen ingress. For safety assessment purposes a cautious approach in the modelling is required when estimating the extent of oxygen ingress. In this report, a cautious approach has been applied both in the conceptualisation of the problem and in the choice of input parameters used in the models. Oxygen consuming processes are only neglected in the modelling if they are expected to further decrease the extent of oxygen ingress. Several oxygen consuming processes have been identified, each of which may play an important role in the scavenging of oxygen along recharge flow paths in the rock. These processes include biological pathways with degradation of organic material of ground surface origin, and biotically mediated reactions with reduced rock minerals and with various materials expected to be present in the backfilled repository volume. In the absence of microbes most of these reactions may also follow abiotic pathways. Present day observations show that degradation of organic material is the most powerful oxygen scavenging process. At Forsmark, oxygen is generally depleted within a few metres under present day temperate conditions. Although biological activity is likely to exist also during different phases of a glaciation, large uncertainties exist regarding e.g. the population growth dynamics, the biotic reaction rates and the availability of organic material under the highly varying conditions expected. Microbial activity and degradation of organic material is therefore pessimistically neglected in the calculations in this report. In the absence of organic material, ferrous iron present in minerals in the rock

Design of a power conversion system for an indirect cycle, helium cooled pebble bed reactor system

International Nuclear Information System (INIS)

Wang, C.; Ballinger, R.G.; Stahle, P.W.; Demetri, E.; Koronowski, M.

2002-01-01

A design is presented for the turbomachinery for an indirect cycle, closed, helium cooled modular pebble bed reactor system. The design makes use of current technology and will operate with an overall efficiency of 45%. The design uses an intermediate heat exchanger which isolated the reactor cycle from the turbomachinery. This design excludes radioactive fission products from the turbomachinery. This minimizes the probability of an air ingress accident and greatly simplifies maintenance. (author)

40 CFR 68.168 - Five-year accident history.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Five-year accident history. 68.168 Section 68.168 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Risk Management Plan § 68.168 Five-year accident history...

The Chernobyl accident

International Nuclear Information System (INIS)

Berg, J.O.; Christensen, G.; Lingjaerde, R.; Smidt Olsen, H.; Wethe, P.I.

1986-10-01

In connection with the Chernobyl accident the report gives a description of the technical features of importance to the accident, the course of events, and the estimated health hazards in the local environment. Dissimilarities in western and Sovjet reactor safety philosophy are dealt with, as well as conceivable concequences in relation to technology and research in western nuclear power programmes. Results of activity level measurements of air and foodstuff, made in Norway by Institute for Energy Technology, are given

Pressurizing Behavior on Ingress of Coolant into Pebble Bed of Blanket of Fusion DEMO Reactor

International Nuclear Information System (INIS)

Daigo Tsuru; Mikio Enoeda; Masato Akiba

2006-01-01

Solid breeder blankets are being developed as candidate blankets for the Fusion DEMO reactor in Japan. JAEA is performing the development of the water cooled and helium cooled solid breeder blankets. The blanket utilizes ceramic breeder pebbles and multiplier pebbles beds cooled by high pressure water or high pressure helium in the cooling tubes placed in the blanket box structure. In the development of the blanket, it is very important to incorporate the safety technology as well as the performance improvement on tritium production and energy conversion. In the safety design and technology, coolant ingress in the blanket box structure is one of the most important events as the initiators. Especially the thermal hydraulics in the pebble bed in the case of the high pressure coolant ingress is very important to evaluate the pressure propagation and coolant flow behavior. This paper presents the preliminary results of the pressure loss characteristics by the coolant ingress in the pebble bed. Experiments have been performed by using alumina pebble bed (4 litter maximum volume of the pebble bed) and nitrogen gas to simulate the helium coolant ingress into breeder and multiplier pebble beds. Reservoir tank of 10 liter is filled with 1.0 MPa nitrogen. The nitrogen gas is released at the bottom part of the alumina pebble bed whose upper part is open to the atmosphere. The pressure change in the pebble bed is measured to identify the pressure loss. The measured values are compared with the predicted values by Ergun's equation, which is the correlation equation on pressure loss of the flow through porous medium. By the results of the experiments with no constraint on the alumina pebble bed, it was clarified that the measured value agreed in the lower flow rate. However, in the higher flow rate where the pressure loss is high, the measured value is about half of the predicted value. The differences between the measured values and the predicted values will be discussed from

Effects of the Air Flow Rate on The Oxidation of NBG-18 and 25 Nuclear Graphite Grades

International Nuclear Information System (INIS)

Chi, Se-Hwan; Kim, Gen-Chan; Jang, Joon-Hee

2007-01-01

For a VHTR, graphite oxidation is regarded as a critical phenomenon for degrading the integrity of graphite components under normal or abnormal conditions. The oxidation of a graphite core component can occur by air which may permeate into the primary coolant operation and/or by impurities contained in the He coolant, or by air ingress during a severe accident. It is well known that the oxidation properties of a graphite are highly dependent on the source of raw materials, impurities, microstructures (crystallites, pore structure), and on the processing and environmental parameters, such as the forming methods, the coolant type, moisture and impurity content, temperature, flow rate and the oxygen potential of the coolants. A lot of work has been performed on the oxidation of graphite since the 1960s, and, for example, in the case of the temperature, a widely accepted oxidation model on the effects of a temperature has already been developed. However, in the case of the flow rate, even for its expected effects in a VHTR, for example, as to the expected changes in the bypass flow (10-20 %) during an operation, no systematic works have been performed. In this respect, as a preliminary study, the effects of an air flow rate on the oxidation of NBG-18 and 25 nuclear graphite were investigated

The major regularities of the air radioactive contamination of Belarus territory after the Chernobyl accident

International Nuclear Information System (INIS)

Konoplya, E. F.; Mironov, V. P.; Drugachenok, M. A.; Kudryashov, V.P; Grushevich, L.E; Adamovich, A.A.

2001-01-01

In the first days after the Chernobyl accident the radioactivity of surface air in different regions of Belarus has increased in tens and hundred thousand time. The regular control of air radioactive contamination in the zone of alienation and nearest to it is carried out from the end 1989. The radioactive air monitoring carried out in territories with various contamination density (from 0,2 up to 9,6 MBq/m 2 on Cs-137). The measurements of the Cs-137 contents in ashes of air sampler filters were carried out by gamma - spectrometer ADCAM-300 MCA. The measurement of the Pu-240,239 and Sr-90 contents was carried out by usual radiochemical techniques. The used techniques and equipment allow to define in a sample of the minimal activity: Pu-239,240 - 0,001 Bq, Cs-137 - 0,5 Bq, Sr-90 - 0,1 Bq. The initial contamination of ground happened a near zone within approximately first 2 weeks after accident. Since the end of May, 1986 till present time the air radioactive contamination is formed under action of processes of secondary wind rise and carry of radioactive particles with contaminated territories, which depends from a number factors of both natural and anthropogenous origin. The relations of Pu-238/Pu-239 and Sr-90/Pu-239 in aerosol particles practically coincide with the same relations in fuel blown up reactor. The ratio of Cs-137/Pu-239 activities in aerosols considerably exceeds the resettlement ratio for fuel. Strontium and plutonium are in structure of fuel particles, and cesium aerosol have other origin. The analysis of changes annual radioisotopes concentration in air of towns of Belarus specifies existence of the tendency to slow decrease of contamination of atmosphere by radioisotopes of industrial origins. The basic tendency of formation of air radioactive contamination is determined by the contents of a dust at surface layer of an atmosphere and its specific activity. Annual average dust content of air in a zone resettlement was least and made about 10 mk g

Oxygen Selective Membranes for Li-Air (O₂ Batteries

Directory of Open Access Journals (Sweden)

Mark Salomon

2012-05-01

Full Text Available Lithium-air (Li-air batteries have a much higher theoretical energy density than conventional lithium batteries and other metal air batteries, so they are being developed for applications that require long life. Water vapor from air must be prevented from corroding the lithium (Li metal negative electrode during discharge under ambient conditions, i.e., in humid air. One method of protecting the Li metal from corrosion is to use an oxygen selective membrane (OSM that allows oxygen into the cell while stopping or slowing the ingress of water vapor. The desired properties and some potential materials for OSMs for Li-air batteries are discussed and the literature is reviewed.

Magnitude and reactivity consequences of moisture ingress into the modular High-Temperature Gas-Cooled Reactor core

International Nuclear Information System (INIS)

Smith, O.L.

1992-12-01

Inadvertent admission of moisture into the primary system of a modular high-temperature gas-cooled reactor has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moisture-ingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. The rate and magnitude of steam buildup are found to be dominated by major system features such as break size compared with safety valve capacity and reliability and less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions

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

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)

A micro-gap, air-filled ionisation chamber as a detector for criticality accident dosimetry

International Nuclear Information System (INIS)

Murawski, I.; Zielczynski, M.; Gryzinski, M.A.; Golnik, N.

2014-01-01

A micro-gap air-filled ionisation chamber was designed for criticality dosimetry. The special feature of the chamber is its very small gap between electrodes of only 0.3 mm. This prevents ion recombination at high dose rates and minimises the influence of gas on secondary particles spectrum. The electrodes are made of polypropylene because of higher content of hydrogen in this material, when compared with soft tissue. The difference between neutron and gamma sensitivity in such chamber becomes practically negligible. The chamber's envelope contains two specially connected capacitors, one for polarising the electrodes and the other for collecting the ionisation charge. Air-filled ionisation chamber with very small gap is a simple dosemeter, which fulfills the most desired properties of criticality accident dosemeters. Short ion collection time is achieved by combination of small gap and relatively high polarising voltage. For the same reason, parasitic recombination of ions in the chamber is negligibly small even at high dose rates. The difference between neutron and gamma sensitivity is small for tissue-equivalent chamber and is expected to become practically negligible when the chamber electrodes are made of polypropylene. Additional capacitor provides a broad measuring range from ∼0.1 Gy up to ∼25 Gy; however, leakage of electrical charge from polarising capacitor has to be observed and taken into account. Periodical re-charging of the device is necessary. Obviously, final test of the device in conditions simulating criticality accident is needed and will be performed as soon as available. (authors)

Non-destructive evaluation of water ingress in photovoltaic modules

Energy Technology Data Exchange (ETDEWEB)

Bora, Mihail; Kotovsky, Jack

2017-03-07

Systems and techniques for non-destructive evaluation of water ingress in photovoltaic modules include and/or are configured to illuminate a photovoltaic module comprising a photovoltaic cell and an encapsulant with at least one beam of light having a wavelength in a range from about 1400 nm to about 2700 nm; capture one or more images of the illuminated photovoltaic module, each image relating to a water content of the photovoltaic module; and determine a water content of the photovoltaic module based on the one or more images. Systems preferably include one or more of a light source, a moving mirror, a focusing lens, a beam splitter, a stationary mirror, an objective lens and an imaging module.

Some observations on the problem of jelly fish ingress in a power station cooling system at Kalpakkam, east coast of India

International Nuclear Information System (INIS)

Rajagopal, S.; Azariah, Jayapaul; Nair, K.V.K.

1989-01-01

The paper reports results of a study on the seasonal distribution and abundance of jelly fishes in the coastal waters of Kalpakkam in the context of their ingress into the cooling system of a power plant. Three species of jelly fishes Dactylometra quinquicirrha (L. agassiz), Crambionella stuhlmanni (Chun) and Chiropsalmus buitendijki (Horst) were found in such abundance as to cause blockage of the cooling water intake screens. The seasonal variation in the incidence of the jelly fishes clearly showed that the periods of maxima and minima were different for each species. During the period of study (February 1988 to April 1989) three maxima were noted - May, July and October. On any given single day, maximum quantity of jelly fishes collected from the travelling screens was 29 tonnes (21 July 1988). The data are discussed with respect of (i)qualitative and quantitative variations in the ingress of jelly fishes at Kalpakkam, (ii)operational problems associated with jelly fish ingress and (iii)possible approaches to combat the problem. (author). 1 tab., 4 figs., 9 ref s

[Accidents in travellers - the hidden epidemic].

Science.gov (United States)

Walz, Alexander; Hatz, Christoph

2013-06-01

The risk of malaria and other communicable diseases is well addressed in pre-travel advice. Accidents are usually less discussed. Thus, we aimed at assessing accident figures for the Swiss population, based on data of the register from 2004 to 2008 of the largest Swiss accident insurance organization (SUVA). More than 139'000 accidents over 5 years showed that 65 % of the accidents overseas are injuries, and 24 % are caused by poisoning or harm by cold, heat or air pressure. Most accidents happened during leisure activities or sports. More than one third of the non-lethal and more than 50 % of the fatal accidents happened in Asia. More than three-quarters of non-lethal accidents take place in people between 25 and 54 years. One out of 74 insured persons has an accident abroad per year. Despite of many analysis short-comings of the data set with regard to overseas travel, the figures document the underestimated burden of disease caused by accidents abroad and should affect the given pre-health advice.

First retrieval of hourly atmospheric radionuclides just after the Fukushima accident by analyzing filter-tapes of operational air pollution monitoring stations.

Science.gov (United States)

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

2014-10-22

No observed data have been found in the Fukushima Prefecture (FP) for the time-series of atmospheric radionuclides concentrations just after the Fukushima Daiichi Nuclear Power Plant (FD1NPP) accident. Accordingly, current estimates of internal radiation doses from inhalation, and atmospheric radionuclide concentrations by atmospheric transport models are highly uncertain. Here, we present a new method for retrieving the hourly atmospheric (137)Cs concentrations by measuring the radioactivity of suspended particulate matter (SPM) collected on filter tapes in SPM monitors which were operated even after the accident. This new dataset focused on the period of March 12-23, 2011 just after the accident, when massive radioactive materials were released from the FD1NPP to the atmosphere. Overall, 40 sites of the more than 400 sites in the air quality monitoring stations in eastern Japan were studied. For the first time, we show the spatio-temporal variation of atmospheric (137)Cs concentrations in the FP and the Tokyo Metropolitan Area (TMA) located more than 170 km southwest of the FD1NPP. The comprehensive dataset revealed how the polluted air masses were transported to the FP and TMA, and can be used to re-evaluate internal exposure, time-series radionuclides release rates, and atmospheric transport models.

First retrieval of hourly atmospheric radionuclides just after the Fukushima accident by analyzing filter-tapes of operational air pollution monitoring stations

Science.gov (United States)

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

2014-01-01

No observed data have been found in the Fukushima Prefecture (FP) for the time-series of atmospheric radionuclides concentrations just after the Fukushima Daiichi Nuclear Power Plant (FD1NPP) accident. Accordingly, current estimates of internal radiation doses from inhalation, and atmospheric radionuclide concentrations by atmospheric transport models are highly uncertain. Here, we present a new method for retrieving the hourly atmospheric 137Cs concentrations by measuring the radioactivity of suspended particulate matter (SPM) collected on filter tapes in SPM monitors which were operated even after the accident. This new dataset focused on the period of March 12–23, 2011 just after the accident, when massive radioactive materials were released from the FD1NPP to the atmosphere. Overall, 40 sites of the more than 400 sites in the air quality monitoring stations in eastern Japan were studied. For the first time, we show the spatio-temporal variation of atmospheric 137Cs concentrations in the FP and the Tokyo Metropolitan Area (TMA) located more than 170 km southwest of the FD1NPP. The comprehensive dataset revealed how the polluted air masses were transported to the FP and TMA, and can be used to re-evaluate internal exposure, time-series radionuclides release rates, and atmospheric transport models. PMID:25335435

Conceptual Design of Portable Filtered Air Suction Systems For Prevention of Released Radioactive Gas under Severe Accidents of NPP

Energy Technology Data Exchange (ETDEWEB)

Gu, Beom W.; Choi, Su Y.; Yim, Man S.; Rim, Chun T. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-05-15

It becomes evident that severe accidents may occur by unexpected disasters such as tsunami, heavy flood, or terror. Once radioactive material is released from NPP through severe accidents, there are no ways to prevent the released radioactive gas spreading in the air. As a remedy for this problem, the idea on the portable filtered air suction system (PoFASS) for the prevention of released radioactive gas under severe accidents was proposed. In this paper, the conceptual design of a PoFASS focusing on the number of robot fingers and robot arm rods are proposed. In order to design a flexible robot suction nozzle, mathematical models for the gaps which represent the lifted heights of extensible covers for given convex shapes of pipes and for the covered areas are developed. In addition, the system requirements for the design of the robot arms of PoFASS are proposed, which determine the accessible range of leakage points of released radioactive gas. In this paper, the conceptual designs of the flexible robot suction nozzle and robot arm have been conducted. As a result, the minimum number of robot fingers and robot arm rods are defined to be four and three, respectively. For further works, extensible cover designs on the flexible robot suction nozzle and the application of the PoFASS to the inside of NPP should be studied because the radioactive gas may be released from connection pipes between the containment building and auxiliary buildings.

Source terms in relation to air cleaning

International Nuclear Information System (INIS)

Bernero, R.M.

1985-01-01

There are two sets of source terms for consideration in air cleaning, those for routine releases and those for accident releases. With about 1000 reactor years of commercial operating experience in the US done, there is an excellent data base for routine and expected transient releases. Specifications for air cleaning can be based on this body of experience with confidence. Specifications for air cleaning in accident situations is another matter. Recent investigations of severe accident behavior are offering a new basis for source terms and air cleaning specifications. Reports by many experts in the field describe an accident environment notably different from previous models. It is an atmosphere heavy with aerosols, both radioactive and inert. Temperatures are sometimes very high; radioiodine is typically in the form of cesium iodide aerosol particles; other nuclides, such as tellurium, are also important aerosols. Some of the present air cleaning requirements may be very important in light of these new accident behavior models. Others may be wasteful or even counterproductive. The use of the new data on accident behavior models to reevaluate requirements promptly is discussed

Revised accident source terms for light-water reactors

Energy Technology Data Exchange (ETDEWEB)

Soffer, L. [Nuclear Regulatory Commission, Washington, DC (United States)

1995-02-01

This paper presents revised accident source terms for light-water reactors incorporating the severe accident research insights gained in this area over the last 15 years. Current LWR reactor accident source terms used for licensing date from 1962 and are contained in Regulatory Guides 1.3 and 1.4. These specify that 100% of the core inventory of noble gases and 25% of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental iodine. These assumptions have strongly affected present nuclear air cleaning requirements by emphasizing rapid actuation of spray systems and filtration systems optimized to retain elemental iodine. A proposed revision of reactor accident source terms and some im implications for nuclear air cleaning requirements was presented at the 22nd DOE/NRC Nuclear Air Cleaning Conference. A draft report was issued by the NRC for comment in July 1992. Extensive comments were received, with the most significant comments involving (a) release fractions for both volatile and non-volatile species in the early in-vessel release phase, (b) gap release fractions of the noble gases, iodine and cesium, and (c) the timing and duration for the release phases. The final source term report is expected to be issued in late 1994. Although the revised source terms are intended primarily for future plants, current nuclear power plants may request use of revised accident source term insights as well in licensing. This paper emphasizes additional information obtained since the 22nd Conference, including studies on fission product removal mechanisms, results obtained from improved severe accident code calculations and resolution of major comments, and their impact upon the revised accident source terms. Revised accident source terms for both BWRS and PWRS are presented.

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

Accident resistant transport container

Science.gov (United States)

Anderson, J.A.; Cole, K.K.

The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

Accident resistant transport container

International Nuclear Information System (INIS)

Andersen, J.A.; Cole, J.K.

1980-01-01

The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident

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)

Moisture ingress into electronics enclosures under isothermal conditions

International Nuclear Information System (INIS)

Staliulionis, Ž.; Jabbari, M.; Hattel, J. H.

2016-01-01

The number of electronics used in outdoor environment is constantly growing. The humidity causes about 19 % of all electronics failures and, especially, moisture increases these problems due to the ongoing process of miniaturization and lower power consumption of electronic components. Moisture loads are still not understood well by design engineers, therefore this field has become one of the bottlenecks in the electronics system design. The objective of this paper is to model moisture ingress into an electronics enclosure under isothermal conditions. The moisture diffusion model is based on a 1D quasi-steady state (QSS) approximation for Fick’s second law. This QSS approach is also described with an electrical analogy which gives a fast tool in modelling of the moisture response. The same QSS method is applied to ambient water vapour variations. The obtained results are compared to an analytical solution and very good agreement is found.

Transport Accident Costs and the Value of Safety

DEFF Research Database (Denmark)

Koornstra, Matthijs; Evans, Andrew; Glansdorp, Cees

The publication descibes a study of costs of passenger transport accident by road, rail, air and sea. It is argued that "willingness to pay" theory should be preferred to "human capital" theory in valuations of life and limb. The total costs of passenger transport accidents in the EU is estimated...... to about 165 billion ECU in 1995 year prices. Road accident costs account for more than 95% of the costs....

32 CFR 256.7 - Accident potential zone guidelines.

Science.gov (United States)

2010-07-01

... 32 National Defense 2 2010-07-01 2010-07-01 false Accident potential zone guidelines. 256.7 Section 256.7 National Defense Department of Defense (Continued) OFFICE OF THE SECRETARY OF DEFENSE (CONTINUED) MISCELLANEOUS AIR INSTALLATIONS COMPATIBLE USE ZONES § 256.7 Accident potential zone guidelines...

Three Mile Island Accident Data

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Three Mile Island Accident Data consists of mostly upper air and wind observations immediately following the nuclear meltdown occurring on March 28, 1979, near...

The transgressive-regressive cycle of the Romualdo Formation (Araripe Basin): Sedimentary archive of the Early Cretaceous marine ingression in the interior of Northeast Brazil

Science.gov (United States)

Custódio, Michele Andriolli; Quaglio, Fernanda; Warren, Lucas Veríssimo; Simões, Marcello Guimarães; Fürsich, Franz Theodor; Perinotto, José Alexandre J.; Assine, Mario Luis

2017-08-01

Geologic events related to the opening of the South Atlantic Ocean deeply influenced the sedimentary record of the Araripe Basin. As consequence, upper stratigraphic units of the basin record a marine ingression in northeastern Brazil during the late Aptian. The timing and stratigraphic architecture of these units are crucial to understand the paleogeography of Gondwana and how the proto-Atlantic Ocean reached interior NE Brazil during the early Cretaceous. This marine ingression is recorded in the Araripe Basin as the Romualdo Formation, characterized by a transgressive-regressive cycle bounded by two regional unconformities. In the eastern part of the basin, the Romualdo depositional sequence comprises coastal alluvial and tide-dominated deposits followed by marine transgressive facies characterized by two fossil-rich intervals: a lower interval of black shales with fossil-rich carbonate concretions (Konservat-Lagerstätten) and an upper level with mollusk-dominated shell beds and shelly limestones. Following the marine ingression, an incomplete regressive succession of marginal-marine facies records the return of continental environments to the basin. The stratigraphic framework based on the correlation of several sections defines a transgressive-regressive cycle with depositional dip towards southeast, decreasing in thickness towards northwest, and with source areas located at the northern side of the basin. The facies-cycle wedge-geometry, together with paleocurrent data, indicates a coastal onlap towards NNW. Therefore, contrary to several paleogeographic scenarios previously proposed, the marine ingression would have reached the western parts of the Araripe Basin from the SSE.

Perspectives on understanding and verifying the safety terrain of modular high temperature gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Carlson, Donald E., E-mail: donald@carlsonperin.net [11221 Empire Lane, Rockville, MD 20852 (United States); Ball, Sydney J., E-mail: beckysyd@comcast.net [100 Greywood Place, Oak Ridge, TN 37830 (United States)

2016-09-15

The passive safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving loss of forced cooling, coolant depressurization, air ingress, moisture ingress, and reactivity events. In addition to discussing associated uncertainties and potential measures to address them, this paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs.

Perspectives on Understanding and Verifying the Safety Terrain of Modular High Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

Carlson, Donald E.

2014-01-01

The inherent safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving coolant depressurization, air ingress, moisture ingress, and reactivity insertion. In addition to discussing associated uncertainties and potential measures to address them, the paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs. (author)

Paralysis from sport and diving accidents.

Science.gov (United States)

Schmitt, H; Gerner, H J

2001-01-01

To examine the causes of sport-related spinal cord injuries that developed into paraplegia or tetraplegia, and to compare data from different sports with previous studies in the same geographical region. A retrospective epidemiological study and comparison with previous studies. The Orthopedic Department, specializing in the treatment and rehabilitation of paralyzed patients, at the University of Heidelberg, Germany. Between 1985 and 1997, 1,016 cases of traumatic spinal cord injury presented at the Orthopedic Department at the University of Heidelberg: 6.8% were caused by sport and 7.7% by diving accidents. Sport-related spinal cord injuries with paralysis. A total of 1.016 cases of traumatic spinal cord injury were reviewed. Of these, 14.5% were caused by sport accidents (n = 69) or diving accidents (n = 78). Age of patients ranged from 9 to 52 years. 83% were male. 77% of the patients developed tetraplegia, and 23%, paraplegia. 16 of the sport accidents resulted from downhill skiing, 9 resulted from horseback riding, 7 from modern air sports, 6 from gymnastics, 5 from trampolining, and 26 from other sports. Previous analyses had revealed that paraplegia had mainly occurred from gymnastics, trampolining, or high diving accidents. More recently, however, the number of serious spinal injuries caused by risk-filled sports such as hang gliding and paragliding has significantly increased (p = 0.095), as it has for horseback riding and skiing. Examinations have shown that all patients who were involved in diving accidents developed tetraplegia. An analysis of injury from specific sports is still under way. Analysis of accidents resulting in damage to the spinal cord in respect to different sports shows that sports that have become popular during the last 10 years show an increasing risk of injury. Modern air sports hold the most injuries. Injury-preventing strategies also are presented.

Zagreb and Tenerife: Airline Accidents Involving Linguistic Factors

Science.gov (United States)

Cookson, Simon

2009-01-01

The International Civil Aviation Organization (ICAO) is currently implementing a program to improve the language proficiency of pilots and air traffic controllers worldwide. In justifying the program, ICAO has cited a number of airline accidents that were at least partly caused by language factors. Two accidents cited by ICAO are analysed in this…

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

Investigation of Zircaloy-2 oxidation model for SFP accident analysis

Energy Technology Data Exchange (ETDEWEB)

Nemoto, Yoshiyuki, E-mail: nemoto.yoshiyuki@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirakata, Ohaza, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Kaji, Yoshiyuki; Ogawa, Chihiro; Kondo, Keietsu [Japan Atomic Energy Agency, 2-4 Shirakata, Ohaza, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Nakashima, Kazuo; Kanazawa, Toru; Tojo, Masayuki [Global Nuclear Fuel – Japan Co., Ltd., 2-3-1, Uchikawa, Yokosuka-shi, Kanagawa, 239-0836 (Japan)

2017-05-15

The authors previously conducted thermogravimetric analyses on Zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study. - Highlights: •An oxidation model of Zircaloy-2 in air environment was developed. •The oxidation model was validated by the comparison with oxidation tests using long cladding tubes in hypothetical spent fuel pool accident condition. •The oxidation model successfully reproduced the typical oxidation behavior in air.

Protective action evaluation, Part 1. Effectiveness of sheltering as a protective action against nuclear accidents involving gaseous releases

International Nuclear Information System (INIS)

Anno, G.H.; Dore, M.A.

1978-04-01

In an airborne release of radioactive material from a nuclear power plant accident, sheltering of individuals is of importance in emergency protective action planning. An analysis to estimate the effectiveness or benefit that might be derived from sheltering is described. The objective of this effort is the development of sheltering effectiveness information for those responsible for formulating required emergency plans for nuclear power plant siting. Shelter effectiveness is specifically defined as the dose reduction factor (DRF). DRF estimates for different conditions of source release, shelter structure assumptions, and operational time parameters are made for both whole-body and thyroid doses separately, based on a single-compartment structural model of the time-varying outside and inside gaseous radionuclide sources of krypton, zenon, and iodine. Design basis accident (DBA) assumptions are made for the gaseous radionuclide release. The magnitude of the release and dose estimates are based on radionuclide data from The Reactor Safety Study (WASH-1400). Source release time and duration assumptions are related to release categories PWR 1, PWR 3, and PWR 4, for which release times range from 1.5 to 2.5 hr and the release duration ranges from 0.5 to 3 hr. The basic shelter model characteristics considered are gamma ray attentuation, source geometry, gaseous fission-product ingress, and air change rate. Temporal parameters considered are source release time and duration, cloud travel time, and time spent in the shelter structure. Also, the analysis of shelter effectiveness is based on a time-frame model, which can be conveniently related to other operational times important for emergency planning. In addition to developing shelter-effectiveness estimates parametrically, the advantage of exiting and evacuating the vicinity of the shelter area after some initial time in the shelter is analyzed from the standpoint of the DRF and temporal considerations

Detection device for off-gas system accidents

International Nuclear Information System (INIS)

Kubota, Ryuji; Tsuruoka, Ryozo; Yamanari, Shozo.

1984-01-01

Purpose: To rapidly isolate the off-gas system by detecting the off-gas system failure accident in a short time. Constitution: Radiation monitors are disposed to ducts connecting an exhaust gas area and an air conditioning system as a portion of a turbine building. The ducts are disposed independently such that they ventilate only the atmosphere in the exhaust gas area and do not mix the atmosphere in the turbine building. Since radioactivity issued upon off-gas accidents to the exhaust gas area is sucked to the duct, it can be detected by radiation detection monitors in a short time after the accident. Further, since the operator judges it as the off-gas system accident, the off-gas system can be isolated in a short time after the accident. (Moriyama, K.)

Confinement of airborne particulate radioactivity in the case of an accident

International Nuclear Information System (INIS)

Ruedinger, V.; Wilhelm, J.G.

1984-01-01

In the case of an accident, the filter elements on the inlets and exhausts of the air-cleaning systems of a nuclear facility may become a part of the remaining fission product barrier. Among others, the Project Nuclear Safety is pursuing the information necessary to insure safe operation of air-cleaning systems under accident conditions. Experimental investigations into the response of HEPA filters to differential presssures involving both dry and moist air have demonstrated the occurrence of structural failure with subsequent loss of efficiency at low values of differential pressures. Contributions are being made to the development and verification of computer codes used to calculate those fluid-dynamic and thermodynamic conditions expected to prevail in an air-cleaning system as a result of potential accident situations. With regard to further investigations, a new test facility was put into operation for the realization of superimposed challenges and a new method for testing particulate removal efficiency under high temperature or high humidity was developed. (orig./HP)

Air temperature determination inside residual heat removal pump room of Angra-1 nuclear power plant after a design basic accident

International Nuclear Information System (INIS)

Siniscalchi, Marcio Rezende

2005-01-01

This work develops heat transfer theoretical models for determination of air temperature inside the Residual Heat Removal Pump Room of Angra 1 Nuclear Power Plant after a Design Basis Accident without forced ventilation. Two models had been developed. The differential equations are solved by analytical methods. A software in FORTRAN language are developed for simulations of temperature inside rooms for different geometries and materials. (author)

On the air coolability of TRIGA reactors following a loss-of-coolant accident

International Nuclear Information System (INIS)

El-Genk, Mohamed S.; Kim, Sung-Ho; Zaki, Galal M.; Foushee, Fabian; Philbin, Jeffrey S.; Schulze, James

1986-01-01

This paper describes the experiments on the air-coolability of a heated rod in a vertical open annulus at near atmospheric pressure. This data can be applied to the coolability of reactor fuel rods that are totally uncovered in a Loss-of-Coolant Accident (LOCA). As a prelude to measuring air coolability of specific core geometries (bundles), heat transfer data was collected for natural convection of atmospheric air in open vertical annuli with an isoflux inner wall and an insulated outer wall (diameter ratios, annulus ratio, of 1.155, 1.33, 1.63, and 12). Although the inner heated tube had the same overall dimensions as the fuel rod in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (3.81 cm o.d. and 55.5 cm long), the heated length was only 36.0 cm rather than the entire 50.5 cm for the ACRR's rods. The test assembly was operated at heat fluxes up to 1.38 W/cm 2 with a corresponding surface temperature of 852 K. The annulus data was extrapolated to an equilibrium surface temperature of 1200 K (as a coolability limit of TRIGA reactors) to provide a qualitative estimate of the coolability of multirod bundles by free convection of atmospheric air. The results suggest that for a typical pitch-to-diameter ratio of 1.12 in the ACRR the decay heat removal level is about 1.0 kW/m. This corresponds to an initial decay power following sustained operations at about 12.5 kW/m in the ACRR. However, because of the uncertainties in duplicating the actual thermal-hydraulic conditions in a multirod bundle using a single rod annulus, the actual coolability of open pool reactors could be different from those suggested in this paper. (author)

Radioactive fallout from the Chernobyl nuclear reactor accident

International Nuclear Information System (INIS)

Beiriger, J.M.; Failor, R.A.; Marsh, K.V.; Shaw, G.E.

1987-08-01

This report describes the detection of fallout in the United States from the Chernobyl nuclear reactor accident. As part of its environmental surveillance program, Lawrence Livermore National Laboratory maintained detectors for gamma-emitting radionuclides. Following the reactor accident, additional air filters were set out. Several uncommon isotopes were detected at the time the plume passed into the US

Chloride Ingress in Concrete with Different Age at Time of First Chloride Exposure

DEFF Research Database (Denmark)

Hansen, Esben Østergaard; Iskau, Martin Riis; Hasholt, Marianne Tange

2016-01-01

Concrete structures cast in spring have longer time to hydrate and are therefore denser and more resistant to chloride ingress when first subjected to deicing salts in winter than structures cast in autumn. Consequently, it is expected that a spring casting will have a longer service life....... This hypothesis is investigated in the present study by testing drilled cores from concrete cast in 2012 and 2013 on the Svendborgsund Bridge. The cores are subject to petrographic examination and mapping of chloride profiles. Moreover, chloride migration coefficients have been measured. The study shows...

Observations on radioactivity from the Chernobyl accident

International Nuclear Information System (INIS)

Cambray, R.S.; Cawse, P.A.; Garland, J.A.; Gibson, J.A.B.; Johnson, P.; Lewis, G.N.J.; Newton, D.; Salmon, L.; Wade, B.O.

1987-02-01

A preliminary study of radioactivity from the Chernobyl accident for the Department of the Environment was started in June 1986 which involved taking on an opportunistic basis, samples of air, rain, grass and soil in the UK. This study was integrated into a programme of other investigations funded by the Departments of Health and Social Security and of Energy including measurements on people, in air, deposition and soil overseas, on deposition to buildings and the derivation where possible of parameters of interest in accident assessment. This report is a comprehensive account of all these initial investigations and presented in fulfilment of the Preliminary Study under DoE contract PECD 7/9/359. (author)

Air transport pilot involvement in general aviation accidents

Science.gov (United States)

1986-01-01

General aviation (GA) fatal accident records of airport transport pilots (ATPs) : were : compared to those of private pilots (PVTs). : ATPs are safer GA pilots than the PVTs. : They have comparable exposure in GA airplanes and account for 7.5% of all...

Biophysical processes leading to the ingress of temperate fish larvae into estuarine nursery areas: A review

Science.gov (United States)

Teodósio, Maria Alexandra; Paris, Claire B.; Wolanski, Eric; Morais, Pedro

2016-12-01

A series of complementary hypotheses have been proposed to explain the recruitment of marine and temperate pelagic fish larvae originated from pelagic eggs in coastal environments. In this review, we propose a new and complementary hypothesis describing the biophysical processes intervening in the recruitment of temperate fish larvae into estuaries. This new hypothesis, the Sense Acuity And Behavioral (SAAB) hypothesis, recognizes that recruitment is unlikely if the larvae drift passively with the water currents, and that successful recruitment requires the sense acuity of temperate fish larvae and their behavioral response to the estuarine cues present in coastal areas. We propose that temperate fish larvae use a hierarchy of sensory cues (odor, sound, visual and geomagnetic cues) to detect estuarine nursery areas and to aid during navigation towards these areas. The sensorial acuity increases along ontogeny, which coincides with increased swimming capabilities. The swimming strategies of post-flexion larvae differ from offshore areas to the tidal zone. In offshore areas, innate behavior might lead larvae towards the coast guided by a sun compass or by the earth's geomagnetic field. In areas under limited influence of estuarine plumes (either in energetic nearshore areas or offshore), post-flexion larvae display a searching swimming behavior for estuarine disconnected patches (infotaxis strategy). After finding an estuarine plume, larvae may swim along the increasing cue concentration to ingress into the estuary. Here, larvae exhibit a rheotaxis behavior and avoid displacement by longshore currents by keeping bearing during navigation. When larvae reach the vicinity of an estuary, merging diel rhythms with feeding and predator avoidance strategies with tidally induced movements is essential to increase their chances of estuarine ingress. A fish larva recruitment model developed for the Ria Formosa lagoon supports the general framework of the SAAB hypothesis. In

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.)

36 CFR 14.5 - Nature of interest granted; settlement on right-of-way; rights of ingress and egress.

Science.gov (United States)

2010-07-01

... 36 Parks, Forests, and Public Property 1 2010-07-01 2010-07-01 false Nature of interest granted; settlement on right-of-way; rights of ingress and egress. 14.5 Section 14.5 Parks, Forests, and Public Property NATIONAL PARK SERVICE, DEPARTMENT OF THE INTERIOR RIGHTS-OF-WAY Nature of Interest § 14.5 Nature...

Bubble-vacuum system of accident localization of reference nuclear power plant with two WWER's

International Nuclear Information System (INIS)

Sykora, D.; Sykorova, I.

1988-01-01

Higher efficiency of the safety system for removing the consequences of project design accidents and higher radiation safety of a nuclear power plant with two WWER-440 units is the subject of Czechoslovak patent document 243961. The principle consists in interconnecting air chambers which are the end parts of safety systems for the two units. The air chamber is separated from the other parts of the safety system by double swing-check valves or closures. The connecting pipes of the two air chambers do not in any way reduce the reliability of the safety system thanks to their high technical safety and totally passive function. The benefits of the interconnection of the air chambers are given by the fact that it reduces maximum accident overpressure both in the air chambers and in the airtight zones. The reduction of the overpressure reduces the total leakage of radioactive substances and the radiation burden of the environment in case of a nuclear power plant accident. (Z.M.). 2 figs

I-NERI QUARTERLY TECHNICAL PROGRESS REPORT - JANUARY 1 - MARCH 31, 2005

International Nuclear Information System (INIS)

Chang Oh

2005-01-01

The objective of this Korean/United States/laboratory/university collaboration is to develop new advanced computational methods for safety analysis codes for very-high-temperature gas-cooled reactors (VHTGRs) and numerical and experimental validation of these computer codes. The research will improve two well-respected light water reactor transient response codes (RELAP5/ATHENA and MELCOR) in the modeling of molecular diffusion and chemical equilibrium, and to validate these codes against the VHTGR accident data, i.e., air ingress and others from the literature. The VHTGR is intrinsically safe, has a proliferation-resistant fuel cycle, and many advantages relative to light water reactors (LWRs). This study consists of five tasks for FY-03: (1) development of computational methods for the VHTGR, (2) theoretical modification of aforementioned computer codes for molecular diffusion (RELAP5/ATHENA) and modeling CO and CO 2 equilibrium (MELCOR), (3) development of a state-of-the-art methodology for VHTGR neutronic analysis and calculation of accurate power distributions and decay heat deposition rates, (4) reactor cavity cooling system experiment, and (5) graphite oxidation experiment. First quarter of Year 3: (A) Prof. NO and Kim continued Task 1. We first performed the chemical reaction test for the VELUNA pebble oxidation experiment and then the analysis of the air ingress accident for PBMR 268MWt. In the GAMMA analysis, significant rise in pebble temperature was observed at the bottom of the core due to graphite oxidation. Since the air ingress process depends on the vault conditions, further analysis coupled with more detailed vault or containment modeling would be necessary as a future study. (B) Prof. Park continued Task 2. The experiments for SNU-RCCS were continued to provide the experimental data for the validation of the thermal hydraulic code being developed at KAIST and to evaluate the performance of the system using the experiments and system analysis

Source term estimation for small sized HTRs: status and further needs - a german approach

International Nuclear Information System (INIS)

Moormann, R.; Schenk, W.; Verfondern, K.

2000-01-01

The main results of German studies on source term estimation for small pebble-bed HTRs with their strict safety demands are outlined. Core heat-up events are no longer dominant for modern high quality fuel, but fission product transport during water ingress accidents (steam cycle plants) and depressurization is relevant, mainly due to remobilization of fission products which were plated-out in the course of normal operation or became dust borne. An important lack of knowledge was identified as concerns data on plate-out under normal operation, as well as on the behaviour of dust borne activity as a whole. Improved knowledge in this field is also important for maintenance/repair and design/shielding. For core heat-up events the influence of burn-up on temperature induced fission product release has to be measured for future high burn-up fuel. Also, transport mechanisms out of the He circuit into the environment require further examination. For water/steam ingress events mobilization of plated-out fission products by steam or water has to be considered in detail, along with steam interaction with kernels of particles with defective coatings. For source terms of depressurization, a more detailed knowledge of the flow pattern and shear forces on the various surfaces is necessary. In order to improve the knowledge on plate-out and dust in normal operation and to generate specimens for experimental remobilization studies, planning/design of plate-out/dust examination facilities which could be added to the next generation of HTRs (HTR10,HTTR) is proposed. For severe air ingress and reactivity accidents, behaviour of future advanced fuel elements has to be experimentally tested. (authors)

Database on aircraft accidents

International Nuclear Information System (INIS)

Nishio, Masahide; Koriyama, Tamio

2012-09-01

probability of aircraft crash are considered to be 4 for large fixed-wing aircraft, 35 for small fixed-wing aircraft, 2 for large bladed aircraft and 30 for small bladed aircraft. (2) The data of SDF aircraft accidents is based on newspapers and aircraft magazines. The total of 45 accidents occurred, 24 of them are large fixed-wing aircrafts, 4 are small fixed-wing aircrafts and 17 are bladed aircrafts. 23 to 45 accidents fell into land. (3) The data of US aircraft accidents is based on newspapers and aircraft magazines. The total of 16 accidents occurred, 13 of which are fixed-wing aircrafts, 3 are bladed aircrafts. 6 of 16 accidents fell into land. (4) The instrument flight information on commercial aircrafts is based on 'Air transport statics' of Ministry of Land, Infrastructure, Transport and Tourism. The total number of takeoffs and landings is 29,971,662 and the total flight distance is 9,232,867,684 km. (author)

Early air sampling in Higashi-Hiroshima after the Fukushima Daiichi Nuclear Power Plant accident and subsequent sampling in Minami-Souma City from October 2011 to September 2012

International Nuclear Information System (INIS)

Shizuma, Kiyoshi; Kajimoto, Tsuyoshi; Endo, Satoru

2013-01-01

Soon after the accident that occurred at the Fukushima Daiichi Nuclear Power Plant on March 11, 2011, air sampling was performed in Higashi-Hiroshima (about 800 km from Fukushima). The sampling began on March 20, and a high-volume air sampler was used. The first radionuclide observed was "1"3"1I (on March 30), and "1"3"7Cs, "1"3"4Cs, "1"3"6Cs, and "1"3"2Te were observed thereafter. According to the sampling, the maximum concentration in the air occurred on April 7 and again on April 18. After April 30, no radionuclides released due to the accident were observed. Since "1"3"7Cs and "1"3"4Cs have a long half-life (30 and 2 y, respectively), all of the air filters were measured again in February through March of 2012 using a well-type Ge detector. The results of these measurements showed that "1"3"7Cs was already observed on March 21. Later on, air sampling was performed at Minami-Souma City, Fukushima (October 2011 to September 2012). The purpose of this sampling was to investigate whether any radionuclides were released from the forest and flew up into the air from the ground, or were carried by the wind in winter or via pollen in spring. Results showed that the radioactive concentration was quite low and no seasonal variation occurred, indicating that no radionuclides flew up into the air from the ground, nor were any released from the forest via wind or pollen. (author)

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

Analysis of the crush environment for lightweight air-transportable accident-resistant containers

International Nuclear Information System (INIS)

McClure, J.D.; Hartman, W.F.

1981-12-01

This report describes the longitudinal dynamic crush environment for a Lightweight Air-Transportable Accident-Resistant Container (LAARC, now called PAT-2) that can be used to transport small quantities of radioactive material. The analysis of the crush environment involves evaluation of the forces imposed upon the LAARC package during the crash of a large, heavily loaded, cargo aircraft. To perform the analysis, a cargo load column was defined which consisted of a longitudinal prism of cargo of cross-sectional area equal to the projected area of the radioactive-material package and length equal to the longitudinal extent of the cargo compartment in a commercial cargo jet aircraft. To bound the problem, two analyses of the cargo load column were performed, a static stability analysis and a dynamic analysis. The results of these analyses can be applied to other packaging designs and suggest that the physical limits or magnitude of the longitudinal crush forces, which are controlled in part by the yield strength of the cargo and the package size, are much smaller than previously estimated

Accommodation of potential hydrogen formation in LMFBR accidents

International Nuclear Information System (INIS)

Stepnewski, D.D.; Peak, R.D.; Mahaffey, M.K.

1981-01-01

Results of design verification tests for the FFTF reactor cavity liner system are presented which suggest that steel liners would retain their integrity even under certain hypothetical accident conditions, thus avoiding the formation of hydrogen. When liner failures are postulate in hypothetical reactor vessel meltthrough accidents, hydrogen levels can be controlled by an air purging system. The design of a containment purging and effluent scrubbing system is discussed

Criticality accident alarm system

International Nuclear Information System (INIS)

Malenfant, R.E.

1991-01-01

The American National Standard ANSI/ANS-8.3-1986, Criticality Accident Alarm System provides guidance for the establishment and maintenance of an alarm system to initiate personnel evacuation in the event of inadvertent criticality. In addition to identifying the physical features of the components of the system, the characteristics of accidents of concern are carefully delineated. Unfortunately, this ANSI Standard has led to considerable confusion in interpretation, and there is evidence that the ''minimum accident of concern'' may not be appropriate. Furthermore, although intended as a guide, the provisions of the standard are being rigorously applied, sometimes with interpretations that are not consistent. Although the standard is clear in the use of absorbed dose in free air of 20 rad, at least one installation has interpreted the requirement to apply to dose in soft tissue. The standard is also clear in specifying the response to both neutrons and gamma rays. An assembly of uranyl fluoride enriched to 5% 235 U was operated to simulate a potential accident. The dose, delivered in a free run excursion 2 m from the surface of the vessel, was greater than 500 rad, without ever exceeding a rate of 20 rad/min, which is the set point for activating an alarm that meets the standard. The presence of an alarm system would not have prevented any of the five major accidents in chemical operations nor is it absolutely certain that the alarms were solely responsible for reducing personnel exposures following the accident. Nevertheless, criticality alarm systems are now the subject of great effort and expense. 13 refs

Agricultural implications for Fukushima nuclear accident

International Nuclear Information System (INIS)

Nakanishi, Tomoko M.

2013-01-01

The overview of our research projects for Fukushima is presented including how they were derived. Then, where the fallout was found, right after the accident, is briefly summarized for soil, plants, trees, etc. The time of the accident was late winter, there were hardly any plants growing except for the wheat in the farming field. Most of the fallout was found at the surface of soil, tree barks, etc., which were exposed to the air at the time of the accident. The fallout found was firmly adsorbed to anything and did not move for months from the site when they first touched. Therefore, the newly emerged tissue after the accident showed very low radioactivity. The fallout contamination was not uniform, therefore, when radiograph of contaminated soil or leaves were taken, fallout was shown as spots. Generally, plants could not absorb radiocesium adsorbed to soil. Some of the results we obtained will be presented. (author)

DOZIM - evaluation dose code for nuclear accident

International Nuclear Information System (INIS)

Oprea, I.; Musat, D.; Ionita, I.

2008-01-01

During a nuclear accident an environmentally significant fission products release can happen. In that case it is not possible to determine precisely the air fission products concentration and, consequently, the estimated doses will be affected by certain errors. The stringent requirement to cope with a nuclear accident, even minor, imposes creation of a computation method for emergency dosimetric evaluations needed to compare the measurement data to certain reference levels, previously established. These comparisons will allow a qualified option regarding the necessary actions to diminish the accident effects. DOZIM code estimates the soil contamination and the irradiation doses produced either by radioactive plume or by soil contamination. Irradiations either on whole body or on certain organs, as well as internal contamination doses produced by isotope inhalation during radioactive plume crossing are taken into account. The calculus does not consider neither the internal contamination produced by contaminated food consumption, or that produced by radioactive deposits resuspension. The code is recommended for dose computation on the wind direction, at distances from 10 2 to 2 x 10 4 m. The DOZIM code was utilized for three different cases: - In air TRIGA-SSR fuel bundle destruction with different input data for fission products fractions released into the environment; - Chernobyl-like accident doses estimation; - Intervention areas determination for a hypothetical severe accident at Cernavoda Nuclear Power Plant. For the first case input data and results (for a 60 m emission height without iodine retention on active coal filters) are presented. To summarize, the DOZIM code conception allows the dose estimation for any nuclear accident. Fission products inventory, released fractions, emission conditions, atmospherical and geographical parameters are the input data. Dosimetric factors are included in the program. The program is in FORTRAN IV language and was run on

GAMMA-FR and MELCOR Validation using HCS Heat Exchanger Break Accident

International Nuclear Information System (INIS)

Jin, Hyung Gon; Hong, Yun Jeong; Cho, Seung Yon

2016-01-01

To confirm the HCCR-TBS integrity, enveloped cases from the conceivable events were evaluated and demonstrated compliance with the General Safety Objectives of ITER. In this analysis, amount of discharged helium is the key parameter to examine total tritium ingress to CCWS-1. In this regard, radiation heat transfer and temperature distribution along the pipes did not take account. Due to the same reason, flow network inside of TBM is simplified as one fluid volume (FB1300). In principle, transient of this accident is similar to LOHSA, therefore, TBM temperature is expected to be cool down by passive cooling and isolation valves avoid CCWS-1 pressure build-up during the accident. With relief valve, pressure of CCWS-1 is under 0.43 MPa during LOCA happens. (CCWS-1 max. design pressure: 1MPa). On the other hand, primary concern is tritium concentration increase in CCWS-1 because of tritium contents in HCS coolant. The important point is that CCWS-1 is an ESP device and its ESP level should be confirmed when operating with HCCR-TBS as well. Key parameters, which govern this transient, are relief valve operation, nitrogen in the pressurizer and flow area of the ruptured channels. Relief valve in CCWS-1 pressurizer opens at 0.41 MPa and closes 0.39 MPa, therefore, CCWS-1 pressure is impossible to exceed 0.41 MPa globally. As a comparison, calculation was conducted against CCWS-1 with relief valve (with RV) and without relief valve (without RV)

Effects of secondary containment air cleanup system leakage on the accident offsite dose as determined during preop tests of the Sequoyah Nuclear Plant

International Nuclear Information System (INIS)

Klaes, L.J.; Nass, S.A.; Proctor, L.D.

1981-01-01

The Sequoyah Nuclear Plant has two secondary containments. One is the annular region between the primary containment and the shield building surrounding the primary containment. The second is the auxiliary building secondary containment enclosure which is potentially subject to direct airborne radioactivity. Two air cleanup systems are provided to serve these areas. The emergency gas treatment system (EGTS) serves the annulus between the primary containment and the shield building, and the auxiliary building gas treatment system (ABGTS) serves the area inside of the auxiliary building secondary containment enclosure. The major function served by these air cleanup systems is that of controlling and processing airborne contamination released in these areas during any accident up to a design basis accident. This is accomplished by (1) creating a negative pressure in the areas served to ensure that no unprocessed air is released to the atmosphere, (2) providing filtration units to process all air exhausted from the secondary containment spaces, and (3) providing a low-leakage enclosure to limit exhaust flows. Offsite dose effects due to secondary containment release rates, bypass leakage, and duct and damper leakages are presented and parameter variations are considered. For the EGTS, a recirculation system, the most important parameter is the total inleakage of the system which causes an increase in both whole body (gamma) and thyroid (iodine) doses. For the ABGTS, a once-through system, the most important paramter is the inleakage which bypasses the filters resulting in an increase in the thyroid dose only. Actual preoperational test data are utilized. Problems encountered during the preop test are summarized. Solutions incorporated to bring the EGTS and ABGTS air cleanup systems within the test acceptance criteria required to meet offsite dose limitations are discussed and the resultant calculated offsite dose is presented

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

Transport of 131I and 137Cs from air to cows milk produced in north-western Italian farms following the Chernobyl accident

International Nuclear Information System (INIS)

Spezzano, P.; Giacomelli, R.

1991-01-01

After the Chernobyl accident, the levels of 131 I and 137 Cs were measured in surface air, pasture grass and milk produced by cows fed exclusively on fresh grass. The observed values of the vegetation-to-air, milk-to-vegetation and milk-to-air concentration ratios were compared with the values predicted by an internationally recognized assessment model for the transfer of radionuclides through terrestrial food chains. Predicted values were higher than the observed results by factors of 2.6, 2.1 and 5.6 for 131 I and 4.3, 3.7 and 16 for 137 Cs, for the vegetation-to-air, milk-to-vegetation and milk-to-air ratios, respectively. A better agreement between prediction and observation was achieved by using experimental values of the following parameters; the mass interception factor (R/Y), the effective decay constants on vegetation (λ ν ) and the milk transfer coefficients (F m ), these being lower than the model default values. Milk produced by dairy cows fed on a mixed diet showed a different behaviour with regard to excretion of 137 Cs. (author)

Air radioactivity levels following the Fukushima reactor accident measured at the Laboratoire Souterrain de Modane, France.

Science.gov (United States)

Loaiza, P; Brudanin, V; Piquemal, F; Reyss, J-L; Stekl, I; Warot, G; Zampaolo, M

2012-12-01

The radioactivity levels in the air of the radionuclides released by the Fukushima accident were measured at the Laboratoire Souterrain de Modane, in the South-East of France, during the period 25 March-18 April 2011. Air-filters from the ventilation system exposed for one or two days were measured using low-background gamma-ray spectrometry. In this paper we present the activity concentrations obtained for the radionuclides (131)I, (132)Te, (134)Cs, (137)Cs, (95)Nb, (95)Zr, (106)Ru, (140)Ba/La and (103)Ru. The activity concentration of (131)I was of the order of 100 μBq/m(3), more than 100 times higher than the activities of other fission products. The highest activities of (131)I were measured as a first peak on 30 March and a second peak on 3-4 April. The activity concentrations of (134)Cs and (137)Cs varied from 5 to 30 μBq/m(3). The highest activity concentration recorded for Cs corresponded to the same period as for (131)I, with a peak on 2-3 April. The results of the radioactivity concentration levels in grass and mushrooms exposed to the air in the Modane region were also measured. Activity concentrations of (131)I of about 100 mBq/m(2) were found in grass. Copyright © 2012 Elsevier Ltd. All rights reserved.

Ventilation-air conditioner system in nuclear power plant

International Nuclear Information System (INIS)

Kubota, Ryuji; Sugisaki, Toshihiko.

1989-01-01

This invention concerns a ventilation-air conditioner system which enables, upon occurrence of accidents in a nuclear power plant, continuous operation for other adjacent nuclear power plants with no effect of accidents. Air supply system and exhaust system are operated during usual operaiton. If loss of coolants accidents should occur in an adjacent nuclear power plants, operation is switched from ventilation operaiton to the operation of re-cycling system based on an AND logic of three signals, that is, a pressure HIGH signal for the reactor container, a water level LOW signal for the reactor and a radioactivity signal of the ventilation-air conditioner sytem on the side of air supply in the nuclear power plant. Thus, nuclear reactor buildings of the nuclear power plant are from the external atmosphere. Therefore, the radioactivity HIGH signal for switching to the emergency air conditioner system of the nuclear power plant is not actuated due to the loss of coolant accidents in the adjacent nuclear power plant. In addition, since the atmospheric temperature in the nuclear reactor building can be maintained by a cooling device disposed to the recycling system, reactor shutdown can be prevented. (I.S.)

Mitigation of corrosion product ingress into SG's

International Nuclear Information System (INIS)

Han, S.H.

1988-01-01

Design and operation experiences to mitigate corrosion product ingress into SGs in Korea nuclear power plants are briefly reviewed. Maintaining the feedwater pH above 9.6 with morpholine seems to contribute significantly to reduction of iron transport to SGs. Measured iron transport rates were 4.8 g/hr/100 MWe at pH 9.8 and 2.8 g/hr/100 MWe at 9.3, respectively. Removal of corrosion products through SG blowdown is very limited. Its removal efficiency at the higher pH plant was in the neighborhood of 10 %. In one of the Korea Nuclear Units, a large amount of sludge piles were found in the middle of tube bundles especially on the cold leg side. Damaged tubes were identified by the multi-frequency eddy current tests and plugged later during the refueling period. Intermittent blowdown-rate increase was tried to enhance ionic impurity removal through SG blowdown. Even though it was not effective against Na, removal other impurity was improved, resulting in prolonged condensate polisher operation periods by 1 - 2 days. Two-bed polisher design, a cation bed followed by a mixed bed, was chosen for future PWR plants to enhance corrosion product filtering capability of the polishers. Condensate pump discharge polishing and divided hot well polishing methods are currently in consideration. (Nogami, K.)

Initial substantial reduction in air dose rates of Cs origin and personal doses for residents owing to the Fukushima nuclear accident

International Nuclear Information System (INIS)

Yoshida, Hiroko; Saito, Junko; Hirasawa, Noriyasu; Kobayashi, Ikuo

2013-01-01

The initial substantial reduction in the air dose rate and personal dose equivalent [Hp(10)] for residents were compared between the Marumori and Kosugo regions for the period from September 2011 to September 2012 after the occurrence of the Fukushima nuclear accident. Marumori is a rural settlement, and Kosugo is a suburban city along a freeway. A similar tendency was observed in the Hp(10) results for Marumori residents and in the air dose rates for both regions: values dropped during the heavy snow season and a faster reduction in the air dose rate than the radioactive decay of 134 Cs and 137 Cs was observed after the snow had thawed. These reductions are considered to be caused by the weathering and/or migration of radionuclides down the soil column. However, neither a drop due to an accumulation of snow nor faster reduction was observed in Hp(10) for Kosugo residents. This discrepancy between the air dose rate and Hp(10) for Marumori and Kosugo residents might be caused by differences in their living environment. (author)

Suppressing hydrogen ingress during aqueous corrosion of CANDU Zr-2.5 Nb pressure tube material

International Nuclear Information System (INIS)

Elmoselhi, M.B.; Donner, A.; Brennenstuhl, A.; Warr, B.D.; Ellis, P.J.; Evans, D.W.

2002-01-01

As a result of their special properties, including low neutron cross-section and intrinsic corrosion resistance, Zr alloys are used in the fabrication of nuclear core components, particularly fuel cladding (in most reactor types) and also Zr-2.5 Nb pressure tubes in CANDU trademark (Canada Deuterium Uranium) reactors. Corrosion and H uptake during service can limit the life of these components. Therefore, remedial action may be appropriate to slow the H uptake rate and prolong the working life of these reactor components. This work has explored the possibility of reducing H uptake in pressure tube material by incorporating an inhibiting agent into the corrosion environment. Two approaches have been tested, depositing a thin metallic film on the initial oxide surface and adding an inhibiting agent to the solution. The latter approach appears more practical. Screening experiments were conducted in short-term (∝30 day) exposures in high temperature (340 C) aqueous out-reactor environments, simulating the CANDU trademark heat transport coolant with various chemistries. Compounds tested included aluminum acetate, aluminum nitrate, lithium nitrate, rhodium nitrate and yttrium nitrate. Comparison of results from the aluminum nitrate additives and aluminum acetate additives suggests that the nitrate anion is the effective ingredient for H ingress inhibition. The nitrate anion appears to reduce the rate of H ingress regardless of the associated cation. However, each cation appears to affect the rate of corrosion differently. These cations were found to be incorporated in the oxide film. (authors)

Radiological Emergency Preparedness after the Early Phase of an Accident : Focusing on an Air Contamination Event

International Nuclear Information System (INIS)

Jeong, Hyo Joon; Hwang, Won Tae; Kim, Eun Han; Han, Moon Hee

2010-01-01

Toxic materials in an urban area can be caused by a variety of events, such as accidental releases on industrial complexes, accidents during the transportation of hazardous materials and intentional explosions. Most governments around the world and their citizens have become increasingly worried about intentional accidents in urban area after the 911 terrorist attack in the United States of America. Even though there have been only a few attempted uses of Radiological Dispersal Devices (RDDs), accidental releases have occurred many times at commercial nuclear power plants and nuclear waste disposal sites. When an intentional release of radioactive materials occurs in an urban area, air quality for radioactive materials in the environment is of great importance to take action for countermeasures and environmental risk assessments. Atmospheric modeling is part of the decision making tasks and that it is particularly important for emergency managers as they often need to take actions quickly on very inadequate information(1). A simple model such as HOTSPOT required wind direction and source term would be enough to support the decision making in the early phase of an accident, but more sophisticated atmospheric modeling is required to adjust decontamination area and relocation etc after the early phase of an accidental event. In this study, we assume an explosion of 137 Cs using RDDs in the metropolitan area of Soul, South Korea. California Puff Model (CALPUFF) is used to calculate an atmospheric dispersion and transport for 137 Cs. Atmospheric dispersion and quantitative radiological risk analysis for 137 Cs were performed assuming an intentional explosion in the metropolitan area of Soul, South Korea after the early phase of emergency. These kinds of atmospheric modeling and risk analysis could provide a means for decision makers to take action on important issues such as the cleanup of the contaminated area and countermeasures to protect the public caused by

Consequences of the Chernobyl accident in France. Thematic sheets

International Nuclear Information System (INIS)

2006-01-01

This document proposes a set of commented maps, graphs and drawings which illustrate and describe various consequences of the Chernobyl accident in France, such as air contamination (scattering of radioactive particles emitted by the reactor explosion by the wind over thousands of kilometres, evolution of air contamination between April 30 and May 5 1986), ground deposits (influence of rain, heterogeneity of these deposits), contamination of farm products (relationship between the accident date and the deposit characteristics, variable decrease rate of contamination, faster decrease of farm product contamination that caesium radioactive decay since 1987, particular cases of some more sensitive products), health effects (low doses received by the French population, concerns about thyroid cancers)

XENON-133 IN CALIFORNIA, NEVADA, AND UTAH FROM THE CHERNOBYL ACCIDENT (JOURNAL VERSION)

Science.gov (United States)

The accident at the Chernobyl nuclear reactor in the USSR introduced numerous radioactive nuclides into the atmosphere, including the noble gas xenon-133. EPA's Environmental Monitoring Systems Laboratory, Las Vegas, NV, detected xenon-133 from the Chernobyl accident in air sampl...

Strength degradation of oxidized graphite support column in VHTR

International Nuclear Information System (INIS)

Park, Byung Ha; No, Hee Cheon

2010-01-01

Air-ingress events caused by large pipe breaks are important accidents considered in the design of Very High Temperature Gas-Cooled Reactors (VHTRs). A main safety concern for this type of event is the possibility of core collapse following the failure of the graphite support column, which can be oxidized by ingressed air. In this study, the main target is to predict the strength of the oxidized graphite support column. Through compression tests for fresh and oxidized graphite columns, the compressive strength of IG-110 was obtained. The buckling strength of the IG-110 column is expressed using the following empirical straight-line formula: σ cr,buckling =91.34-1.01(L/r). Graphite oxidation in Zone 1 is volume reaction and that in Zone 3 is surface reaction. We notice that the ultimate strength of the graphite column oxidized in Zones 1 and 3 only depends on the slenderness ratio and bulk density. Its strength degradation oxidized in Zone 1 is expressed in the following nondimensional form: σ/σ 0 =exp(-kd), k=0.114. We found that the strength degradation of a graphite column, oxidized in Zone 3, follows the above buckling empirical formula as the slenderness of the column changes. (author)

Use of a multi-species reactive transport model to simulate chloride ingress in mortar exposed to NaCl solution or sea-water

DEFF Research Database (Denmark)

Jensen, Mads Mønster; De Weerdt, K.; Johannesson, Björn

2015-01-01

Simulations of ion ingress in Portland cement mortar using a multi-species reactive mass transport model are compared with experimental test results. The model is an extended version of the Poisson–Nernst–Planck equations, accounting for chemical equilibrium. Saturated mortar samples were exposed...

THE AIRCRAFT ACCIDENT RATE IN CIVIL AVIATION DURING AIR TRANSPORT OPERATIONS AT THE AIRPORT

Directory of Open Access Journals (Sweden)

О. Запорожець

2012-04-01

Full Text Available The aircraft accident dates in civil aviation Ukraine and in republics of participants Agreement werecollected. The aircraft accident rate per 1 million flights was defined for civil aviation Ukraine and republicsof participants Agreement. Dynamics of aircraft accident rate was represented for civil aviation Ukraine.This dynamics was done for civil aviation of republics of participants Agreement and worldwide.

Exploring Civil Drone Accidents and Incidents to Help Prevent Potential Air Disasters

Directory of Open Access Journals (Sweden)

Graham Wild

2016-07-01

Full Text Available A recent alleged “drone” collision with a British Airways Airbus A320 at Heathrow Airport highlighted the need to understand civil Remotely Piloted Aircraft Systems (RPAS accidents and incidents (events. This understanding will facilitate improvements in safety by ensuring efforts are focused to reduce the greatest risks. One hundred and fifty two RPAS events were analyzed. The data was collected from a 10-year period (2006 to 2015. Results show that, in contrast to commercial air transportation (CAT, RPAS events have a significantly different distribution when categorized by occurrence type, phase of flight, and safety issue. Specifically, it was found that RPAS operations are more likely to experience (1 loss of control in-flight, (2 events during takeoff and in cruise, and (3 equipment problems. It was shown that technology issues, not human factors, are the key contributor in RPAS events. This is a significant finding, as it is contrary to the industry view which has held for the past quarter of a century that human factors are the key contributor (which is still the case for CAT. Regulators should therefore look at technologies and not focus solely on operators.

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.)

Microbial aerosol generation during laboratory accidents and subsequent risk assessment.

Science.gov (United States)

Bennett, A; Parks, S

2006-04-01

To quantify microbial aerosols generated by a series of laboratory accidents and to use these data in risk assessment. A series of laboratory accident scenarios have been devised and the microbial aerosol generated by them has been measured using a range of microbial air samplers. The accident scenarios generating the highest aerosol concentrations were, dropping a fungal plate, dropping a large bottle, centrifuge rotor leaks and a blocked syringe filter. Many of these accidents generated low particle size aerosols, which would be inhaled into the lungs of any exposed laboratory staff. Spray factors (SFs) have been calculated using the results of these experiments as an indicator of the potential for accidents to generate microbial aerosols. Model risk assessments have been described using the SF data. Quantitative risk assessment of laboratory accidents can provide data that can aid the design of containment laboratories and the response to laboratory accidents. A methodology has been described and supporting data provided to allow microbiological safety officers to carry out quantitative risk assessment of laboratory accidents.

Database on aircraft accidents

International Nuclear Information System (INIS)

Nishio, Masahide; Koriyama, Tamio

2013-11-01

for evaluating probability of aircraft crash are considered to be 4 for large fixed-wing aircraft, 35 for small fixed-wing aircraft, 1 for large bladed aircraft and 25 for small bladed aircraft. (2) The data of SDF aircraft accidents is based on newspapers and aircraft magazines. The total number of accidents is 42 comprised of 21 for large fixed-wing aircraft, 4 for small fixed-wing aircraft and 17 for bladed aircraft. The 23 accidents of the total fell into land. (3) The data of US aircraft accidents is based on newspapers and aircraft magazines. The total number of accidents is 16 comprised of 13 for fixed-wing aircraft and 3 for bladed aircraft. The 6 accidents of the total fell into land. (4) The instrument flight information on commercial aircrafts is based on 'Air transport statics' of Ministry of Land, Infrastructure, Transport and Tourism. The total number of takeoffs and landings is 30,685,564 and the total flight distance is 9,499,283,168 km. (author)

Sensitivity analysis on the zirconium ignition in a postulated SFP loss of coolant accident

International Nuclear Information System (INIS)

Park, Sanggil; Lee, Jaeyoung; Kim, Sun-ki; Chun, Tae-hyun; Bang, Je-geon

2016-01-01

From both SFP complete LOCA experiments, it was observed that zirconium alloy cladding temperature was abruptly increased at a certain point and the cladding was almost fully oxidized. To capture this phenomenon, the concept of air oxidation breakaway model was adopted in MELCOR code. This paper examines this air oxidation breakaway model by comparing the SFP project test data and MELCOR code calculation results by using this model. The air oxidation model parameters are slightly altered to see their sensitivities on the occurrence of the zirconium ignition. Through such sensitivity analysis, limitations of the air oxidation breakaway model are revealed in comparison to the actual zirconium ignition phenomenon during air ingress scenarios. In addition, ways to overcome the identified limitations of the air oxidation model are recommended to estimate better the zirconium ignition phenomenon in SFP sequences. In this paper, the zirconium ignition phenomenon was reviewed and the model to capture this phenomenon was investigated. The model is the air oxidation breakaway model in MELCOR code, and its sensitivity of the model parameters on the time to ignition was studied. From the sensitivity analysis, the slight change of model parameters induce the large variation of the time to ignition. The model itself includes its weakness to fully represent both the air oxidation breakaway phenomenon and the followed zirconium ignition behavior. Furthermore, this model considers no effect of N2 on the cladding degradation and its promoted exothermic heat release

Sensitivity analysis on the zirconium ignition in a postulated SFP loss of coolant accident

Energy Technology Data Exchange (ETDEWEB)

Park, Sanggil; Lee, Jaeyoung [Handong Global Univ., Pohang (Korea, Republic of); Kim, Sun-ki; Chun, Tae-hyun; Bang, Je-geon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

From both SFP complete LOCA experiments, it was observed that zirconium alloy cladding temperature was abruptly increased at a certain point and the cladding was almost fully oxidized. To capture this phenomenon, the concept of air oxidation breakaway model was adopted in MELCOR code. This paper examines this air oxidation breakaway model by comparing the SFP project test data and MELCOR code calculation results by using this model. The air oxidation model parameters are slightly altered to see their sensitivities on the occurrence of the zirconium ignition. Through such sensitivity analysis, limitations of the air oxidation breakaway model are revealed in comparison to the actual zirconium ignition phenomenon during air ingress scenarios. In addition, ways to overcome the identified limitations of the air oxidation model are recommended to estimate better the zirconium ignition phenomenon in SFP sequences. In this paper, the zirconium ignition phenomenon was reviewed and the model to capture this phenomenon was investigated. The model is the air oxidation breakaway model in MELCOR code, and its sensitivity of the model parameters on the time to ignition was studied. From the sensitivity analysis, the slight change of model parameters induce the large variation of the time to ignition. The model itself includes its weakness to fully represent both the air oxidation breakaway phenomenon and the followed zirconium ignition behavior. Furthermore, this model considers no effect of N2 on the cladding degradation and its promoted exothermic heat release.

Development of supporting system for emergency response to maritime transport accidents involving radioactive material

International Nuclear Information System (INIS)

Odano, N.; Matsuoka, T.; Suzuki, H.

2004-01-01

National Maritime Research Institute has developed a supporting system for emergency response of competent authority to maritime transport accidents involving radioactive material. The supporting system for emergency response has functions of radiation shielding calculation, marine diffusion simulation, air diffusion simulation and radiological impact evaluation to grasp potential hazard of radiation. Loss of shielding performance accident and loss of sealing ability accident were postulated and impact of the accidents was evaluated based on the postulated accident scenario. Procedures for responding to emergency were examined by the present simulation results

Measurement of the Portsmouth Gaseous Diffusion Plant criticality accident alarm

International Nuclear Information System (INIS)

Tayloe, R.W. Jr.; D'Aquila, D.M.; McGinnis, R.B.

1991-01-01

The nuclear criticality accident radiation alarm system installed at the Portsmouth Gaseous Diffusion Plant was tested extensively at critical facilities located at the Los Alamos National Laboratory. The ability of the neutron scintillator radiation detection units to respond to a minimum accident of concern as defined in Standard ANSI/ANS-83.-1986 was demonstrated. Detector placement and the established trip point are based on shielding calculations performed by the Oak Ridge National Laboratory and criticality specialists at the Portsmouth plant. Based on these experiments and calculations, a detector trip point of 5 mrad/h in air is used. Any credible criticality accident is expected to produce neutron radiation fields >5 mrad/h in air at one or more radiation alarm locations. Each radiation alarm location has a cluster of three detectors that employs a two-out-of-three alarm logic. Earlier work focused on testing the alarm logic latching circuitry. This work was directed toward measurements involving the actual audible alarm signal delivered

F4E R and D programme and results on in-vessel dust and tritium

International Nuclear Information System (INIS)

Le Guern, F.; Gulden, W.; Ciattaglia, S.; Counsell, G.; Bengaouer, A.; Brinster, J.; Dabbene, F.; Denkevitz, A.; Jordan, T.; Kuznetsov, M.; Porfiri, M.T.; Redlinger, R.; Roblin, Ph.; Roth, J.; Segre, J.; Sugiyama, K.; Tkatschenko, I.; Xu, Z.

2011-01-01

In a Tokamak vacuum vessel, plasma-wall interactions can result in the production of radioactive dust and H isotopes (including tritium) can be trapped both in in-vessel material and in dust. The vacuum vessel represents the most important confinement barrier to this radioactive material. In the event of an accident involving ingress of steam to the vacuum vessel, hydrogen could be produced by chemical reactions with hot metal and dust. Hydrogen isotopes could also be desorbed from in-vessel components, e.g. cryopumps. In events where an ingress of air to the vacuum vessel occurs, reaction of the air with hydrogen and/or dust therefore cannot be completely excluded. Due to the radiological risks highlighted by the safety evaluation studies for ITER in normal conditions (e.g. in-vessel maintenance chronic release) and accidental ones (e.g. challenge of vacuum vessel tightness in the event of a hydrogen/dust explosion with air), limitations on the accumulation of dust and tritium in the vacuum vessel are imposed as well as controls over the maximum extent of the quantity of accidental air ingress. ITER IO has defined a strategy for the control of in-vessel dust and tritium inventories below the safety limits based primarily on the measurement and removal of dust and tritium. In this context, this paper will report on the efforts under F4E responsibility to develop a number of the new ITER baseline systems. In particular this paper, after a review of safety constraints and ITER strategy, provides the status of: (1) tasks being launched on diagnostics for in-vessel dust inventory measurement, (2) experiments to enrich the data about the effectiveness of desorption of tritium from Be at 350 o C (divertor baking aiming to release significant amount of tritium trapped in Be co-deposit), (3) on-going R and D programme (experimental and numerical simulation) at FZK, CEA and ENEA on in-vacuum vessel H2 dust explosion.

The question of the possibility of being killed or injured in a transport accident

Directory of Open Access Journals (Sweden)

Kazakov N.A.

2017-02-01

Full Text Available the article presents an analysis of statistics, which shows that the probability to get in a road accident is higher and it is most likely to die in an air transport accident, by becoming a party of it.

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

PARC (Plutonium Accident Resistant Container) project

International Nuclear Information System (INIS)

Andersen, J.A.

1978-01-01

Response by the US Nuclear Regulatory Commission (NRC) to a public law limiting the air transport of plutonium resulted in a new Qualification Criteria and led to the PARC project. The PARC project resulted in the design, development, and certification testing of a crashworthy air transportable radioactive materials package for certification by the NRC. This package, identified by the NRC as the Model PAT-1 Package (PAT = Plutonium Air Transportable), has a mass capacity of 2 kg of PuO 2 and a thermal capacity of 25 watts; the internal volume of the inner containment vessel is 1460 cm 3 ; the package dimensions are 62 cm (24 - 1/2 in.) O.D. x 108 cm (42 - 1/2 in.) length and a mass of 227 kg (500 lbs). The design rationale for very high energy absorption (impact, crush, puncture, and slash protection) with residual high-level fire protection, resulted in a reasonably small air-transportable package, advancing the packaging state-of-the-art. Optimization design interactions were utilized in the areas of impact energy absorption and stress and thermal analysis. Package test results are presented in relation to the containment acceptance criteria of the Qualification Criteria and ANSI N 14.5 leaktight standards and the IAEA Safety Series No. 6 A2 radioactive material release quantities. Acceptability of the pre-accident and post-accident package configuration with respect to shielding and criticality standards are ascertained

THE USE OF AVIATION ACCIDENT INVESTIGATION REPORTS AS EVIDENCE IN COURT

Directory of Open Access Journals (Sweden)

Sorana POP PĂUN

2016-05-01

Full Text Available Air transport is an essential part of the international society, constituting a liaison between people and continents and an important contributor to the world economy and globalization. Aircraft operation has grown in complexity needing for a safety level to be maintained and constantly grown. Along with the development of the aviation industry, the legal system in the aviation field has registered significant challenges, one of them being the claims related to air crashes which are contested. The investigation process of an accident or incident has become not only important for the safety of operations but also to the establishment of legal fault and blame. The article proposes to present the principles of conducting and accident and incident investigation, the value of the report and new developments in relation to the recent case law on the use of the accident investigation report in Court.

The Transit Ingress and the Tilted Orbit of the Extraordinarily Eccentric Exoplanet HD 80606b

Science.gov (United States)

Winn, Joshua N.; Howard, Andrew W.; Johnson, John A.; Marcy, Geoffrey W.; Gazak, J. Zachary; Starkey, Donn; Ford, Eric B.; Colon, Knicole D.; Reyes, Francisco; Nortmann, Lisa;

High-temperature gas-cooled reactor safety studies. Progress report for January 1, 1974--June 30, 1975

International Nuclear Information System (INIS)

Cole, T.E.; Sanders, J.P.; Kasten, P.R.

1977-07-01

Progress is reported in the following areas: systems and safety analysis; fission product technology; primary coolant technology; seismic and vibration technology; confinement components; primary system materials technology; safety instrumentation; loss of flow accident analysis using HEATUP code; use of coupled-conduction-convection model for core thermal analysis; development of multichannel conduction-convection program HEXEREI; cooling system performance after shutdown; core auxiliary cooling system performance; development of FLODIS code; air ingress into primary systems following DBDA; performance of PCRV thermal barrier cover plates; temperature limits for fuel particle coating failure; tritium distribution and release in HTGR; energy release to PCRV during DBDA; and mathematical models for HTGR reactor safety studies

Effects of New MCCI Models Implemented in MAAP5 on MCCI in NPP Application

Energy Technology Data Exchange (ETDEWEB)

Kim, Byung Jo; Ryu, In Chul; Moon, Young Tae [KEPCO Engineering and Construction Co., Gimcheon (Korea, Republic of)

2016-10-15

In the case of severe accident, the containment integrity is a key issue, because the containment is the final barrier against the release of fission products into the environment. Therefore the progression of Molten Core-Concrete Interaction (MCCI) takes significant importance and has a role to threaten the integrity of the containment and the subsequent release of radioactivity in the case of core melting leading to vessel melt-through in a way of pressurization due to heat and steam/gas generation, generation of combustible gases, release of radioactive aerosols, and containment liner failure. To evaluate MCCI phenomenon under the severe accident condition, Modular Accident Analysis Program (MAAP) which developed by Electric Power Research Institute (EPRI) has been widely used in the utility field. This paper introduces the features of new MCCI models realized in MAAP5 and the effects of new models on MCCI consequences for a large dry type PWR plant application with the pre-flooding accident management strategy. MAAP5.03 calculation using new MCCI models were carried out for the NPP application. Among the known cooling mechanism during MCCI, dominant two mechanisms - water ingression and melt eruption are incorporated in MAAP5. Regarding the water ingression model two bounding value of user-input parameter is applied in parametric model. Comparing with the mechanistic water ingression model, the bounding FCHF value case show a limiting result of ablation and quenching. When one employ the parametric water ingression model, a precisely review and choose FCHF value is essential to reduce the user-dependency in MCCI evaluation, however mechanistic water ingression model derived in MAAP5 clearly provides the realistic and best-estimate heat removal through the upper crust. Melt eruption and resultant formation of particle bed on top of the upper crust is tested using MAAP5. Once MCCI starts the formation of particle bed is apparent due to off gases from the

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

Drivers of accident preparedness and safety: evidence from the RMP Rule

International Nuclear Information System (INIS)

Kleindorfer, Paul R.; Elliott, Michael R.; Wang Yanlin; Lowe, Robert A.

2004-01-01

This paper provides an overview of recent results derived from the accident history data collected under 112(r) of the Clean Air Act Amendments (the Risk Management Program (RMP) Rule) covering the period 1994-2000, together with a preliminary assessment of the effectiveness of the RMP Rule as a form of Management System Regulation. These were undertaken at the University of Pennsylvania by a multi-disciplinary team of economists, statisticians and epidemiologists with the support of the US Environmental Protection Agency and its Office of Emergency Prevention, Preparedness and Response (OEPPR, formerly CEPPO). Section 112(r) of the Clean Air Act Amendments of 1990 requires that chemical facilities in the US that had on premises more than specified quantities of toxic or flammable chemicals file a 5-year history of accidents. The initial data reported under the RMP Rule covered roughly the period from mid-1994 through mid-2000, and provided details on economic, environmental and acute health affects resulting from accidents at some 15,000 US chemical facilities for this period. This paper reviews research based on this data. The research is in the form of a retrospective cohort study that considers the statistical associations between accident frequency and accident severity at covered facilities (the outcome variables of interest) and a number of facility characteristics (the available predictor variables provided by the RMP Rule), the latter including such facility characteristics as size, hazardousness, financial characteristics of parent company-owners of the facility, regulatory programs in force at the facility, and host community characteristics for the surrounding county in which the facility was located, as captured in the 1990 Census. Among the findings reviewed are: (1) positive associations with (a measure of) facility hazardousness and accident, injury and economic costs of accidents; (2) positive (resp., negative) associations between accident

Gamma-spectrometric examination of hot particles emitted during the Chernobyl accident

International Nuclear Information System (INIS)

Balashazy, I.; Szabadine-Szende, G.; Loerinc, M.; Zombori, P.

1987-05-01

Ge(Li) gamma-spectrometric examination of hot particles prepared from air filtered dust of Budapest air after the Chernobyl accident is presented. The method of separating hot particles is described and their concentration in the air is determined. The radioactive isotope composition of hot particles is discussed and compared with that of dust samples. Finally, the inhalation probability and radiation burden of hot particles are evaluated. (author)

Durable concrete for a waste repository - Measurement of ionic ingress

International Nuclear Information System (INIS)

Feldman, R.F.; Beaudoin, J.J.; Philipose, K.E.

1990-01-01

A waste repository for the below ground disposal of low level radioactive waste is planned at Chalk River Nuclear Laboratories. It relies greatly on the durability of concrete for the required 500 year service life. A research program to design durable concrete and predict its service life is in progress. The degradation of the concrete depends to a large extent on the rate of ingress of corrosive agents. Penetration of chloride and sulfate ions are particularly relevant. Twenty mix formulations were developed to create various types and qualities of concrete, and to study their behavior in different site environmental conditions. A total of 1,000 concrete specimens are being exposed at 20C and 45C to 25 different combinations of the corrosive agents including CO 2 . Procedures to measure the ionic profiles and to determine the factors controlling diffusion of the ions in the various concretes have been developed. Results of selected concrete systems exposed to chloride and sulfate solutions for 1 year are presented and discussed in terms of pore structure and permeability parameters of the concrete

Criticality accident dosimetry with ESR spectroscopy.

Science.gov (United States)

d'Errico, F; Fattibene, P; Onori, S; Pantaloni, M

1996-01-01

The suitability of the ESR alanine and sugar detectors for criticality accident dosimetry was experimentally investigated during an intercomparison of dosimetry techniques. Tests were performed irradiating detectors both free-in-air and on-phantom during controlled critcality excursions at the SILENE reactor in Valduc, France. Several grays of absorbed dose were imparted in neutron gamma-ray fields of various relative intensities and spectral distributions. Analysed results confirmed the potential of these systems which can immediately provide an acute dose assessment with an average underestimate of 30%in the various fields. This performance allows for the screening of severely exposed individuals and meets the IAEA recommendations on the early estimate of accident absorbed doses.

Study of the ruthenium fission-product behavior in the containment, in the case of a nuclear reactor severe accident

International Nuclear Information System (INIS)

Mun, Ch.

2007-03-01

Ruthenium tetroxide is an extremely volatile and highly radio-toxic species. During a severe accident with air ingress in the reactor vessel, ruthenium oxides may reach the reactor containment building in significant quantities. Therefore, a better understanding of the RuO 4 (g) behaviour in the containment atmosphere is of primary importance for the assessment of radiological consequences, in the case of potential releases of this species into the environment. A RuO 4 (g) decomposition kinetic law was determined. Steam seems to play a catalytic role, as well as the presence of ruthenium dioxide deposits. The temperature is also a key parameter. The nature of the substrate, stainless steel or paint, did not exhibit any chemical affinities with RuO 4 (g). This absence of reactivity was confirmed by XPS analyses, which indicate the presence of the same species in the Ru deposits surface layer whatever the substrates considered. It has been concluded that RuO 4 (g) decomposition corresponds to a bulk gas phase decomposition. The ruthenium re-volatilization phenomenon under irradiation from Ru deposits was also highlighted. An oxidation kinetic law was determined. The increase of the temperature and the steam concentration promote significantly the oxidation reaction. The establishment of Ru behavioural laws allowed making a modelling of the Ru source term. The results of the reactor calculations indicate that the values obtained for 106 Ru source term are closed to the reference value considered currently by the IRSN, for 900 MWe PWR safety analysis. (author)

Evaluation of carbons exposed to the Three Mile Island accident

International Nuclear Information System (INIS)

Deitz, V.R.; Romans, J.B.; Bellamy, R.R.

1981-01-01

One of the lines of defense that served to mitigate the radiological effects of the accident at Three Mile Island was the activated carbon installed in ventilation air flows. Filters in the Auxiliary and Fuel Handling Buildings of Unit 2 adsorbed tens to hundreds of curies of iodine-131, preventing the release to the environment. The carbon exposed to the accident has been replaced and the spent carbon has been analyzed in the laboratory. Independent analyses were performed for the two filter trains in both the Auxiliary and Fuel Handling Buildings, replaced at various times after the accident. The results of these analyses are compared to new (unexposed) carbons

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

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

Effects of the Fukushima Daiichi nuclear accident on goshawk reproduction

Science.gov (United States)

Murase, Kaori; Murase, Joe; Horie, Reiko; Endo, Koichi

2015-01-01

Although the influence of nuclear accidents on the reproduction of top predators has not been investigated, it is important that we identify the effects of such accidents because humans are also top predators. We conducted field observation for 22 years and analysed the reproductive performance of the goshawk (Accipiter gentilis fujiyamae), a top avian predator in the North Kanto area of Japan, before and after the accidents at the Fukushima Daiichi nuclear power plant that occurred in 2011. The reproductive performance declined markedly compared with the pre-accident years and progressively decreased for the three post-accident study years. Moreover, it was suggested that these declines were primarily caused by an increase in the air dose rate of radio-active contaminants measured under the nests caused by the nuclear accidents, rather than by other factors. We consider the trends in the changes of the reproductive success rates and suggest that internal exposure may play an important role in the reproductive performance of the goshawk, as well as external exposure. PMID:25802117

Simulation of hypothetical criticality accidents involving homogeneous damp low-enriched UO2 powder systems

International Nuclear Information System (INIS)

Basoglu, B.; Brewer, R.W.; Haught, C.F.; Hollenbach, D.F.; Wilkinson, A.D.; Dodds, H.L.; Pasqua, P.F.

1994-01-01

This paper describes the development of a computer model for predicting the excursion characteristics of a postulated, hypothetical, critically accident involving a homogeneous mixture of low-enriched UO 2 powder and water contained in a cylindrical blender. The model uses point neutronics coupled with simple lumped-parameter thermal-hydraulic feedback. The temperature of the system is calculated using a simple time-dependent energy balance where two extreme conditions for the thermal behavior of the system are considered, which bound the real life situation. Using these extremes, three different models are developed. To evaluate the models, the authors compared the results with the results of the POWDER code, which was developed by the Commissariat a l'Energie Atomique/United Kingdom Atomic Energy Authority (CEA/UKAEA) for damp powder systems. The agreement in these comparisons is satisfactory. Results of the excursion studies in this work show that approximately 10 19 fissions occur as a result of accidental water ingress into powder blenders containing 5,000 kg of low-enriched (5%) UO 2 powder

Agricultural implications of the Fukushima nuclear accident

International Nuclear Information System (INIS)

Nakanishi, Tomoko M.

2016-01-01

More than 4 years has passed since the accident at the Fukushima Nuclear Power Plant. Immediately after the accident, 40 to 50 academic staff of the Graduate School of Agricultural and Life Sciences at the University of Tokyo created an independent team to monitor the behavior of the radioactive materials in the field and their effects on agricultural farm lands, forests, rivers, animals, etc. When the radioactive nuclides from the nuclear power plant fell, they were instantly adsorbed at the site where they first touched; consequently, the fallout was found as scattered spots on the surface of anything that was exposed to the air at the time of the accident. The adsorption has become stronger over time, so the radioactive nuclides are now difficult to remove. The findings of our study regarding the wide range of effects on agricultural fields are summarized in this report

Study of the ruthenium fission-product behavior in the containment, in the case of a nuclear reactor severe accident; Etude du comportement du produit de fission ruthenium dans l'enceinte de confinement d'un reacteur nucleaire, en cas d'accident grave

Energy Technology Data Exchange (ETDEWEB)

Mun, Ch

2007-03-15

Ruthenium tetroxide is an extremely volatile and highly radio-toxic species. During a severe accident with air ingress in the reactor vessel, ruthenium oxides may reach the reactor containment building in significant quantities. Therefore, a better understanding of the RuO{sub 4}(g) behaviour in the containment atmosphere is of primary importance for the assessment of radiological consequences, in the case of potential releases of this species into the environment. A RuO{sub 4}(g) decomposition kinetic law was determined. Steam seems to play a catalytic role, as well as the presence of ruthenium dioxide deposits. The temperature is also a key parameter. The nature of the substrate, stainless steel or paint, did not exhibit any chemical affinities with RuO{sub 4}(g). This absence of reactivity was confirmed by XPS analyses, which indicate the presence of the same species in the Ru deposits surface layer whatever the substrates considered. It has been concluded that RuO{sub 4}(g) decomposition corresponds to a bulk gas phase decomposition. The ruthenium re-volatilization phenomenon under irradiation from Ru deposits was also highlighted. An oxidation kinetic law was determined. The increase of the temperature and the steam concentration promote significantly the oxidation reaction. The establishment of Ru behavioural laws allowed making a modelling of the Ru source term. The results of the reactor calculations indicate that the values obtained for {sup 106}Ru source term are closed to the reference value considered currently by the IRSN, for 900 MWe PWR safety analysis. (author)

Radiological emergency: road map for radiation accident victim transport

International Nuclear Information System (INIS)

Costa, V.S.G.; Alcantara, Y.P.; Lima, C.M.A.; Silva, F. C. A. da

2017-01-01

During a radiological or nuclear emergency, a number of necessary actions are taken, both within the radiation protection of individuals and the environment, involving many institutions and highly specialized personnel. Among them it is possible to emphasize the air transportation of radiation accident victims.The procedures and measures for the safe transport of these radiation accident victims are generally the responsibility of the armed forces, specifically the Aeronautics, with the action denominated 'Aeromedical Military Evacuation of Radiation Accident Victims'. The experience with the Radiological Accident of Goiânia demonstrated the importance of adequate preparation and response during a radiological emergency and the need for procedures and measures with regard to the transport of radiation victims are clearly defined and clearly presented for the effectiveness of the actions. This work presents the necessary actions for the transport of radiation accident victim during a radiological emergency, through the road map technique, which has been widely used in scientific technical area to facilitate understanding and show the way to be followed to reach the proposed objectives

Investigation of hot air balloon fatalities.

Science.gov (United States)

McConnell, T S; Smialek, J E; Capron, R G

1985-04-01

The rising popularity of the sport of hot air ballooning has been accompanied by several recent incidents, both in this country and other parts of the world, where mechanical defects and the improper operation of balloons have resulted in several fatalities. A study was conducted to identify the location and frequency of hot air ballooning accidents. Furthermore, the study attempted to identify those accidents that were the result of improper handling on the part of the balloon operators and those that were related to specific defects in the construction of the balloon. This paper presents a background of the sport of hot air ballooning, together with an analysis of the construction of a typical hot air balloon, pointing out the specific areas where defects may occur that could result in a potential fatal balloon crash. Specific attention is given to the two recent balloon crashes that occurred in Albuquerque, N.M., hot air balloon capital of the world, and that resulted in multiple fatalities.

Analysis of Sertraline in Postmortem Fluids and Tissues in 11 Aviation Accident Victims

Science.gov (United States)

2012-11-01

likely undergoes significant postmortem redistribution. 17. Key Words 18. Distribution Statement Forensic Toxicology , Sertraline, Norsertraline... Toxicology .. Forensic Sci Int,.142:.75-100.(2004) . 29 .. Skopp,.G ..Postmortem.Toxicology .. Forensic Sci Med Pathol,.6:.314-25.(2010) . ... toxicological . analysis. on. specimens.from.….aircraft.accident.fatalities”.and.“in- vestigate.….general.aviation.and.air.carrier.accidents. and. search

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

The reactor accident of Chernobyl

International Nuclear Information System (INIS)

Koenig, L.A.; Schuettelkopf, H.; Erat, S.; Fessler, H.; Hempelmann, S.; Maurer, K.; Pimpl, M.; Radziwill, A.

1986-08-01

The contamination, caused by the radioactivity released during the reactor accident of Chernobyl was measured in samples taken in the environment of the Karlsruhe Nuclear Research Center. The radioactivity was determined in air, fodder, milk, vegetables, other plants, foodstuffs, soil, precipitations, drinking water, sludge and other samples. Results of measurements are reported which were received with considerably more than 1000 samples. The evaluation of the data will be presented in KfK 4140. (orig.) [de

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

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

Occupational Radiation Protection in Severe Accident Management

International Nuclear Information System (INIS)

2015-01-01

As an early response to the Fukushima Daiichi NPP accident, the Information System on Occupational Exposure (ISOE) Bureau decided to focus on the following issues as an initial response of the joint program after having direct communications with the Japanese official participants in April 2011: - Management of high radiation area worker doses: It has been decided to make available the experience and information from the Chernobyl accident in terms of how emergency worker / responder doses were legally and practically managed, - Personal protective equipment for highly-contaminated areas: It was agreed to collect information about the types of personnel protective equipment and other equipment (e.g. air bottles, respirators, air-hoods or plastic suits, etc.), as well as high-radiation area worker dosimetry use (e.g. type, number and placement of dosimetry) for different types of emergency and high-radiation work situations. Detailed information was collected on dose criteria which are used for emergency workers /responders and their basis, dose management criteria for high dose/dose rate areas, protective equipment which is recommended for emergency workers / responders, recommended individual monitoring procedures, and any special requirement for assessment from the ISOE participating nuclear utilities and regulatory authorities and made available for Japanese utilities. With this positive response of the ISOE official participants and interest in the situation in Fukushima, the Expert Group on Occupational Radiation Protection in Severe Accident Management (EG-SAM) was established by the ISOE Management Board in May 2011. The overall objective of the EG-SAM is to contribute to occupational exposure management (providing a view on management of high radiation area worker doses) within the Fukushima plant boundary with the ISOE participants and to develop a state-of-the-art ISOE report on best radiation protection management practices for proper radiation

Contaminant ingress into multizone buildings: An analytical state-space approach

KAUST Repository

Parker, Simon

2013-08-13

The ingress of exterior contaminants into buildings is often assessed by treating the building interior as a single well-mixed space. Multizone modelling provides an alternative way of representing buildings that can estimate concentration time series in different internal locations. A state-space approach is adopted to represent the concentration dynamics within multizone buildings. Analysis based on this approach is used to demonstrate that the exposure in every interior location is limited to the exterior exposure in the absence of removal mechanisms. Estimates are also developed for the short term maximum concentration and exposure in a multizone building in response to a step-change in concentration. These have considerable potential for practical use. The analytical development is demonstrated using a simple two-zone building with an inner zone and a range of existing multizone models of residential buildings. Quantitative measures are provided of the standard deviation of concentration and exposure within a range of residential multizone buildings. Ratios of the maximum short term concentrations and exposures to single zone building estimates are also provided for the same buildings. © 2013 Tsinghua University Press and Springer-Verlag Berlin Heidelberg.

Comparative analysis of the countermeasures taken to mitigate exposure of the public to radioiodine following the Chernobyl and Fukushima accidents: lessons from both accidents.

Science.gov (United States)

Uyba, Vladimir; Samoylov, Alexander; Shinkarev, Sergey

2018-04-01

In the case of a severe radiation accident at a nuclear power station, the most important radiation hazard for the public is internal exposure of the thyroid to radioiodine. The purposes of this paper were (i) to compare countermeasures conducted (following the Chernobyl and Fukushima accidents) aimed at mitigation of exposure to the thyroid for the public, (ii) to present comparative estimates of doses to the thyroid and (iii) to derive lessons from the two accidents. The scale and time of countermeasures applied in the early phase of the accidents (sheltering, evacuation, and intake of stable iodine to block the thyroid) and at a later time (control of 131I concentration in foodstuffs) have been described. After the Chernobyl accident, the estimation of the thyroid doses for the public was mainly based on direct thyroid measurements of ~400 000 residents carried out within the first 2 months. The highest estimates of thyroid doses to children reached 50 Gy. After the Fukushima accident, the estimation of thyroid doses was based on radioecological models due to a lack of direct thyroid measurements (only slightly more than 1000 residents were measured). The highest estimates of thyroid doses to children were a few hundred mGy. Following the Chernobyl accident, ingestion of 131I through cows' milk was the dominant pathway. Following the Fukushima accident, it appears that inhalation of contaminated air was the dominant pathway. Some lessons learned following the Chernobyl and Fukushima accidents have been presented in this paper.

Safety analysis of RA reactor operation, I-III, Part III - Environmental effect of the maximum credible accident

International Nuclear Information System (INIS)

Raisic, N.

1963-02-01

Maximum credible accident at the RA reactor would consider release of fission products into the environment. This would result from fuel elements failure or meltdown due to loss of coolant. The analysis presented in this report assumes that the reactor was operating at nominal power at the moment of maximum possible accident. The report includes calculations of fission products activity at the moment of accident, total activity release during the accident, concentration of radioactive material in the air in the reactor neighbourhood, and the analysis of accident environmental effects

Environmental radioactivity and dose evaluation in Taiwan after the Chernobyl accident

International Nuclear Information System (INIS)

Chung, C.E.

1989-01-01

A substantial increase in fission product activity was observed in various environmental samples taken in Taiwan after the Chernobyl accident. The concentration of long-lived fission products in air above ground, precipitation, grass, vegetation and milk were monitored in the next 7 wk. The individual effective dose equivalent committed by the first year of exposure and intake following the accident were evaluated. Average individual doses for the population in Taiwan are estimated at 0.9 microSv due to global fallout from the Chernobyl accident. This value is lower than that reported in neighboring countries in the Far East and poses no increased health impact to the public in Taiwan

Plutonium accident resistant container project

International Nuclear Information System (INIS)

Andersen, J.A.

1978-05-01

The PARC (plutonium accident resistant container) project resulted in the design, development, and certification testing of a crashworthy air-transportable plutonium package (shipping container) for certification by the USNRC. This PAT-1 (plutonium air transportable) package survives a very severe sequential test program of impact, crush, puncture, slash, burn, and water immersion. There is also an individual hydrostatic pressure test. The package has a payload mass capacity of 2 kg of PuO2 and a thermal capacity of 25 watts. The design rationale for very high energy absorption (impact, crush, puncture, and slash protection) with residual high-level fire protection, resulted in a reasonalby small air-transportable package, advancing the packaging state-of-art. Optimization design iterations were utilized in the areas of impact energy absorption and stress and thermal analysis. Package test results are presented in relation to radioactive materials containment acceptance criteria, shielding and criticality standards

EPRI nuclear fuel-cycle accident risk assessment

International Nuclear Information System (INIS)

Anon.

1981-01-01

The present results of the nuclear fuel-cycle accident risk assessment conducted by the Electric Power Research Institute show that the total risk contribution of the nuclear fuel cycle is only approx. 1% of the accident risk of the power plant; hence, with little error, the accident risk of nuclear electric power is essentially that of the power plant itself. The power-plant risk, assuming a very large usage of nuclear power by the year 2005 is only approx. 0.5% of the radiological risk of natural background. The smallness of the fuel-cycle risk relative to the power-plant risk may be attributed to the lack of internal energy to drive an accident and the small amount of dispersible material. This work aims at a realistic assessment of the process hazards, the effectiveness of confinement and mitigation systems and procedures, and the associated likelihood of errors and the estimated size of errors. The primary probabilistic estimation tool is fault-tree analysis, with the release source terms calculated using physicochemical processes. Doses and health effects are calculated with CRAC (Consequences of Reactor Accident Code). No evacuation or mitigation is considered; source terms may be conservative through the assumption of high fuel burnup (40,000 MWd/t) and short cooling period (90 to 150 d); high-efficiency particulate air filter efficiencies are derived from experiments

Non-destructive vacuum decay method for pre-filled syringe closure integrity testing compared with dye ingress testing and high-voltage leak detection.

Science.gov (United States)

Simonetti, Andrea; Amari, Filippo

2015-01-01

solution, preventing possible prefilled syringe plunger movement during container closure integrity testing execution, is presented as well. The growing need to meet sterile drug products' regulatory, quality, and safety expectations has progressively driven new developments and improvements both in container closure integrity testing methods and in the respective equipment, over the last years. Indeed, container closure integrity testing establishes the container closure system capability to provide required protection to the drug product and to demonstrate maintenance of product sterility over its shelf life. This article describes the development of four container closure integrity testing approaches for the evaluation of glass prefilled syringe closure integrity, including two destructive (pharmacopoeial and Novartis specific dye ingress test) and two non-destructive (vacuum decay and high-voltage leak detection) methods. The important finding from the validation of comparative studies was that the vacuum decay method resulted in the most effective, reliable and repeatable detection of defective samples, whether the defect was exposed to sterile water, to drug product, or to air. Complete sets of known defects were created for this purpose (5 μm, 10 μm, 20 μm certified leakages by laser drilled holes and capillary tubes). All investigations and studies were conducted at Bonfiglioli Engineering S.r.l. (Vigarano Pieve, Ferrara, Italy) and at Novartis Vaccines (Sovicille, Siena, Italy). © PDA, Inc. 2015.

The External Cost Evaluation of the Nuclear Severe Accident Using CVM

International Nuclear Information System (INIS)

Lee, Yong Suk; Lee, Byung Chul

2006-01-01

The external cost of energy can be defined as 'the cost not included in the energy market price', such as air pollution, noise, etc. Within the evaluation of the external cost of nuclear energy, the estimation of the external cost of severe accident is one of the major topics to be addressed. For the evaluation of the external cost of severe accident, the effect of risk aversion of the public against the severe accident must be addressed, because people are more concerned about low probability - high consequence events than about high probability - low consequence events having the same mean damage. It is generally recognized that there is a discrepancy between the social acceptability of the risk and the average monetary value which corresponds in principle to the compensation of the consequences for each individual of the population affected by the accident. In this paper, the CVM (Contingent Valuation Method) is used to integrate the risk aversion in the external costs of nuclear severe accidents in Korea

[Research on accidents in a tire-producing plant].

Science.gov (United States)

Mete, R; Sabatucci, A

1989-09-30

In the autumn of 1987 the U.S.L. health service (prevention, hygiene and occupational safety section) began a study about the accidents in a firm manufacturing tyres, placed in its own area. The retrospective enquiry starts from the analysis of typology, diffusion and seriousness of occupational accidents. The firm's accident register has been analyzed and integrated with other necessary information provided by the firm, by I.N.A.I.L. and by the air force metereological service. The study has been carried out on data concerning the following years: 1984-1985-1986. The accidents considered, implied absence from work and were divided as follows: for absence up till 3 days (in franchise), and more than 3 days (indemnified), applying the average value calculated on one year of the three analyzed. Every accident has been analyzed per year, month, day, hour of event. According to the classes: circumstances, kind of lesion, site of lesion, period of absence from work. The indices of: frequency, seriousness, incidence, mean duration have been calculated. The average monthly values of temperature: max and min. of the area and to the average monthly amount of processed elastomer (rate of production). The statistics we obtained, justified the study and showed the operative solution. The aspect of sanitary education and the general psychological aspect regarding the accident have been considered. Moreover the general operative solutions for the firm and specific ones for every department and for every position have been shown and faced up to. In this way, according to the risks that have emerged from the enquiries on previous accidents and thanks to direct inspection. it was possible to prevent accidents.

Evaluation of major polluting accidents in China-Results and perspectives

International Nuclear Information System (INIS)

Hou Yu; Zhang Tianzhu

2009-01-01

Lessons learnt from accidents are essential sources for updating state-of-the-art requirements in pollution accident prevention. To improve this input in the People's Republic of China in a systematic way, a database for collecting and evaluating major pollution accidents is being established. This is being done in co-operation with Chinese Society for Environment Sciences and other national Institutions. At the time of writing over 80 major events from 2002-2006 have been collected. In this paper, a summary evaluation on the major polluting events in China from 2002 to 2006 is presented and some basic lessons drawn shown. There is no a systematic pollution accident notification system currently in China. The results from root cause analysis underline the importance of emergency measures, maintenance, human factor issues and the role of safety organization. Chronic pollution, especially water pollution and air pollution should be paid the same attention as the sudden pollution. It is important to keep in mind that collecting information from major accidents represents a small percentage of the actual number of events taking place.

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)

Development of Very High Temperature Reactor Technology

International Nuclear Information System (INIS)

Lee, Won Jae; Noh, J. M.; Kim, Y. H.

2009-04-01

For an efficient production of nuclear hydrogen, the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature and the interfacing system for the hydrogen production are required. We have developed various evaluation technologies for the performance and safety of VHTR through the accomplishment of this project. First, to evaluate the performance of VHTR, a series of analyses has been performed such as core characteristics at 950 .deg. C, applicability of cooled-vessel, intermediate loop system and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC and the analysis of the risk/performance-informed method, VHTR safety evaluation has been also performed. In addition, various design analysis codes have been developed for a nuclear design, system loop design, system performance analysis, air-ingress accident analysis, fission product/tritium transport analysis, graphite structure seismic analysis and hydrogen explosion analysis, and they are being verified and validated through a lot of international collaborations

Development of Very High Temperature Reactor Design Technology

International Nuclear Information System (INIS)

Lee, Won Jae; Noh, J. M.; Kim, K. S.

2007-05-01

To develop design technologies for the VHTR (Very High Temperature Gas-cooled Reactor) of 950 .deg. C outlet temperature for an efficient hydrogen production, key studies were performed, which include evaluation technology for the performance and safety of VHTR, and development of design and analysis codes. First, to evaluate the performance of VHTR, a series of analyses has been carried out for core characteristics at 950 .deg. C, cooled-vessel adopting internal flow path through the graphite structure, compact heat exchanger with periodic channel configuration, intermediate loop system, risk/performance-informed method, and high temperature structural integrity. Through the analyses of major accidents such as HPCC and LPCC, safety evaluation of both VHTR and RCCS has been also performed. In addition, prototype codes have been developed for a nuclear design, system loop design, system performance analysis, air-ingress accident analysis, fission product/tritium transport analysis, graphite structure seismic analysis and hydrogen explosion analysis, and they are being verified and validated through a lot of international collaborations

Highlights from the IAEA coordinated research programme on fuel performance and fission product data

International Nuclear Information System (INIS)

Nabielek, H.; Schenk, W.; Verfondern, K.

1996-01-01

Seven countries are cooperating with the objectives (i) to document the status of the experimental data base and of the predictive methods for Gas-Cooled Reactor fuel performance and fission product behaviour; (ii) to verify and validate methods in fuel performance and fission product retention prediction. These countries are China, France, Germany, Japan, Russia, USA and the UK. Duration of the programme is 1993-96. The technology areas addressed in this IAEA Coordinated Research Programme are: Fuel design and manufacture, Normal operation fuel performance and fission product behaviour, Accident condition fuel performance and fission product behaviour, -core heatup, -fast transients, -oxidising conditions (water and air ingress), Plateout, re-entrainment of plateout, fission product behaviour in the reactor building, and Performance of advanced fuels. Work performed so far has generated a 300-page draft document with important information for normal operations (Germany, Japan, China, Russia) and accident conditions (USA, Japan, Germany, Russia) and, additionally, a special chapter on advanced fuels (Japan). (author)

Helicopter emergency medical services response to equestrian accidents.

Science.gov (United States)

Lyon, Richard M; Macauley, Ben; Richardson, Sarah; de Coverly, Richard; Russell, Malcolm

2015-04-01

Horse riding is a common leisure activity associated with a significant rate of injury. Helicopter emergency medical services (HEMS) may be called to equestrian accidents. Accurate HEMS tasking is important to ensure appropriate use of this valuable medical resource. We sought to review HEMS response to equestrian accidents and identify factors associated with the need for HEMS intervention or transport of the patient to a major trauma centre. Retrospective case review of all missions flown by Kent, Surrey & Sussex Air Ambulance Trust over a 1-year period (1 July 2011 to 1 July 2012). All missions were screened for accidents involving a horse. Call details, patient demographics, suspected injuries, clinical interventions and patient disposition were all analysed. In the 12-month data collection period there were 47 equestrian accidents, representing ∼3% of the total annual missions. Of the 42 cases HEMS attended, one patient was pronounced life extinct at the scene. In 15 (36%) cases the patient was airlifted to hospital. In four (10%) cases, the patient underwent prehospital anaesthesia. There were no specific predictors of HEMS intervention. Admission to a major trauma centre was associated with the rider not wearing a helmet, a fall onto their head or the horse falling onto the rider. Equestrian accidents represent a significant proportion of HEMS missions. The majority of patients injured in equestrian accidents do not require HEMS intervention, however, a small proportion have life-threatening injuries, requiring immediate critical intervention. Further research is warranted, particularly regarding HEMS dispatch, to further improve accuracy of tasking to equestrian accidents.

Some perspective decisions for the regeneration system equipment of the thermal and nuclear power plants decreasing the probability of water ingress into the turbine and rotor acceleration by return steam flow

Science.gov (United States)

Trifonov, N. N.; Svyatkin, F. A.; Sintsova, T. G.; Ukhanova, M. G.; Yesin, S. B.; Nikolayenkova, E. K.; Yurchenko, A. Yu.; Grigorieva, E. B.

2016-03-01

The regeneration system heaters are one of the sources of possible ingress of the water into the turbine. The water penetrates into the turbine either at the heaters overflow or with the return flow of steam generated when the water being in the heater boils up in the dynamic operation modes or at deenergization of the power-generating unit. The return flow of steam and water is dangerous to the turbine blades and can result in the rotor acceleration. The known protective devices used to prevent the overflow of the low-pressure and high-pressure heaters (LPH and HPH), of the horizontal and vertical heaters of heating-system water (HWH and VWH), as well as of the deaerators and low-pressure mixing heaters (LPMH) were considered. The main protective methods of the steam and water return flows supplied by the heaters in dynamic operation modes or at deenergization of the power-generating unit are described. Previous operating experience shows that the available protections do not fully prevent water ingress into the turbine and the rotor acceleration and, therefore, the development of measures to decrease the possibility of ingress of the water into the turbine is an actual problem. The measures allowing eliminating or reducing the water mass in the heaters are expounded; some of them were designed by the specialists of OAO Polzunov Scientific and Development Association on Research and Design of Power Equipment (NPO CKTI) and are efficiently introduced at heat power plants and nuclear power plants. The suggested technical solutions allow reducing the possibility of the water ingress into the turbine and rotor acceleration by return steam flow in the dynamic operation modes or in the case of power generating unit deenergization. Some of these solutions have been tested in experimental-industrial exploitation and can be used in industry.

Natural hazard impacts on transport systems: analyzing the data base of transport accidents in Russia

Science.gov (United States)

Petrova, Elena

2015-04-01

We consider a transport accident as any accident that occurs during transportation of people and goods. It comprises of accidents involving air, road, rail, water, and pipeline transport. With over 1.2 million people killed each year, road accidents are one of the world's leading causes of death; another 20-50 million people are injured each year on the world's roads while walking, cycling, or driving. Transport accidents of other types including air, rail, and water transport accidents are not as numerous as road crashes, but the relative risk of each accident is much higher because of the higher number of people killed and injured per accident. Pipeline ruptures cause large damages to the environment. That is why safety and security are of primary concern for any transport system. The transport system of the Russian Federation (RF) is one of the most extensive in the world. It includes 1,283,000 km of public roads, more than 600,000 km of airlines, more than 200,000 km of gas, oil, and product pipelines, 115,000 km of inland waterways, and 87,000 km of railways. The transport system, especially the transport infrastructure of the country is exposed to impacts of various natural hazards and weather extremes such as heavy rains, snowfalls, snowdrifts, floods, earthquakes, volcanic eruptions, landslides, snow avalanches, debris flows, rock falls, fog or icing roads, and other natural factors that additionally trigger many accidents. In June 2014, the Ministry of Transport of the RF has compiled a new version of the Transport Strategy of the RF up to 2030. Among of the key pillars of the Strategy are to increase the safety of the transport system and to reduce negative environmental impacts. Using the data base of technological accidents that was created by the author, the study investigates temporal variations and regional differences of the transport accidents' risk within the Russian federal regions and a contribution of natural factors to occurrences of different

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

Plutonium accident resistant container project

International Nuclear Information System (INIS)

Andersen, J.A.

1978-09-01

The PARC (plutonium accident resistant container) project resulted in the design, development, and certification testing of a crashworthy air-transportable plutonium package (shipping container) for certification by the USNRC (Nuclear Regulatory Commission). This PAT-1 (plutonium air transportable) package survives a very severe sequential test program of impact, crush, puncture, slash, burn, and water immersion. There is also an individual hydrostatic pressure test. The package has a payload mass capacity of 2 kg of PuO 2 and a thermal capacity of 25 watts. The design rationale for very high energy absorption (impact, crush, puncture, and slash protection) with residual high-level fire protection, resulted in a reasonably small air-transportable package, advancing the packaging state-of-art. Optimization design iterations were utilized in the areas of impact energy absorption and stress and thermal analysis. Package test results are presented in relation to radioactive materials containment acceptance criteria, shielding and criticality standards

Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

Directory of Open Access Journals (Sweden)

Rainer Moormann

2008-01-01

Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

Accidents of Electrical and Mechanical Works for Public Sector Projects in Hong Kong.

Science.gov (United States)

Wong, Francis K W; Chan, Albert P C; Wong, Andy K D; Hon, Carol K H; Choi, Tracy N Y

2018-03-10

A study on electrical and mechanical (E&M) works-related accidents for public sector projects provided the opportunity to gain a better understanding of the causes of accidents by analyzing the circumstances of all E&M works accidents. The research aims to examine accidents of E&M works which happened in public sector projects. A total of 421 E&M works-related accidents in the "Public Works Programme Construction Site Safety and Environmental Statistics" (PCSES) system were extracted for analysis. Two-step cluster analysis was conducted to classify the E&M accidents into different groups. The results identified three E&M accidents groups: (1) electricians with over 15 years of experience were prone to 'fall of person from height'; (2) electricians with zero to five years of experience were prone to 'slip, trip or fall on same level'; (3) air-conditioning workers with zero to five years of experience were prone to multiple types of accidents. Practical measures were recommended for each specific cluster group to avoid recurrence of similar accidents. The accident analysis would be vital for industry practitioners to enhance the safety performance of public sector projects. This study contributes to filling the knowledge gap of how and why E&M accidents occur and promulgating preventive measures for E&M accidents which have been under researched.

Accidents of Electrical and Mechanical Works for Public Sector Projects in Hong Kong

Directory of Open Access Journals (Sweden)

Francis K. W. Wong

2018-03-01

Full Text Available A study on electrical and mechanical (E&M works-related accidents for public sector projects provided the opportunity to gain a better understanding of the causes of accidents by analyzing the circumstances of all E&M works accidents. The research aims to examine accidents of E&M works which happened in public sector projects. A total of 421 E&M works-related accidents in the “Public Works Programme Construction Site Safety and Environmental Statistics” (PCSES system were extracted for analysis. Two-step cluster analysis was conducted to classify the E&M accidents into different groups. The results identified three E&M accidents groups: (1 electricians with over 15 years of experience were prone to ‘fall of person from height’; (2 electricians with zero to five years of experience were prone to ‘slip, trip or fall on same level’; (3 air-conditioning workers with zero to five years of experience were prone to multiple types of accidents. Practical measures were recommended for each specific cluster group to avoid recurrence of similar accidents. The accident analysis would be vital for industry practitioners to enhance the safety performance of public sector projects. This study contributes to filling the knowledge gap of how and why E&M accidents occur and promulgating preventive measures for E&M accidents which have been under researched.

Accidents of Electrical and Mechanical Works for Public Sector Projects in Hong Kong

Science.gov (United States)

Wong, Francis K. W.; Chan, Albert P. C.; Wong, Andy K. D.; Choi, Tracy N. Y.

2018-01-01

A study on electrical and mechanical (E&M) works-related accidents for public sector projects provided the opportunity to gain a better understanding of the causes of accidents by analyzing the circumstances of all E&M works accidents. The research aims to examine accidents of E&M works which happened in public sector projects. A total of 421 E&M works-related accidents in the “Public Works Programme Construction Site Safety and Environmental Statistics” (PCSES) system were extracted for analysis. Two-step cluster analysis was conducted to classify the E&M accidents into different groups. The results identified three E&M accidents groups: (1) electricians with over 15 years of experience were prone to ‘fall of person from height’; (2) electricians with zero to five years of experience were prone to ‘slip, trip or fall on same level’; (3) air-conditioning workers with zero to five years of experience were prone to multiple types of accidents. Practical measures were recommended for each specific cluster group to avoid recurrence of similar accidents. The accident analysis would be vital for industry practitioners to enhance the safety performance of public sector projects. This study contributes to filling the knowledge gap of how and why E&M accidents occur and promulgating preventive measures for E&M accidents which have been under researched. PMID:29534429

A compendium of the measurements related to the Chernobyl nuclear accident

International Nuclear Information System (INIS)

Deworm, J.P.

1987-01-01

Results of radiation measurements performed in Belgium after the Chernobyl accident are presented. Contamination of air, soil, milk, grass, fruit, vegetables and water is studied. The committed effective dose equivalents for the population are estimated. (MCB)

K Basins floor sludge retrieval system knockout pot basket fuel burn accident

International Nuclear Information System (INIS)

HUNT, J.W.

1998-01-01

The K Basins Sludge Retrieval System Preliminary Hazard Analysis Report (HNF-2676) identified and categorized a series of potential accidents associated with K Basins Sludge Retrieval System design and operation. The fuel burn accident was of concern with respect to the potential release of contamination resulting from a runaway chemical reaction of the uranium fuel in a knockout pot basket suspended in the air. The unmitigated radiological dose to an offsite receptor from this fuel burn accident is calculated to be much less than the offsite risk evaluation guidelines for anticipated events. However, because of potential radiation exposure to the facility worker, this accident is precluded with a safety significant lifting device that will prevent the monorail hoist from lifting the knockout pot basket out of the K Basin water pool

Radioactive contamination characteristics in China following Chernobyl accident

International Nuclear Information System (INIS)

Wang Zuoyuan

1987-01-01

In the aftermath of Chernobyl nuclear reactor accident, the Environmental Radiation Surveillance Network of Ministry of Public Health of China has done monitoring on environmental samples to determine the contamination levels of radioactivity. Radionuclides, such as I-131, I-132, Cs-137, Cs-134 and Te-132, were found on surface of airplanes, which flew in domestic airlines between May 1-3, that means the radionuclides from Chernobyl accident already reached high altitude atmosphere over China, but the concentration was much lower than that in Europe. During the period of May 2-15, in most stations, radionuclides were found in different environmental samples, such as air, milk, vegetables, rain water, river and lake water, and sheep thyroid. Radioactivity levels of samples were higher in north part of China than in south. The amounts of radionuclides in all samples were well below the derived air concentrations and derived intake concentrations specified in the National Basic Health Standards for Radiological Protection. Thus, the public need not to take any precautions for the purpose of radiation protection

Use of analytical aids for accident management

International Nuclear Information System (INIS)

Ward, L.W.

1991-01-01

The use of analytical aids by utility technical support teams can enhance the staff's ability to manage accidents. Since instrumentation is exposed to environments beyond design-basis conditions, instruments may provide ambiguous information or may even fail. While it is most likely that many instruments will remain operable, their ability to provide unambiguous information needed for the management of beyond-design-basis events and severe accidents is questionable. Furthermore, given these limitation in instrumentation, the need to ascertain and confirm current plant status and forecast future behavior to effectively manage accidents at nuclear facilities requires a computational capability to simulate the thermal and hydraulic behavior in the primary, secondary, and containment systems. With the need to extend the current preventive approach in accident management to include mitigative actions, analytical aids could be used to further enhance the current capabilities at nuclear facilities. This need for computational or analytical aids is supported based on a review of the candidate accident management strategies discussed in NUREG/CR-5474. Based on the review of the NUREG/CR-5474 strategies, two major analytical aids are considered necessary to support the implementation and monitoring of many of the strategies in this document. These analytical aids include (1) An analytical aid to provide reactor coolant and secondary system behavior under LOCA conditions. (2) An analytical aid to predict containment pressure and temperature response with a steam, air, and noncondensable gas mixture present

Plant air systems safety study: Portsmouth Gaseous Diffusion Plant

International Nuclear Information System (INIS)

1982-05-01

The Portsmouth Gaseous Diffusion Plant Air System facilities and operations are reviewed for potential safety problems not covered by standard industrial safety procedures. Information is presented under the following section headings: facility and process description (general); air plant equipment; air distribution system; safety systems; accident analysis; plant air system safety overview; and conclusion

Engineering and licensing progress of the HTR-Module

Energy Technology Data Exchange (ETDEWEB)

Weisbrodt, I A

1988-07-01

This report deals not only with the latest status of Siemens/Interatom's HTR-Module but also reflects the latest engineering and licensing progress of the HTR-Module against the background of the specified design requirements and of the discussions on passively safe reactors. Therefore, I intend to report also about two examples of the accident analysis - one design basis accident, i.e. the leak-before-break of the reactor pressure vessel and one beyond design accident, i. e. massive water ingress.

Engineering and licensing progress of the HTR-Module

International Nuclear Information System (INIS)

Weisbrodt, I.A.

1988-01-01

This report deals not only with the latest status of Siemens/Interatom's HTR-Module but also reflects the latest engineering and licensing progress of the HTR-Module against the background of the specified design requirements and of the discussions on passively safe reactors. Therefore, I intend to report also about two examples of the accident analysis - one design basis accident, i.e. the leak-before-break of the reactor pressure vessel and one beyond design accident, i. e. massive water ingress

Modeling and analyses of postulated UF6 release accidents in gaseous diffusion plant

International Nuclear Information System (INIS)

Kim, S.H.; Taleyarkhan, R.P.; Keith, K.D.; Schmidt, R.W.; Carter, J.C.; Dyer, R.H.

1995-10-01

Computer models have been developed to simulate the transient behavior of aerosols and vapors as a result of a postulated accident involving the release of uranium hexafluoride (UF 6 ) into the process building of a gaseous diffusion plant. UF 6 undergoes an exothermic chemical reaction with moisture (H 2 O) in the air to form hydrogen fluoride (HF) and radioactive uranyl fluoride (UO 2 F 2 ). As part of a facility-wide safety evaluation, this study evaluated source terms consisting of UO 2 F 2 as well as HF during a postulated UF 6 release accident in a process building. In the postulated accident scenario, ∼7900 kg (17,500 lb) of hot UF 6 vapor is released over a 5 min period from the process piping into the atmosphere of a large process building. UO 2 F 2 mainly remains as airborne-solid particles (aerosols), and HF is in a vapor form. Some UO 2 F 2 aerosols are removed from the air flow due to gravitational settling. The HF and the remaining UO 2 F 2 are mixed with air and exhausted through the building ventilation system. The MELCOR computer code was selected for simulating aerosols and vapor transport in the process building. MELCOR model was first used to develop a single volume representation of a process building and its results were compared with those from past lumped parameter models specifically developed for studying UF 6 release accidents. Preliminary results indicate that MELCOR predicted results (using a lumped formulation) are comparable with those from previously developed models

Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-6 test data report: thermal hydraulic results, Rev. 0

International Nuclear Information System (INIS)

Lomperski, S.; Farmer, M.T.; Kilsdonk, D.; Aeschlimann, B.

2011-01-01

The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure? (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium (∼ φ 30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength is being addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus measures the fracture strength of the crust while it is either at room temperature or above, the latter state being achieved with a heating element placed below the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the sixth water ingression test, designated SSWICS-6. This test investigated

Environmental monitoring activities in JAERI at JCO accident

International Nuclear Information System (INIS)

Yamaguchi, Takenori

2001-01-01

The Japan Atomic Energy Research Institute (JAERI) was involved in a various environmental monitoring activities, such as environmental radiation monitoring by monitoring post and monitoring car, air dust, soil, green vegetables and water sampling and measurements, neutron and gamma radiation dose rate survey around the JCO site for emergency response actions to the JCO accident on September 30, 1999. These activities were performed from Sep. 30 to Oct. 2, and were the initial and first stage activities in the emergency environmental monitoring activities. JAERI has been assigned to the public organization to support the government by Disaster Prevention Fundamental Law. These activities were performed to ensure the public safety to avoid the effluent of the accident. Through the environmental monitoring activities, I recognized that the importance of the accident information to make the best use for the initial environmental monitoring, and the monitoring information exchange is important to perform the effective monitoring activities for taking the early countermeasures such as evacuation to the public. (author)

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)

Improved emergency elevated air release for simplified PWR

International Nuclear Information System (INIS)

Naitoh, T.; Bruce, R.A.; Hirota, K.; Tajiri, Y.

1992-01-01

In developing the application of the simplified PWR in Japan, one of the most important areas is to limit post-accident site boundary whole body dose. In addressing this, the concept of Emergency Passive Air Filtration System (EPAFS) and it's feasibility is developed. The efficiency of charcoal filtering and the atmospheric diffusion effect of an elevated air release are important for dose reduction. The performance of these functions was evaluated by confirmatory testing. The test results confirmed a 99 percent efficiency of charcoal filter and an atmospheric diffusion effect higher than that of a conventional plant. The Emergency Passive Air Filtration System (EPAFS) and the atmospheric diffusion effect of elevated air release contribute to making the calculated post-accident site boundary whole body dose of simplified PWR as low as that of the conventional Japanese PWR plant. (author)

Aspects of severe accidents in transmutation systems

International Nuclear Information System (INIS)

Wider, H.U.; Karlson, J.; Jones, A.V.

2001-01-01

The different types of transmutation systems under investigation include accelerator driven (ADS) and critical systems. To switch off an accelerator in case of an accident initiation is quite important for all accidents. For a fast ADS the grace times available for doing so depend strongly on the total heat capacity and the natural circulation capability of the primary coolant. Cooling with heavy metal Pb-Bi has considerable advantages in this regard compared to gas cooling. Moreover it allows passive ex-vessel cooling with natural air or water circulation. In the remote likelihood of fuel melting, oxide fuel appears to mix with the Pb-Bi coolant. Fast critical systems that are cooled by Pb-Bi will automatically shut off if the flow or heat sink is lost. Reactivity accidents can be limited by a low total control rod worth. High temperature reactors can achieve only incomplete burning of actinides. If an accelerator is added to increase burn-up, a fast spectrum region is needed, which has a low heat capacity. (author)

Accident loads for a VVER-440/213 containment

Energy Technology Data Exchange (ETDEWEB)

Techy, Z. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Lajtha, G. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Taubner, R. [Institute for Electric Power Research (VEIKI), Budapest (Hungary)

1995-08-01

Specific features of the VVER-440/213 containment are the subdivided rectangular building and the localization system including the bubbler trays and air traps. Accident loads are calculated for a large break loss of coolant accident (LBLOCA). The maximum pressure and temperature loads are calculated with different codes during the blowdown phase of the LBLOCA. The uncertainty margins of the maximum pressure are given in this case. Sensitivity studies are performed for different leakage rates and hydraulic data of the containment. The effects of the active and passive spray systems on the depressurization are presented in this paper. The maximum pressure loads are also examined in case of degraded conditions of the localization system. (orig.).

Safety demonstration analyses on criticality for severe accident during overland transport of fresh nuclear fuel

International Nuclear Information System (INIS)

Takahashi, Satoshi; Okuno, Hiroshi; Yamada, Kenji; Watanabe, Kouji; Nomura, Yasushi; Miyoshi, Yoshinori

2005-01-01

Criticality safety analysis was performed for transport packages of uranium dioxide powder or of fresh PWR fuel involved in a severe accident during overland transportation, and as a result, sub-criticality was confirmed against impact accident conditions such as loaded by a drop from high position to a concrete or asphalt surface, and fire accident conditions such as caused by collisions with an oil tank trailer carrying lots of inflammable material in open air, or with a commonly used two-ton-truck inside an unventilated tunnel. (author)

Accident management insights after the Fukushima Daiichi NPP accident

International Nuclear Information System (INIS)

Degueldre, Didier; Viktorov, Alexandre; Tuomainen, Minna; Ducamp, Francois; Chevalier, Sophie; Guigueno, Yves; Tasset, Daniel; Heinrich, Marcus; Schneider, Matthias; Funahashi, Toshihiro; Hotta, Akitoshi; Kajimoto, Mitsuhiro; Chung, Dae-Wook; Kuriene, Laima; Kozlova, Nadezhda; Zivko, Tomi; Aleza, Santiago; Jones, John; McHale, Jack; Nieh, Ho; Pascal, Ghislain; ); Nakoski, John; Neretin, Victor; Nezuka, Takayoshi; )

2014-01-01

The Fukushima Daiichi nuclear power plant (NPP) accident, that took place on 11 March 2011, initiated a significant number of activities at the national and international levels to reassess the safety of existing NPPs, evaluate the sufficiency of technical means and administrative measures available for emergency response, and develop recommendations for increasing the robustness of NPPs to withstand extreme external events and beyond design basis accidents. The OECD Nuclear Energy Agency (NEA) is working closely with its member and partner countries to examine the causes of the accident and to identify lessons learnt with a view to the appropriate follow-up actions to be taken by the nuclear safety community. Accident management is a priority area of work for the NEA to address lessons being learnt from the accident at the Fukushima Daiichi NPP following the recommendations of Committee on Nuclear Regulatory Activities (CNRA), Committee on the Safety of Nuclear Installations (CSNI), and Committee on Radiation Protection and Public Health (CRPPH). Considering the importance of these issues, the CNRA authorised the formation of a task group on accident management (TGAM) in June 2012 to review the regulatory framework for accident management following the Fukushima Daiichi NPP accident. The task group was requested to assess the NEA member countries needs and challenges in light of the accident from a regulatory point of view. The general objectives of the TGAM review were to consider: - enhancements of on-site accident management procedures and guidelines based on lessons learnt from the Fukushima Daiichi NPP accident; - decision-making and guiding principles in emergency situations; - guidance for instrumentation, equipment and supplies for addressing long-term aspects of accident management; - guidance and implementation when taking extreme measures for accident management. The report is built on the existing bases for capabilities to respond to design basis

High-efficiency particulate air (HEPA) filter performance following service and radiation exposure

International Nuclear Information System (INIS)

Jones, L.R.

1975-01-01

Small HEPA filters were exposed to a 60 Co source with a radiation strength of 3 x 10 7 rads per hour and then exposed to steam--air mixtures at several times filter design flow, followed by extended exposure to steam and air at reduced flow. Additional filters were exposed to air flow in a reactor confinement system and then similarly tested with steam--air mixture flows. The test data and calculated effects of filter pluggage with moisture on confinement system performance following potential reactor accidents are described. Gamma radiation exposure impaired the performance of new filters only slightly and temporarily improved performance of service aged filters. Normal confinement system service significantly impaired filter performance although not sufficiently to prevent adequate performance of the SRP confinement system following an unlikely reactor accident. Calculations based on measured filter pluggage indicate that during an accident air flow could be reduced approximately 50 percent with service-degraded HEPA filters present, or approximately 10 percent with new filters damaged by the radiation exposure. (U.S.)

Contamination of the air and other environmental samples of the Ulm region by radioactive fission products after the accident of the Chernobyl reactor

International Nuclear Information System (INIS)

Krivan, V.; Egger, K.P.; Hausbeck, R.; Schmid, W.

1986-01-01

Since April 30, 1986, the radioactivity of the fission products released by the accident of the Chernobyl reactor has been measured in the air of the city of Ulm. The airborne dust samples were collected with flow calibrated samplers on cellulose acetate membrane filters and counted with a high resolution gamma ray spectrometer. Later on, the radioactivity measurements were expanded to other relevant environmental samples contaminated by radioactive atmospheric precipitates including grass, spruce needles, mosses, lichens, various kinds of food, drinking water, asphalt and concrete surface layers, municipal sewage sludge and sewage sludge ash. This paper reports the obtained results. (orig.) [de

Migration of 137Cs from air to soil and plants in the Gulsvik area, Norway after the Chernobyl reactor accident

International Nuclear Information System (INIS)

Pacyna, J.M.; Semb, A.; Christensen, G.C.

1988-02-01

The migration of 137 Cs from air to soil and vegetation after the Chernobyl accident has been studied using the concentrations measured in the Gulsvik area in Norway. The major part of the 137 Cs deposition seems to be in the soil. An uptake of 137 Cs from soil to plants through their root system is not a rapid process. Only a few percent of the deposition can be traced in plants. This seems to suggest that as far as 137 Cs is concerned, an effect of the Chernobyl releases is not an acute but a long-term phenomenon. The 137 Cs accumulation in soils is rather high, but doses not result in 137 Cs levels in plants and diet higher than acceptable in Norway

Participation of IRD/CNEN-Br in International Intercomparison of Criticality Accident Dosimetry Systems at Silene reactor, France

International Nuclear Information System (INIS)

Mauricio, Claudia Lucia P.; Fonseca, Evaldo S. da

1996-01-01

IRD has participated in an International Intercomparison of Criticality Accident Dosimetry Systems at the SILENE reactor, France on June 1993. The dosemeters were irradiated on phantoms and free in air, in bare and lead shield reactor pulses, simulating different irradiation fields that can be found in criticality accidents. Comparing with the reference measurements, the calculated mean neutron kerma found by IRD was only 2% greater for lead shield and 14% greater for bare reactor. For gamma absorbed dose, the differences were, respectively + 22% and -9% for the dosemeters free in air and -19% and -9% for dosemeters on phantoms. IRD results are closer to the real values than the mean values measured by the participants. IRD results show a good performance if its simple criticality accident system. (author)

Civilian Helicopter Search and Rescue Accidents in the United States: 1980 Through 2013.

Science.gov (United States)

Worley, Gordon H

2015-12-01

Helicopters are commonly used in search and rescue operations, and accidents have occurred during helicopter search and rescue (HSAR) missions. The purposes of this study were to investigate whether the HSAR accident rate in the United States could be determined and whether any common contributing factors or trends could be identified. Searches were conducted of the National Transportation Safety Board aviation accident database, the records of the major search and rescue and air medical organizations, and the medical and professional literature for reports of HSAR accidents. A total of 47 civilian HSAR accidents were identified during the study. Of these, 43% involved fatal injuries, compared with a 19% fatality rate for US helicopter general aviation accidents during the same time period and a 40% rate for helicopter emergency medical services. The HSAR accidents carried a significantly higher risk of fatal outcomes when compared with helicopter general aviation accidents (2-tailed Fisher's exact test, P .05). The number of HSAR missions conducted annually could not be established, so an overall accident rate could not be calculated. Although the overall number of HSAR accidents is small, the percentage of fatal outcomes from HSAR accidents is significantly higher than that from general helicopter aviation accidents and is comparable to that seen for helicopter emergency medical services operations. Further study could help to improve the safety of HSAR flights. Copyright © 2015 Wilderness Medical Society. Published by Elsevier Inc. All rights reserved.

Large population center and core melt accident considerations in siting

International Nuclear Information System (INIS)

Camarinopoulos, L.; Yadigaroglu, G.

1983-01-01

The problem of providing suitable demographic siting criteria in the presence of a very large population center in an otherwise sparsely populated region is addressed. Simple calculations were performed making maximum use of pretabulated results of studies where core melt accidents are considered. These show that taking into consideration the air flow patterns in the region can lower the expected population doses from core melt accidents more effectively than distance alone. Expected doses are compared to the annual background radiation dose. A simple siting criterion combining geographical considerations with the probability of a release reaching the large population center is proposed

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...

Environmental radioactivity measurements at BNL following the Chernobyl accident

International Nuclear Information System (INIS)

Hancock, R.; Woollam, P.B.

1986-06-01

Measurements are reported of the concentrations at Berkeley in Gloucestershire of radioactivity in the air, rainwater, tap water, soil, herbage and fresh vegetables for the period 29 April 1986 to 15 May 1986, following the Chernobyl Power Station accident. Data for up to 18 gamma emitting isotopes are reported, together with some limited actinide-in-air measurements. Deposition velocities are calculated and an assessment is presented of the sensitivity of the techniques employed. Some data are also included on the gaseous composition of the cloud and the isotope dependent dose rate from deposition. (author)

THE FUNCTION AND PURPOSE OF AIRCRAFT ACCIDENT INVESTIGATION ACCORDING TO THE INTERNATIONAL AIR LAW

Directory of Open Access Journals (Sweden)

Atip Latipulhayat

2015-10-01

Full Text Available The main objective of an aircraft accident investigation is to find out the most probable causes of such accident. This represents a technical investigation in nature. At the practical level, however, this report is often used as legal evidence before the court. This paper argues that the main purpose of an aircraft acccident investigation is technical in nature and judicial investigation is carried out when the technical investigation found elements of crime that has been alleged as the most probable cause of the accident.   Menurut Konvensi Chicago 1944, tujuan utama suatu investigasi kecelakaan pesawat udara adalah untuk menemukan penyebab terjadinya kecelakaan tersebut. Jadi, investigasi ini bersifat teknis. Namun, laporan investigasi ini seringkali dijadikan alat bukti di pengadilan khususnya apabila faktor kesalahan manusia dianggap sebagai penyebab utama terjadinya kecelakaan tersebut. Investigasi teknis berubah menjadi investigasi yuridis. Tulisan ini berpendapat bahwa hakikat investigasi kecelakaan pesawat udara adalah bersifat teknis dan investigasi yuridis hanya akan dilakukan apabila laporan investigasi teknis menunjukkan adanya elemen kriminal.

16th DOE nuclear air cleaning conference: proceedings

International Nuclear Information System (INIS)

First, M.W.

1981-02-01

Major topics discussed during the Sixteenth DOE Nuclear Air Cleaning Conference were: waste treatment, including volume reduction and storage; system and component response to stress and accident conditions; Three Mile Island accident; iodine adsorption; treatment and storage of noble gas: treatment of offgases from chemical processing; aerosol; behavior; containment venting; laboratory and in-place filter-testing methods; and particulate filtration. Volume I of the Proceedings has 49 papers from the following sessions; HEPA filter test methods; noble gas separation; air cleaning system design; containment venting; iodine adsorption; reprocessing offgas cleaning; critical review; filtration; filter testing; and aerosols. Volume II contains 44 papers from the sessions on: nuclear waste treatment; critical review; noble gas treatment; carbon-14 and tritium; air cleaning system response to stress; nuclear standards and safety; round table; open end; and air cleaning technology at Three Mile Island. Abstracts are provided for all of these papers

16th DOE nuclear air cleaning conference: proceedings

International Nuclear Information System (INIS)

First, M.W.

1981-02-01

Major topics discussed during the Sixteenth DOE Nuclear Air Cleaning Conference were: waste treatment, including volume reduction and storage; system and component response to stress and accident conditions; Three Mile Island accident; iodine adsorption; treatment and storage of noble gas; treatment of offgases from chemical processing; aerosol behavior; containment venting; laboratory and in-place filter-testing methods; and particulate filtration. Volume I of the Proceedings has 49 papers from the following sessions: HEPA filter test methods; noble gas separation; air cleaning system design; containment venting; iodine adsorption; reprocessing offgas cleaning; critical review; filtration, filter testing, and aerosols. Volume II contains 44 papers from the sessions on: nuclear waste treatment; critical review; noble gas treatment; carbon-14 and tritium; air cleaning system response to stress; nuclear standards and safety; round table; open end; and air cleaning technology at Three Mile Island. Abstracts are provided for all of these papers

Domino effect in chemical accidents: main features and accident sequences

OpenAIRE

Casal Fàbrega, Joaquim; Darbra Roman, Rosa Maria

2010-01-01

The main features of domino accidents in process/storage plants and in the transportation of hazardous materials were studied through an analysis of 225 accidents involving this effect. Data on these accidents, which occurred after 1961, were taken from several sources. Aspects analyzed included the accident scenario, the type of accident, the materials involved, the causes and consequences and the most common accident sequences. The analysis showed that the most frequent causes a...

MLAM assessment of air concentration, deposition, and dose for Chernobyl reactor accident

International Nuclear Information System (INIS)

Olsen, A.R.; Davis, W.E.; Didier, B.T.; Soldat, J.K.; Napier, B.A.; Peloquin, R.A.

1989-12-01

The purpose of this report is to provide estimates for the areas in Europe affected by the accident involving Unit 4 of the Chernobylskaya Atomic Energy Station which resulted in the release of radioactive material to the atmosphere

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)

An Examination of Aviation Accidents Associated with Turbulence, Wind Shear and Thunderstorm

Science.gov (United States)

Evans, Joni K.

2013-01-01

The focal point of the study reported here was the definition and examination of turbulence, wind shear and thunderstorm in relation to aviation accidents. NASA project management desired this information regarding distinct subgroups of atmospheric hazards, in order to better focus their research portfolio. A seven category expansion of Kaplan's turbulence categories was developed, which included wake turbulence, mountain wave turbulence, clear air turbulence, cloud turbulence, convective turbulence, thunderstorm without mention of turbulence, and low altitude wind shear, microburst or turbulence (with no mention of thunderstorms).More than 800 accidents from flights based in the United States during 1987-2008 were selected from a National Transportation Safety Board (NTSB) database. Accidents were selected for inclusion in this study if turbulence, thunderstorm, wind shear or microburst was considered either a cause or a factor in the accident report, and each accident was assigned to only one hazard category. This report summarizes the differences between the categories in terms of factors such as flight operations category, aircraft engine type, the accident's geographic location and time of year, degree of injury to aircraft occupants, aircraft damage, age and certification of the pilot and the phase of flight at the time of the accident.

Effect of alternative aging and accident simulations on polymer properties

International Nuclear Information System (INIS)

Bustard, L.D.; Chenion, J.; Carlin, F.; Alba, C.; Gaussens, G.; LeMeur, M.

1985-05-01

The influence of accident irradiation, steam, and chemical spray exposures on the behavior of twenty-three age-preconditioned polymer sample sets (twenty-one different materials) has been investigated. The test program varied the following conditions: (1) Accident simulations of irradiation and thermodynamic (steam and chemical spray) conditions were performed both sequentially and simultaneously. (2) Accident thermodynamic (steam and chemical spray) exposures were performed both with and without air present during the exposures. (3) Sequential accident irradiations were performed both at 28 0 C and 70 0 C. (4) Age preconditioning was performed both sequentially and simultaneously. (5) Sequential aging irradiations were performed both at 27 0 C and 70 0 C. (6) Sequential aging exposures were performed using two sequences: (1) thermal followed by irradiation and (2) irradiation followed by thermal. We report both general trends applicable to a majority of the tested materials as well as specific results for each polymer. Our data base consists of ultimate tensile properties at the completion of the accident exposure for three XLPO and XLPE, five EPR and EPDM, two CSPE (HYPALON), one CPE, one VAMAC, one polydiallylphtalate, and one PPS material. We also report bend test results at completion of the accident exposures for two TEFZEL materials and permanent set after compression results for three EPR, one VAMAC, one BUNA N, one SILICONE, and one VITON material

Estimates of health risks associated with uranium hexafluoride transport by air

International Nuclear Information System (INIS)

Elert, M.; Skagius, K.

1990-01-01

In Sweden air transport is considered as an alternative for the shipment of uranium hexafluoride (UF 6 ). The radiological consequences of an aeroplane accident involving UF 6 transport have been estimated and are presented as the dose from acute exposure and the dose from long-term exposure caused by ground contamination. Chemical effects of a UF 6 release are also discussed. A number of limiting scenarios have been defined, resulting in different mechanical and thermal impacts on the transport packages. The expected accident environment and the physical and chemical behaviour of the material have been used to derive a source term for the release to the air. A Gaussian dispersion model has been used to calculate the expected air concentration downwind from the accident site. The radiation dose from short-term exposure was found to be higher than the long-term exposure from uranium deposited on the ground. (author)

Reconstructing the Chernobyl Nuclear Power Plant (CNPP) accident 30 years after. A unique database of air concentration and deposition measurements over Europe.

Science.gov (United States)

Evangeliou, Nikolaos; Hamburger, Thomas; Talerko, Nikolai; Zibtsev, Sergey; Bondar, Yuri; Stohl, Andreas; Balkanski, Yves; Mousseau, Timothy A; Møller, Anders P

2016-09-01

30 years after the Chernobyl Nuclear Power Plant (CNPP) accident, its radioactive releases still remain of great interest mainly due to the long half-lives of many radionuclides emitted. Observations from the terrestrial environment, which hosts radionuclides for many years after initial deposition, are important for health and environmental assessments. Furthermore, such measurements are the basis for validation of atmospheric transport models and can be used for constraining the still not accurately known source terms. However, although the "Atlas of cesium deposition on Europe after the Chernobyl accident" (hereafter referred to as "Atlas") has been published since 1998, less than 1% of the direct observations of (137)Cs deposition has been made publicly available. The remaining ones are neither accessible nor traceable to specific data providers and a large fraction of these data might have been lost entirely. The present paper is an effort to rescue some of the data collected over the years following the CNPP accident and make them publicly available. The database includes surface air activity concentrations and deposition observations for (131)I, (134)Cs and (137)Cs measured and provided by Former Soviet Union authorities the years that followed the accident. Using the same interpolation tool as the official authorities, we have reconstructed a deposition map of (137)Cs based on about 3% of the data used to create the Atlas map. The reconstructed deposition map is very similar to the official one, but it has the advantage that it is based exclusively on documented data sources, which are all made available within this publication. In contrast to the official map, our deposition map is therefore reproducible and all underlying data can be used also for other purposes. The efficacy of the database was proved using simulated activity concentrations and deposition of (137)Cs from a Langrangian and a Euleurian transport model. Copyright © 2016. Published by

The nuclear reactor accident at Windscale - October, 1957: Environmental aspects

Energy Technology Data Exchange (ETDEWEB)

Loutit, J F; Marley, W G; Russell, R S

1960-12-01

The nature and cause of the nuclear reactor accident at Windscale in October, 1957, have been described in the summary report of the Committee of Inquiry set up by the Atomic Energy Authority. This report was published in a Command Paper Atomic Energy Office, 1957). The events leading up to the accident occurred on the 8th October, during a routine release of the energy which had become stored in the graphite moderator as a result of the normal operation of the reactor. The Committee concluded that the accident had been caused by local overheating of the uranium fuel elements, the canning of which then failed exposing the uranium and allowing it to oxidize. The temperatures in the affected channels continued to rise, leading to the combustion of the graphite. The amount of radioactivity released during the accident is not known precisely, but approximate estimates were made from the measurements of the radioactive iodine deposited on the ground in this country, and from measurements on air filters obtained both in the United Kingdom and on the continent of Europe.

Nineteenth nuclear accident dosimetry intercomparison study, August 9-13, 1982

International Nuclear Information System (INIS)

Greene, R.T.; Sims, C.C.; Swaja, R.E.

1983-11-01

The Nineteenth Nuclear Accident Dosimetry Intercomparison Study was held August 9 to 13, 1982, at the Oak Ridge National Laboratory using the Health Physics Research Reactor operated in the pulse mode to simulate nuclear criticality accidents. Participants from eight organizations measured neutron and gamma doses at air stations and on phantoms for three different shielding conditions. Measured results were compared to nuclear industry guidelines for criticality accident dosimeters which suggest accuracies of +-25% for neutron dose and +-20% for gamma dose. Seventy-two percent of the neutron dose measurements using foil activation, sodium activation, hair sulfur activation, and thermoluminescent methods met the guidelines while less than 40% of the gamma dose measurements were within +-20% of reference values. The softest neutron energy spectrum (also lowest neutron/gamma dose ratio) provided the most difficulty in measuring neutron and gamma doses. Results of this study indicate the need for continued intercomparison and testing of nuclear accident dosimetry systems and for training of evaluating personnel. 14 references, 7 figures, 16 tables

The nuclear reactor accident at Windscale - October, 1957: Environmental aspects

International Nuclear Information System (INIS)

Loutit, J.F.; Marley, W.G.; Russell, R.S.

1960-01-01

The nature and cause of the nuclear reactor accident at Windscale in October, 1957, have been described in the summary report of the Committee of Inquiry set up by the Atomic Energy Authority. This report was published in a Command Paper Atomic Energy Office, 1957). The events leading up to the accident occurred on the 8th October, during a routine release of the energy which had become stored in the graphite moderator as a result of the normal operation of the reactor. The Committee concluded that the accident had been caused by local overheating of the uranium fuel elements, the canning of which then failed exposing the uranium and allowing it to oxidize. The temperatures in the affected channels continued to rise, leading to the combustion of the graphite. The amount of radioactivity released during the accident is not known precisely, but approximate estimates were made from the measurements of the radioactive iodine deposited on the ground in this country, and from measurements on air filters obtained both in the United Kingdom and on the continent of Europe

Roles of texture in controlling oxidation, hydrogen ingress and hydride formation in Zr alloys

International Nuclear Information System (INIS)

Szpunar, Jerzy A.; Qin, Wen; Li, Hualong; Kumar, Kiran

2011-01-01

Experimental observations shows that the oxide formed on Zr alloys are strongly textured. The texture and grain-boundary characteristics of oxide are dependent on the texture of metal substrate. Computer simulation and thermodynamic modeling clarify the effect of metal substrate on structure of oxide film, and intrinsic factors affecting the microstructure. Models of diffusion process of hydrogen atoms and oxygen diffusion through oxide are presented. Both intra-granular and inter-granular hydrides were found following (0001) α-Zr //(111) δ-ZrH1.5 relationship. The through-thickness texture inhomogeneity in cladding tubes, the effects of hoop stress on the hydride orientation and the formation of interlinked hydride structure were studied. A thermodynamic model was developed to analyze the nucleation and the stress-induced reorientation of intergranular hydrides. These works provide a framework for understanding the oxidation, the hydrogen ingress and the hydride formation in Zr alloys. (author)

The Chernobyl nuclear accident: environmental radioactivity monitoring at the LENA site

International Nuclear Information System (INIS)

Genova, N.; Meloni, S.; Rosti, G.; Caramella Crespi, V.

1986-01-01

Air pumping and filtration stations nearby the LENA site, routinely active for air radioactivity monitoring, were alerted on April 28, 1986 to look for fission products coming from U.S.S.R. after the Chernobyl accident according to weather forecast. Air filters were submitted to direct gamma ray spectrometry and fission products detected. After May 1st 1986, when the maximum radionuclide concentration in air was observed, an environmental radioactivity monitoring program was started. Several matrices such as milk, soil, grass, vegetables, tap and rain water, were systematically analyzed. At the moment the program is still active but only air, milk, vegetables and meat are periodically analyzed by gamma ray spectrometry. Results, distributions and correlations are presented and discussed. (author)

Physical status of the 10 MW test module

International Nuclear Information System (INIS)

Luo Jingyu

1991-01-01

Like most graphite moderated HTR systems, the 10 MW Test-Module reactor is undermoderated. This means if the water ingress is into the pebble bed that this system will gain reactivity as moderator is added to the core. The reactivity increase caused by water ingress strongly depends on the geometry of the core, the temperature, the burnup status and especially the moderator to fuel ratio (the metal content per fuel element). For the equilibrium core 5 gram of heavy metal per fuel element is chosen in order to limit the effect of water ingress. Another important effect of water ingress is to reduce the worth of control rods which are usually located in the reflector. The distance between core and control rods in the reflector are carefully calculated. In the 10 MW Test-Module reactor the core shutdown capability of 12 control rods are selected such that any accident reactivity can be counterbalanced and only half of the rods have to be available to render the reactor down from normal operation to the cold, permanent subcritical condition. (author). 5 figs, 3 tabs

Maximal design basis accident of fusion neutron source DEMO-TIN

Energy Technology Data Exchange (ETDEWEB)

Kolbasov, B. N., E-mail: Kolbasov-BN@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

2015-12-15

When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission–fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.

Simultaneous sampling of indoor and outdoor airborne radioactivity after the Fukushima Daiichi nuclear power plant accident.

Science.gov (United States)

Ishikawa, Tetsuo; Sorimachi, Atsuyuki; Arae, Hideki; Sahoo, Sarata Kumar; Janik, Miroslaw; Hosoda, Masahiro; Tokonami, Shinji

2014-02-18

Several studies have estimated inhalation doses for the public because of the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident. Most of them were based on measurement of radioactivity in outdoor air and included the assumption that people stayed outdoors all day. Although this assumption gives a conservative estimate, it is not realistic. The "air decontamination factor" (ratio of indoor to outdoor air radionuclide concentrations) was estimated from simultaneous sampling of radioactivity in both inside and outside air of one building. The building was a workplace and located at the National Institute of Radiological Sciences (NIRS) in Chiba Prefecture, Japan. Aerosol-associated radioactive materials in air were collected onto filters, and the filters were analyzed by γ spectrometry at NIRS. The filter sampling was started on March 15, 2011 and was continued for more than 1 year. Several radionuclides, such as (131)I, (134)Cs, and (137)Cs were found by measuring the filters with a germanium detector. The air decontamination factor was around 0.64 for particulate (131)I and 0.58 for (137)Cs. These values could give implications for the ratio of indoor to outdoor radionuclide concentrations after the FDNPP accident for a similar type of building.

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....

Tracing of the radioactive cloud in Krakow after the Chernobyl nuclear accident

International Nuclear Information System (INIS)

Florkowski, T.; Grabczak, J.; Kuc, T.; Rozanski, K.

1987-01-01

Results of radioactivity measurements in Cracow after the Chernobyl accident are reported. Methods of sampling air particulates and aerosols, gases (Kr, Xe, CO 2 ) and results of total beta activity measurements by liquid scintillation counting are presented. Drinking water and dairy milk were also controlled. Several samples were analysed by gamma spectrometry for identification of radionuclides. An attempt to determine alpha emitters did not provide positive results. The influence of Chernobyl accident on the levels of natural tritium in precipitation and atmospheric radiocarbon is also shown. 13 refs., 9 figs., 2 tabs. (author)

Chernobyl - what can natural scientists or physicians say to that accident?

International Nuclear Information System (INIS)

1986-01-01

The public discussion meeting was intended to offer to the general public a platform for discussion of questions evoked by the Chernobyl reactor accident, and scientific information on what has happened there. The brief lectures therefore deal with the accident scenario as far as assessable at the time, and with the consequences to be expected for the Federal Republic of Germany, with the fallout situation in the Mainz area, and the atmospheric dispersion and transfer of air masses from Chernobyl to the FRG. The medical experts presented information on the radiation exposure of the population and the possible genetic risk. (DG) [de

Review of US accident/incident experience involving the transportation of radioactive material (RAM) 1971-1980

International Nuclear Information System (INIS)

McClure, J.D.; Emerson, E.L.

1980-01-01

This paper analyzes the transportation accidents and incidents which have occurred in the United States in the period 1971-1980 based upon the information in the Radioactive Material Transportation Accident/Incident Data Base developed by the Transportation Technology Center (TTC) at Sandia National Laboratories. The accident/incident data base incorporates the files of the Hazardous Material Incident Report (HMIR) system operated by the Material Transportation Bureau of the US Department of Transportation (DOT) with additional information obtained from the files of the US Nuclear Regulatory Commission (NRC). A principal objective of this paper is to summarize US accident/incident experience for the past ten years, providing a concise statement of radioactive material (RAM) package failure description for the transport modes of truck, rail and air

Global atmospheric dispersion modelling after the Fukushima accident

Energy Technology Data Exchange (ETDEWEB)

Suh, K.S.; Youm, M.K.; Lee, B.G.; Min, B.I. [Korea Atomic Energy Research Institute (Korea, Republic of); Raul, P. [Universidad de Sevilla (Spain)

2014-07-01

A large amount of radioactive material was released to the atmosphere due to the Fukushima nuclear accident in March 2011. The radioactive materials released into the atmosphere were mostly transported to the Pacific Ocean, but some of them were fallen on the surface due to dry and wet depositions in the northwest area from the Fukushima nuclear site. Therefore, northwest part of the nuclear site was seriously contaminated and it was designated with the restricted zone within a radius of 20 ∼ 30 km around the Fukushima nuclear site. In the early phase of the accident from 11 March to 30 March, the radioactive materials were dispersed to an area of the inland and offshore of the nuclear site by the variations of the wind. After the Fukushima accident, the radionuclides were detected through the air monitoring in the many places over the world. The radioactive plume was transported to the east part off the site by the westerly jet stream. It had detected in the North America during March 17-21, in European countries during March 23-24, and in Asia during from March 24 to April 6, 2011. The radioactive materials were overall detected across the northern hemisphere passed by 15 ∼ 20 days after the accident. Three dimensional numerical model was applied to evaluate the dispersion characteristics of the radionuclides released into the air. Simulated results were compared with measurements in many places over the world. Comparative results had good agreements in some places, but they had a little differences in some locations. The difference between the calculations and measurements are due to the meteorological data and relatively coarse resolutions in the model. Some radioactive materials were measured in Philippines, Taiwan, Hon Kong and South Korea during from March 23-28. It inferred that it was directly transported from the Fukushima by the northeastern monsoon winds. This event was well represented in the numerical model. Generally, the simulations had a good

Major Accidents (Gray Swans) Likelihood Modeling Using Accident Precursors and Approximate Reasoning.

Science.gov (United States)

Khakzad, Nima; Khan, Faisal; Amyotte, Paul

2015-07-01

Compared to the remarkable progress in risk analysis of normal accidents, the risk analysis of major accidents has not been so well-established, partly due to the complexity of such accidents and partly due to low probabilities involved. The issue of low probabilities normally arises from the scarcity of major accidents' relevant data since such accidents are few and far between. In this work, knowing that major accidents are frequently preceded by accident precursors, a novel precursor-based methodology has been developed for likelihood modeling of major accidents in critical infrastructures based on a unique combination of accident precursor data, information theory, and approximate reasoning. For this purpose, we have introduced an innovative application of information analysis to identify the most informative near accident of a major accident. The observed data of the near accident were then used to establish predictive scenarios to foresee the occurrence of the major accident. We verified the methodology using offshore blowouts in the Gulf of Mexico, and then demonstrated its application to dam breaches in the United Sates. © 2015 Society for Risk Analysis.

Domino effect in chemical accidents: main features and accident sequences.

Science.gov (United States)

Darbra, R M; Palacios, Adriana; Casal, Joaquim

2010-11-15

The main features of domino accidents in process/storage plants and in the transportation of hazardous materials were studied through an analysis of 225 accidents involving this effect. Data on these accidents, which occurred after 1961, were taken from several sources. Aspects analyzed included the accident scenario, the type of accident, the materials involved, the causes and consequences and the most common accident sequences. The analysis showed that the most frequent causes are external events (31%) and mechanical failure (29%). Storage areas (35%) and process plants (28%) are by far the most common settings for domino accidents. Eighty-nine per cent of the accidents involved flammable materials, the most frequent of which was LPG. The domino effect sequences were analyzed using relative probability event trees. The most frequent sequences were explosion→fire (27.6%), fire→explosion (27.5%) and fire→fire (17.8%). Copyright © 2010 Elsevier B.V. All rights reserved.

Information support for analysing the radiological impact of areas affected by the Chernobyl accident

Energy Technology Data Exchange (ETDEWEB)

Shershakov, V.M.; Baranov, A.Yu.; Borodin, R.V.; Golubenkov, A.V.; Godko, A.M.; Kosykh, V.S.; Korenev, A.I.; Meleshkin, M.A. [SPA `Typhoon`, Obninsk (Russian Federation)

1996-09-01

The organisation and management of data banks generated using data from monitoring the radiological situation after the Chernobyl accident is of key importance to health care and rehabilitation in the contaminated areas. Measures following the accident were based on large scale studies involving analysis and prediction of radioactive contamination. These studies included measurements of radioactivity in air, soil and water, modelling and prediction of radionuclides transport and transformation. This required the development of a computer system RECASS (RadioEcological Analysis Support System) which is currently being developed in SPA ``Typhoon``. The main tasks of RECASS are to integrate data on existing characteristics of the environment, and data on air, soil, water and biota-contamination with numerical models that account for radionuclide behaviour in all environmental media, and with radiation dose formations that are based on geographic information system (GIS) principles. The data bank of the system includes the following data bases: a data base with measurement of radioactive contamination levels in environmental media (soil, air, water); a meteorological data base; and a data base with administrative and demographic data. A set of models for radionuclide transfer in various environments incorporated in the chain permits short or long-term predictions to be made. The results of implementing RECASS to reconstruct the time and space picture of contamination in the first days after the Chernobyl accident are presented. (Author).

Informatics support for analysing the radiological impact of areas affected by the Chernobyl accident

International Nuclear Information System (INIS)

Shershakov, V.M.; Baranov, A.Yu.; Borodin, R.V.; Golubenkov, A.V.; Godko, A.M.; Kosykh, V.S.; Korenev, A.I.; Meleshkin, M.A.

1996-01-01

The organisation and management of data banks generated using data from monitoring the radiological situation after the Chernobyl accident is of key importance to health care and rehabilitation in the contaminated areas. Measures following the accident were based on large scale studies involving analysis and prediction of radioactive contamination. These studies included measurements of radioactivity in air, soil and water, modelling and prediction of radionuclides transport and transformation. This required the development of a computer system RECASS (RadioEcological Analysis Support System) which is currently being developed in SPA ''Typhoon''. The main tasks of RECASS are to integrate data on existing characteristics of the environment, and data on air, soil, water and biota-contamination with numerical models that account for radionuclide behaviour in all environmental media and with radiation dose formations that are based on geographic information system (GIS) principles. The data bank of the system includes the following databases: a data base with measurement of radioactive contamination levels in environmental media (soil, air, water); a meteorological data base; and a data base with administrative and demographic data. A set of models for radionuclide transfer in various environments incorporated in the chain permits short or long-term predictions to be made. The results of implementing RECASS to reconstruct the time and space picture of contamination in the first days after the Chernobyl accident are presented. (Author)

Information support for analysing the radiological impact of areas affected by the Chernobyl accident

International Nuclear Information System (INIS)

Shershakov, V.M.; Baranov, A.Yu.; Borodin, R.V.; Golubenkov, A.V.; Godko, A.M.; Kosykh, V.S.; Korenev, A.I.; Meleshkin, M.A.

1996-01-01

The organisation and management of data banks generated using data from monitoring the radiological situation after the Chernobyl accident is of key importance to health care and rehabilitation in the contaminated areas. Measures following the accident were based on large scale studies involving analysis and prediction of radioactive contamination. These studies included measurements of radioactivity in air, soil and water, modelling and prediction of radionuclides transport and transformation. This required the development of a computer system RECASS (RadioEcological Analysis Support System) which is currently being developed in SPA ''Typhoon''. The main tasks of RECASS are to integrate data on existing characteristics of the environment, and data on air, soil, water and biota-contamination with numerical models that account for radionuclide behaviour in all environmental media, and with radiation dose formations that are based on geographic information system (GIS) principles. The data bank of the system includes the following data bases: a data base with measurement of radioactive contamination levels in environmental media (soil, air, water); a meteorological data base; and a data base with administrative and demographic data. A set of models for radionuclide transfer in various environments incorporated in the chain permits short or long-term predictions to be made. The results of implementing RECASS to reconstruct the time and space picture of contamination in the first days after the Chernobyl accident are presented. (Author)

Alcohol-control public service announcements (PSAs) and drunk-driving fatal accidents in the United States, 1996-2010.

Science.gov (United States)

Niederdeppe, Jeff; Avery, Rosemary; Miller, Emily N

2017-06-01

Widespread concern regarding the detrimental effects of excessive alcohol consumption (especially by minors) and associated social problems (particularly drunk driving) continues to exist among policymakers, law enforcement officers, and the general public. Alcohol consumption is a leading contributor to death from injuries, which itself is one of the main causes of death for people under 21years of age in the United States. This study examines the relationship between the volume and timing of alcohol-control public service announcements (PSAs) and rates of drunk-driving fatal accidents in the U.S. We estimate ordinary least squares (OLS) regression models to predict rates of drunk-driving fatal accidents by state and month as a function of the volume of alcohol-control PSAs aired during the previous 8months. Models include controls for state anti-drunk-driving laws and regulations, state demographic characteristics, state taxes on alcohol, calendar year, and seasonality. Results indicate that higher volumes of anti-drunk driving PSAs airing in the preceding 2 to 3months are associated, albeit modest in magnitude, with reduced rates of drunk-driving fatal accidents. The regression coefficients are largest for adults (relative to underage drunk drivers) and when the PSAs air during prime time (relative to daytime or nighttime). We conclude that PSAs could play an important contributing role in reducing drunk-driving fatal accidents, although levels of exposure and potential effects likely remain modest due to reliance on donated air time. Well-funded anti-drunk driving campaigns could achieve higher levels of exposure and have a larger impact. Copyright © 2017 Elsevier Inc. All rights reserved.

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

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.

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.

Learning lessons from Natech accidents - the eNATECH accident database

Science.gov (United States)

Krausmann, Elisabeth; Girgin, Serkan

2016-04-01

When natural hazards impact industrial facilities that house or process hazardous materials, fires, explosions and toxic releases can occur. This type of accident is commonly referred to as Natech accident. In order to prevent the recurrence of accidents or to better mitigate their consequences, lessons-learned type studies using available accident data are usually carried out. Through post-accident analysis, conclusions can be drawn on the most common damage and failure modes and hazmat release paths, particularly vulnerable storage and process equipment, and the hazardous materials most commonly involved in these types of accidents. These analyses also lend themselves to identifying technical and organisational risk-reduction measures that require improvement or are missing. Industrial accident databases are commonly used for retrieving sets of Natech accident case histories for further analysis. These databases contain accident data from the open literature, government authorities or in-company sources. The quality of reported information is not uniform and exhibits different levels of detail and accuracy. This is due to the difficulty of finding qualified information sources, especially in situations where accident reporting by the industry or by authorities is not compulsory, e.g. when spill quantities are below the reporting threshold. Data collection has then to rely on voluntary record keeping often by non-experts. The level of detail is particularly non-uniform for Natech accident data depending on whether the consequences of the Natech event were major or minor, and whether comprehensive information was available for reporting. In addition to the reporting bias towards high-consequence events, industrial accident databases frequently lack information on the severity of the triggering natural hazard, as well as on failure modes that led to the hazmat release. This makes it difficult to reconstruct the dynamics of the accident and renders the development of

Environmental radioactivity measurements in greece following the Fukushima Daichi nuclear accident

International Nuclear Information System (INIS)

Potiriadis, C.; Kolovou, M.; Clouvas, A.; Xanthos, S.

2008-01-01

Since the double disaster of the 9.0 magnitude earthquake and tsunami that affected hundreds of thousands of people and seriously damaged the Fukushima Daichi power plant in Japan on 11 March 2011, traces of radioactive emissions from Fukushima have spread across the entire northern hemisphere. The radioactive isotope of iodine 131 I that was generated by the nuclear accident in Fukushima arrived in Greece on 24 March 2011. Radioactive iodine is present in the air either as gas or bound to particles (aerosols). The maximum 131 I concentrations were measured between 3 and 5 April 2011. In aerosols the maximum 131 I values measured in Southern Greece (Athens) and Northern Greece (Thessaloniki) were 585±70 and 408±61 μBq m -3 , respectively 131 I concentrations in gas were about 3.5 times higher than in aerosols. Since 29 April 2011, the 131 I concentration has been below detection limits. Traces of 137 Cs and 134 Cs were also measured in the air filters with an activity ratio of 137 Cs/ 134 Cs equal to 1 and 131 I/ 137 Cs activity ratio of about 3. Since 16 May 2011, the 137 Cs concentration in air has been determined to be about the same as before the Fukushima accident. Traces of 131 I were also measured in grass and milk. The maximum measured activity of 131 I in sheep milk was about 2 Bq l -1 which is 5000 times less than that measured in Greece immediately after the Chernobyl accident. The measured activity concentrations of artificial radionuclides in Greece due to the Fukushima release, have been very low, with no impact on human health. (authors)

Interactions between multiple organizations responding to a reactor accident in Switzerland

International Nuclear Information System (INIS)

Baggenstos, M.; Isaak, H.P.

1987-01-01

The poster deals with the problems that emerge from the distribution of responsibilities in the case of a nuclear accident. It is demonstrated how clear and simple procedures and good co-ordination between the onsite and offsite emergency organizations are necessary for the smooth handling of such a situation. A good communications network is particularly essential. The special problem of co-ordination between different organizations performing measurements of the contaminated air, ground and food is also discussed. From the experience gained during the Chernobyl accident, it has been established that only those organizations, which are correctly trained and have a perspective over the complete accident situation, can perform valuable work. At the beginning of the accident for example, there was no organization responsible for taking samples covering the whole of Switzerland. A first quick assessment of the radiological situation could therefore only be made on the basis of measurements performed in the direct vicinity of the laboratories or from single samples which had to be transported over large distances

Atmospheric radionuclides from Fukushima Dai-ichi nuclear accident detected in Lanzhou, China

Institute of Scientific and Technical Information of China (English)

邬家龙王赟; 孙卫; 罗伟立; 王延俊; 张飙

2015-01-01

After the Fukushima Dai-ichi nuclear power plant accident on March 11, 2011, the radioactivity released from the accident was transported around the globe by atmospheric processes. The radioactivity monitoring program on atmospheric particulate in Lanzhou, China was activated by GSCDC to detect the input radionu-clides through atmospheric transport. Several artificial radionuclides were detected and measured in aerosol samples from March 26 to May 2, 2011. The peaked activity concentrations (in mBq/m3) were: 1.194 (131I), 0.231 (137Cs), 0.173 (134Cs) and 0.008 (136Cs), detected on April 6, 2011. The average activity ratio of 131I/137Cs and 134Cs/137Cs in air were 13.5 and 0.78. The significant increase of 137Cs activity concentration, one order of magnitude higher than pre-Fukushima accident levels, in ground level aerosol was observed in 2013, as its re-suspension from soil. The back-trajectory analysis simulated by NOAA-ARL HYSPLIT shows a direct transfer of the air masses released from Fukushima to Lanzhou across the Pacific Ocean, North America and Europe at the height close to 9000 m AGL. The value of effective dose for inhalation is close to one millionth of the annual limit for the general public.

Impact of Air Entraining Method on the Resistance of Concrete to Internal Cracking

Science.gov (United States)

Wawrzeńczyk, Jerzy; Molendowska, Agnieszka

2017-10-01

This paper presents the test results of air entrained concrete mixtures made at a constant W/C ratio of 0.44. Three different air entraining agents were used: polymer microspheres, glass microspheres and a conventional air entraining admixture. The aim of this study was to compare the effectiveness of the air entraining methods. Concrete mixture tests were performed for consistency (slump test), density and, in the case of AEA series, air content by pressure method. Hardened concrete tests were performed for compressive strength, water absorption, resistance to chloride ingress, and freeze-thaw durability - resistance to internal cracking tests were conducted in accordance with PN-88/B-06250 on cube specimens and with the modified ASTM C666 A test method on beam specimens; porosity characteristics (A, A300, \\bar L) were determined to PN-EN 480-11:1998. No significant mass and length changes were recorded for the concrete air entrained with the conventional methods or with polymer microspheres. The results indicate that polymer microspheres are a very good alternative to traditional air entraining methods for concrete, providing effective air entrainment and protection from freezing and thawing. The glass microsphere-based concretes showed insufficient freeze-thaw resistance. The test results indicate that both the conventional methods (AEA) and the air entrainment by polymer microspheres are effective air entraining methods. It has to be noted that in the case of the use of polymer microspheres, a comparable value of \\bar L and a very good freeze-thaw resistance can be achieved at a noticeably lower air and micropore contents and at lower strength loss.

Radioecological and dosimetric consequences of the Chernobyl accident in France; Consequences radioecologiques et dosimetriques de l'accident de Tchernobyl en France

Energy Technology Data Exchange (ETDEWEB)

Renaud, Ph; Beaugelin, K; Maubert, H; Ledenvic, Ph [Inst. de Protection et de Surete Nucleaire, CEA Centre d' Etudes de Fontenay-aux-Roses, 92 (France)

1997-11-01

This study has as objective a survey of the radioecological and dosimetric consequences of the Chernobyl accident in France, as well as a prognosis for the years to come. It was requested by the Direction of Nuclear Installation Safety (DSIN) in relation to different organisms which effected measurements after this accident. It is based on the use of combined results of measurements and modelling by means of the code ASTRAL developed at IPSN. Various measurements obtained from five authorities and institutions, were made available, such as: activity of air and water, soil, processed food, agricultural and natural products. However, to achieve the survey still a modelling is needed. ASTRAL is a code for evaluating the ecological consequences of an accident. It allows establishing the correspondence between the soil Remnant Surface Activities (RSA, in Bq.m{sup -2}), the activity concentration of the agricultural production and the individual and collective doses resulting from external and internal exposures (due to inhalation and ingestion of contaminated nurture). The results of principal synthesis documents on the Chernobyl accident and its consequences were also used. The report is structured in nine sections, as follows: 1.Introduction; 2.Objective and methodology; 3.Characterization of radioactive depositions; 4;Remnant surface activities; 5.Contamination of agricultural products and foods; 6.Contamination of natural, semi-natural products and of drinking water; 7.Dosimetric evaluations; 8.Proposals for the environmental surveillance; 9.Conclusion. Finally, after ten years, one concludes that at presentthe dosimetric consequences of the Chernobyl accident in France were rather limited. For the period 1986-2046 the average individual effective dose estimated for the most struck zone is lower than 1500 {mu}Sv, which represents almost 1% of the average natural exposure for the same period. At present, the cesium 137 levels are at often inferior to those recorded

Multidisciplinary response for a radiological accident developed in Argentina

International Nuclear Information System (INIS)

Di Giorgio, Marina; Tadic, Matilde; Vazquez, Marina; Cateriano, Miguel; Jordan, Osvaldo; Rojas, Carlos A.; Portas, M.; Coppola, A.; De Lellis, M. del C.; Dovasio, F.

2010-01-01

In January 2009 during an oil well logging task in Rincon de los Sauces, Province of Neuquen, Argentina, and, as a result of the temporary loss of control of a 137 Cs sealed source of 92,5 GBq (2,5 Ci), two workers were allegedly exposed to ionizing radiation. The Health Emergency Coordination System in Neuquen activated the Radiological Emergency Intervention System of the Nuclear Regulatory Authority (NRA). An initial assessment of the accident scenario and the involved persons was performed. Two workers were were transfer to Buenos Aires for a biodosimetry evaluation. In Buenos Aires, the Burn Hospital performed the first diagnostic and therapeutic approach of the radioinduced skin injuries. The NRA estimated the biological absorbed doses and the studies of telethermography. The Italian Hospital was in charge of the studies using ultrasonography and Eco- Doppler. As a result of the multidisciplinary evaluation performed, one of the workers was diagnosed with superficial radioinduced dermatitis. The mentioned intervention highlighted the importance of multidisciplinary and multiparametric work for the evaluation of a radiological accident, as well as for the diagnosis and early treatment of the exposed persons. (authors) [es

Brazilian emergency planning for radiological accidents

International Nuclear Information System (INIS)

Mendonca, A.H.

1986-01-01

Brazilian emergency planning for radiological accidents is organized to respond promptly to any emergency at nuclear power plants or other installations utilizing nuclear fuel. It consists of several committees: a general coordination committee with representatives from several federal departments, with final decision with the Brazilian Nuclear Energy Commission (CNEN), and the Federal Environmental Protection Agency (SEMA). Some committees conduct support activities. For example, the Operational Coordination Committee supervises the tasks undertaken by the Army, Navy, and Air Force in response to the needs and decisions of the general coordination committee

Control room habitability during severe accidents

International Nuclear Information System (INIS)

Siu, R.P.

1989-01-01

The requirements for protection of control room personnel against radiation hazards are specified in 10CFR50, Appendix A, GDC 19. The conventional approach involves a mechanistic evaluation of the radiation doses to control room personnel during design-basis accidents. In this study, an assessment of control room habitability during severe accidents is conducted. The potential levels of radiation hazards to control room personnel are evaluated in terms of both magnitude and probability of occurrence. The expected values for the probabilities of exceeding GDC-19 limits and the cumulative probability distributions of control room doses are determined. In this study, a pressurized water reactor with a large dry containment has been selected for analysis. The types of control rooms evaluated in this study include designs with: (a) filtered local intakes only, (b) filtered recirculation only, (c) filtered local intakes and recirculation, and (d) filtered dual remote intakes and recirculation. From the observations, it is concluded that, except for control room D, all other control room designs may require improvements in order to provide adequate radiation protection during severe accidents, particularly in terms of reducing whole-body gamma doses and skin doses. Potential design improvements include reduction of intake flows for concepts relying on pressurization, reduction in overall leakages, and control room pressurization through the use of bottled air supply

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

Severe accidents at nuclear power plants. Their risk assessment and accident management

International Nuclear Information System (INIS)

Abe, Kiyoharu.

1995-05-01

This document is to explain the severe accident issues. Severe Accidents are defined as accidents which are far beyond the design basis and result in severe damage of the core. Accidents at Three Mild Island in USA and at Chernobyl in former Soviet Union are examples of severe accidents. The causes and progressions of the accidents as well as the actions taken are described. Probabilistic Safety Assessment (PSA) is a method to estimate the risk of severe accidents at nuclear reactors. The methodology for PSA is briefly described and current status on its application to safety related issues is introduced. The acceptability of the risks which inherently accompany every technology is then discussed. Finally, provision of accident management in Japan is introduced, including the description of accident management measures proposed for BWRs and PWRs. (author)

Radioactive pollution in Athens, Greece due to the Fukushima nuclear accident

International Nuclear Information System (INIS)

Kritidis, P.; Florou, H.; Eleftheriadis, K.; Evangeliou, N.; Gini, M.; Sotiropoulou, M.; Diapouli, E.; Vratolis, S.

2012-01-01

As a result of the nuclear accident in Fukushima Dai-ichi power plant, which started on March 11, 2011, radioactive pollutants were transferred by air masses to various regions of the Northern hemisphere, including Europe. Very low concentrations of 131 I, 137 Cs and 134 Cs in airborne particulate matter were measured in Athens, Greece during the period of March 24 to April 28, 2011. The maximum air concentration of 131 I was measured on April 6, 2011 and equaled 490 ± 35 μBq m −3 . The maximum values of the two cesium isotopes were measured on the same day and equaled 180 ± 40 μBq m −3 for 137 Cs and 160 ± 30 μBq m −3 for 134 Cs. The average activity ratio of 131 I/ 137 Cs in air was 3.0 ± 0.5, while the corresponding ratio of 137 Cs/ 134 Cs equaled 1.1 ± 0.3. No artificial radionuclides could be detected in air after April 28, 2011. Traces of 131 I as a result of radioactive deposition were measured in grass, soil, sheep milk and meat. The total deposition of 131 I (dry + wet) was 34 ± 4 Bq m −2 , and of 137 Cs was less than 10 Bq m −2 . The maximum concentration of 131 I in grass was 2.1 ± 0.4 Bg kg −1 , while 134 Cs was not detected. The maximum concentrations of 131 I and 137 Cs in sheep milk were 1.7 ± 0.16 Bq kg −1 and 0.6 ± 0.12 Bq kg −1 respectively. Concentrations of 131 I up to 1.3 ± 0.2 Bq kg −1 were measured in sheep meat. Traces of 131 I were found in a number of soil samples. The radiological impact of the Fukushima nuclear accident in Athens region was practically negligible, especially as compared to that of the Chernobyl accident and also to that of natural radioactivity.

Hydrogen combustion in a flat semi-confined layer with respect to the Fukushima Daiichi accident

International Nuclear Information System (INIS)

Kuznetsov, M.; Yanez, J.; Grune, J.; Friedrich, A.; Jordan, T.

2012-01-01

The hydrogen accumulation at the top of containment or reactor building may occur due to an interaction of molten corium and water followed by a severe accident of a nuclear reactor (TMI, Chernobyl, Fukushima Daiichi). The hydrogen, released from the reactor, accumulates usually as a stratified semi-confined layer of hydrogen-air mixture. A series of large scale experiments on hydrogen combustion and explosion in a semi-confined layer of uniform and non-uniform hydrogen-air mixtures in presence of obstructions or without them was performed at the Karlsruhe Inst. of Technology (KIT). Different flame propagation regimes from slow subsonic to relative fast sonic flames and then to the detonations were experimentally investigated in different geometries and then simulated with COMSD code with respect to evaluate amount of burnt hydrogen taken place during the Fukushima Daiichi Accident (FDA). The experiments were performed in a horizontal semi-confined layer with dimensions of 9x3x0.6 m with/without obstacles opened from below. The hydrogen concentration in the mixtures with air was varied in the range of 0-34 vol. % without or with a gradient of 0-60 vol. %H 2 /m. Effects of hydrogen concentration gradient, thickness of the layer, geometry of the obstructions, average and maximum hydrogen concentration on flame propagation regimes were investigated with respect to evaluate the maximum pressure loads of internal structures. Blast wave strength and dynamics of propagation after explosion of the layer of hydrogen-air mixture was numerically simulated to reproduce the hydrogen explosion process during the Fukushima Daiichi Accident. (authors)

[Measures against Radiation Exposure Due to Large-Scale Nuclear Accident in Distant Place--Radioactive Materials in Nagasaki from Fukushima Daiichi Nuclear Power Plant].

Science.gov (United States)

Yuan, Jun; Sera, Koichiro; Takatsuji, Toshihiro

2015-01-01

To investigate human health effects of radiation exposure due to possible future nuclear accidents in distant places and other various findings of analysis of the radioactive materials contaminating the atmosphere of Nagasaki due to the Fukushima Daiichi Nuclear Power Plant accident. The concentrations of radioactive materials in aerosols in the atmosphere of Nagasaki were measured using a germanium semiconductor detector from March 2011 to March 2013. Internal exposure dose was calculated in accordance with ICRP Publ. 72. Air trajectories were analyzed using NOAA and METEX web-based systems. (134)Cs and (137)Cs were repeatedly detected. The air trajectory analysis showed that (134)Cs and (137)Cs flew directly from the Fukushima Daiichi Nuclear Power Plant from March to April 2011. However, the direct air trajectories were rarely detected after this period even when (134)Cs and (137)Cs were detected after this period. The activity ratios ((134)Cs/(137)Cs) of almost all the samples converted to those in March 2011 were about unity. This strongly suggests that the (134)Cs and (137)Cs detected mainly originated from the Fukushima Daiichi Nuclear Power Plant accident in March 2011. Although the (134)Cs and (137)Cs concentrations per air volume were very low and the human health effects of internal exposure via inhalation is expected to be negligible, the specific activities (concentrations per aerosol mass) were relatively high. It was found that possible future nuclear accidents may cause severe radioactive contaminations, which may require radiation exposure control of farm goods to more than 1000 km from places of nuclear accidents.

Radiological Consequence Analyses Following a Hypothetical Severe Accident in Japan

Energy Technology Data Exchange (ETDEWEB)

Kim, Juyub; Kim, Juyoul [FNC Technology Co., Yongin (Korea, Republic of)

2016-10-15

In order to reflect the lessons learned from the Fukushima Daiichi nuclear power plant accident, a simulator which is named NANAS (Northeast Asia Nuclear Accident Simulator) for overseas nuclear accident has been developed. It is composed of three modules: source-term estimation, atmospheric dispersion prediction and dose assessment. For the source-term estimation module, the representative reactor types were selected as CPR1000, BWR5 and BWR6 for China, Japan and Taiwan, respectively. Considering the design characteristics of each reactor type, the source-term estimation module simulates the transient of design basis accident and severe accident. The atmospheric dispersion prediction module analyzes the transport and dispersion of radioactive materials and prints out the air and ground concentration. Using the concentration result, the dose assessment module calculates effective dose and thyroid dose in the Korean Peninsula region. In this study, a hypothetical severe accident in Japan was simulated to demonstrate the function of NANAS. As a result, the radiological consequence to Korea was estimated from the accident. PC-based nuclear accident simulator, NANAS, has been developed. NANAS contains three modules: source-term estimation, atmospheric dispersion prediction and dose assessment. The source-term estimation module simulates a nuclear accident for the representative reactor types in China, Japan and Taiwan. Since the maximum calculation speed is 16 times than real time, it is possible to estimate the source-term release swiftly in case of the emergency. The atmospheric dispersion prediction module analyzes the transport and dispersion of radioactive materials in wide range including the Northeast Asia. Final results of the dose assessment module are a map projection and time chart of effective dose and thyroid dose. A hypothetical accident in Japan was simulated by NANAS. The radioactive materials were released during the first 24 hours and the source

Visualization of mole fraction distribution of slow jet forming stably stratified field

International Nuclear Information System (INIS)

Fumizawa, Motoo; Hishida, Makoto

1990-01-01

An experimental study has been performed to investigate the behavior of flow and mass transfer in gaseous slow jet in which buoyancy force opposed the flow forming stably stratified field. The study has been performed to understand the basic features of air ingress phenomena at pipe rupture accident of the high temperature gas-cooled reactor. A displacement fringe technique was adopted in Mach-Zehnder interferometer to visualize the mole fraction distribution. As the result, the followings were obtained: (1) The stably stratified fields were formed in the vicinity of the outlet of the slow jet. The penetration distance of the stably stratified fields increased with Froude number. (2) Mass fraction distributions in the stably stratified fields were well correlated with the present model using the ramp mole velocity profile. (author)

Deposition of RuO 4 on various surfaces in a nuclear reactor containment

Science.gov (United States)

Holm, Joachim; Glänneskog, Henrik; Ekberg, Christian

2009-07-01

During a severe nuclear reactor accident with air ingress, ruthenium can be released from the nuclear fuel in the form of ruthenium tetroxide. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. The aim of this work was to investigate the deposition of gaseous ruthenium tetroxide on aluminium, copper and zinc, which all appear in relatively large amounts in reactor containment. The experiments show that ruthenium tetroxide is deposited on all the metal surfaces, especially on the copper and zinc surfaces. A large deposition of ruthenium tetroxide also appeared on the relatively inert glass surfaces in the experimental set-ups. The analyses of the different surfaces, with several analytical methods, showed that the form of deposited ruthenium was mainly ruthenium dioxide.

Deposition of RuO{sub 4} on various surfaces in a nuclear reactor containment

Energy Technology Data Exchange (ETDEWEB)

Holm, Joachim, E-mail: joachim.holm@chalmers.s [Department of Nuclear Chemistry, Chalmers University of Technology, Se-412 96 Gothenburg (Sweden); Glaenneskog, Henrik [Ringhals AB, SE-430 22, Vaeroebacka (Sweden); Ekberg, Christian [Department of Nuclear Chemistry, Chalmers University of Technology, Se-412 96 Gothenburg (Sweden)

2009-07-01

During a severe nuclear reactor accident with air ingress, ruthenium can be released from the nuclear fuel in the form of ruthenium tetroxide. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. The aim of this work was to investigate the deposition of gaseous ruthenium tetroxide on aluminium, copper and zinc, which all appear in relatively large amounts in reactor containment. The experiments show that ruthenium tetroxide is deposited on all the metal surfaces, especially on the copper and zinc surfaces. A large deposition of ruthenium tetroxide also appeared on the relatively inert glass surfaces in the experimental set-ups. The analyses of the different surfaces, with several analytical methods, showed that the form of deposited ruthenium was mainly ruthenium dioxide.

The transport of civil plutonium by air

International Nuclear Information System (INIS)

1988-01-01

The paper presents the review entitled ''The Transport of Civil Plutonium by Air'' reported by the Advisory Committee on the Safe Transport of Radioactive Materials (ACTRAM) 1988. The contents contain chapters on the following topics:- the reasons for air transport, the various regulations, packagings for plutonium transport, testing of the packagings, accidents, the consequences of a release, and emergency arrangements. (U.K.)

Severe accident phenomena

International Nuclear Information System (INIS)

Jokiniemi, J.; Kilpi, K.; Lindholm, I.; Maekynen, J.; Pekkarinen, E.; Sairanen, R.; Silde, A.

1995-02-01

Severe accidents are nuclear reactor accidents in which the reactor core is substantially damaged. The report describes severe reactor accident phenomena and their significance for the safety of nuclear power plants. A comprehensive set of phenomena ranging from accident initiation to containment behaviour and containment integrity questions are covered. The report is based on expertise gained in the severe accident assessment projects conducted at the Technical Research Centre of Finland (VTT). (49 refs., 32 figs., 12 tabs.)

The principles of radioiodine dosimetry following a nuclear accident

International Nuclear Information System (INIS)

Zvonova, I.A.

1996-01-01

Based upon the experience of radioiodine dosimetry after the Chernobyl accident main principals of radioiodine measurements and dosimetry in thyroid glands of population in case of a radiation accident are discussed in the report. For the correct dose estimation following the radioiodine measurement in the thyroid one should know the ''history'' of radionuclide intake into the body of a contaminated person. So a measurement of radioiodine thyroid content should be accompanied by asking questions of investigated persons about, their life style and feeding after a nuclear incident. These data coincidently with data of radionuclides dynamic in the air and food (especially in milk products) are used for the development of radioiodine intake model and then for thyroid dose estimation. The influence of stable iodine prophylaxis and other countermeasures on values are discussed in dependence on the time of its using. Some methods of thyroid dose reconstruction used after the Chernobyl accident in Russia for a situation of thyroid radioiodine measurements lacking in a contaminated settlement are presented in the report. (author). 16 refs, 5 figs, 3 tabs

The principles of radioiodine dosimetry following a nuclear accident

Energy Technology Data Exchange (ETDEWEB)

Zvonova, I A [Institute of Radiation Hygiene, St. Petersburg (Russian Federation)

1996-08-01

Based upon the experience of radioiodine dosimetry after the Chernobyl accident main principals of radioiodine measurements and dosimetry in thyroid glands of population in case of a radiation accident are discussed in the report. For the correct dose estimation following the radioiodine measurement in the thyroid one should know the ``history`` of radionuclide intake into the body of a contaminated person. So a measurement of radioiodine thyroid content should be accompanied by asking questions of investigated persons about, their life style and feeding after a nuclear incident. These data coincidently with data of radionuclides dynamic in the air and food (especially in milk products) are used for the development of radioiodine intake model and then for thyroid dose estimation. The influence of stable iodine prophylaxis and other countermeasures on values are discussed in dependence on the time of its using. Some methods of thyroid dose reconstruction used after the Chernobyl accident in Russia for a situation of thyroid radioiodine measurements lacking in a contaminated settlement are presented in the report. (author). 16 refs, 5 figs, 3 tabs.

Coal Mine Accidents in Bangladesh: Its Causes and Remedial Measures

Directory of Open Access Journals (Sweden)

MD. MINHAJ UDDIN MONIR

2012-12-01

Full Text Available Barapukuria coal mine (BCM is located at Dinajpur district, NW Bangladesh. Total area is about 6.68 km2 and coal was encountered at a depth ranging between 118 and 509 m. Six major coal seams (seam I to seam VI were identified at BCM, of which, thickest one is the seam-VI (~36 m. The estimated coal reserves is about 390 million tons and of this, nearly 64 million tons of coal is extractable. The annual production of coal is about 1 million ton. In this study several parameters were used to explore minimization of accidents and improvement of coal production in BCM. Considering the geological, hydrological and other technical parameters, longwall mining method is applied for extraction of coal. Longwall retreating mining method is also applied for extraction of coal from single face with the operational area protected by self-advancing hydraulic powered roof support (HPRS system. These supports are moved forward causing the roof behind them to form an extensive abandoned area named goaf, which may lead to collapse with air blast. Coal in BCM is extracted from seam VI, using a multislice Longwall top coal caving method. Average ventilated air composition is ~20.94% oxygen, ~79.00% nitrogen, ~0.04% carbon monoxide, and minor water steam with huge dust. Production of coal was completed from 12 Longwall faces of 1st slice from seam VI. During production period several accidents and troubles within BCM have occurred. Poisonous gases such as carbon monoxide were detected during/after development of 1110 Longwall face, which increases up to 6000 ppm, therefore, sealed-off the face with mining equipments. During the development of belt gate roadway of 1101 Longwall face, maximum temperature and humidity increases up to 37 oC and 100%, respectively. Miners of BCM experienced some difficulties to work in that adverse environment. Air return roadway temperature was always 40 to 41oC in 1101 Longwall face during production period, faces temperature 46o

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.)

Professional experience and traffic accidents/near-miss accidents among truck drivers.

Science.gov (United States)

Girotto, Edmarlon; Andrade, Selma Maffei de; González, Alberto Durán; Mesas, Arthur Eumann

2016-10-01

To investigate the relationship between the time working as a truck driver and the report of involvement in traffic accidents or near-miss accidents. A cross-sectional study was performed with truck drivers transporting products from the Brazilian grain harvest to the Port of Paranaguá, Paraná, Brazil. The drivers were interviewed regarding sociodemographic characteristics, working conditions, behavior in traffic and involvement in accidents or near-miss accidents in the previous 12 months. Subsequently, the participants answered a self-applied questionnaire on substance use. The time of professional experience as drivers was categorized in tertiles. Statistical analyses were performed through the construction of models adjusted by multinomial regression to assess the relationship between the length of experience as a truck driver and the involvement in accidents or near-miss accidents. This study included 665 male drivers with an average age of 42.2 (±11.1) years. Among them, 7.2% and 41.7% of the drivers reported involvement in accidents and near-miss accidents, respectively. In fully adjusted analysis, the 3rd tertile of professional experience (>22years) was shown to be inversely associated with involvement in accidents (odds ratio [OR] 0.29; 95% confidence interval [CI] 0.16-0.52) and near-miss accidents (OR 0.17; 95% CI 0.05-0.53). The 2nd tertile of professional experience (11-22 years) was inversely associated with involvement in accidents (OR 0.63; 95% CI 0.40-0.98). An evident relationship was observed between longer professional experience and a reduction in reporting involvement in accidents and near-miss accidents, regardless of age, substance use, working conditions and behavior in traffic. Copyright © 2016 Elsevier Ltd. All rights reserved.

Safety study for HTR concepts under German site conditions. Main volume for phase I B

International Nuclear Information System (INIS)

1981-12-01

Within the spectrum of accidents, core heat-up, caused by failure of the power supplies to the reactor or by interruption of the cooling system, is of major significance. This accident causes no serious effects to the environment if the reactor containment building remains intact; however, in certain infrequent events the containment building could fail due to overpressure after about one week at the earliest and a significant environmental hazard would ensue. The other accidents - depressurisation, water and air ingress - are by comparison of distinctly lower importance to the overall risk. A common feature of all the significant accidents is that within the first five hours the maximum release of radioactivity to the reactor containment building is the inventory of the primary circuit, while the fission product inventory of the core is securely retained. The accidents investigated are characterised in the study by the probability of occurrence, by the types and quantities of the fission products released and by the period of release. It was stipulated that in this phase of the study the health risk to the neighbouring population should not be determined. However, the estimated quantities of released nuclides and their associated time of release lead to the conclusion that the worst possible damage to the environment is very limited. This arises from the inherent safety characteristics of the HTR and the comparatively long period available for counter measures. (orig./HP) [de

Airborne radioactivity in Finland after the Chernobyl accident in 1986

International Nuclear Information System (INIS)

Sinkko, K.; Aaltonen, H.; Mustonen, R.; Taipale, T.K.; Juutilainen, J.

1987-06-01

In the air surveillance programme of the Finnish Centre for Radiation and Nuclear Safety concentrations of artificial radionuclides are monitored in the air close to the ground. Airborne dust is collected continuously on a class fibre filter by a high-volume air sampler at Nurmijaervi, 40 km north of Helsinki, and the concentrations of radionuclides are evaluated. Extensive studies on radionuclide composition in air and spatial distribution were performed in Finland after the Chernobyl accident. The fallout situation was followed by temporary air sampling in Helsinki and Rovaniemi, with short sampling periods and also with air dust samples from the upper atmosphere. In Nurmijaervi, air samples were also taken on an activated carbon bed. All samples were measured by gammaspectrometry, but some radiochemical analyses were also performed. Fallout from Chernobyl arrived in Finland on Sunday, April 27. The maximum concentrations in air were measured on Monday evening, April 28, and ranged from a few microbecquerels to two hundred becquerels per cubic metre. At an altitude of about 1500 m the concentrations of radionuclides were even two decades higher. The radionuclide concentrations in air decreased rapidly being under one hundredth part of their maximum values after few days

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

Hydrogen combustion in a flat semi-confined layer with respect to the Fukushima Daiichi accident

Energy Technology Data Exchange (ETDEWEB)

Kuznetsov, Mike, E-mail: kuznetsov@kit.edu [Karlsruhe Institute of Technology, 76131 Karlsruhe (Germany); Yanez, Jorge [Karlsruhe Institute of Technology, 76131 Karlsruhe (Germany); Grune, Joachim; Friedrich, Andreas [Pro-Science GmbH, 76275 Ettlingen (Germany); Jordan, Thomas [Karlsruhe Institute of Technology, 76131 Karlsruhe (Germany)

2015-05-15

Graphical abstract: - Highlights: • Critical conditions for flame propagation regimes in a layer geometry are analyzed. • Numerical simulation of hydrogen explosion reproduces real strength of shock waves. • From 80 to 200 kg of hydrogen were exploded during Fukushima (Unit I) accident. • A sonic deflagration with TNT equivalent of 800 kg was the most probable regime. - Abstract: Hydrogen accumulations at the top of a containment or reactor building may occur due to the interaction of molten corium and water followed by a severe accident of a nuclear reactor (TMI, Chernobyl, Fukushima Daiichi). The hydrogen that is released from the reactor accumulates usually as a stratified semi-confined layer of hydrogen–air mixture. A series of large scale experiments on hydrogen combustion and explosion in a semi-confined layer of uniform and non-uniform hydrogen–air mixtures in the presence of obstructions or without them was performed at the Karlsruhe Institute of Technology (KIT). Different flame propagation regimes from slow subsonic to relatively fast sonic flames and then to detonations were experimentally investigated in different geometries and then simulated with COM3D code with respect to evaluate the amount of hydrogen that was involved in the Fukushima Daiichi Accident (FDA). The experiments were performed in a horizontal semi-confined layer with the dimensions 9 × 3 × 0.6 m with/without obstacles opened from below. The hydrogen concentration in the mixtures with air was varied in the range of 10–34 vol.% without or with a gradient of 20–60 vol.%H{sub 2}/m. Effects of hydrogen concentration gradient, layer thickness, obstruction geometry, average and maximum hydrogen concentration on the flame propagation regimes were investigated with respect to evaluate the maximum pressure loads on internal structures. Blast wave strength and dynamics of propagation after the explosion of the hydrogen–air mixture layer were numerically simulated to reproduce

Vehicle accidents related to sleep: a review.

Science.gov (United States)

Horne, J; Reyner, L

1999-05-01

Falling asleep while driving accounts for a considerable proportion of vehicle accidents under monotonous driving conditions. Many of these accidents are related to work--for example, drivers of lorries, goods vehicles, and company cars. Time of day (circadian) effects are profound, with sleepiness being particularly evident during night shift work, and driving home afterwards. Circadian factors are as important in determining driver sleepiness as is the duration of the drive, but only duration of the drive is built into legislation protecting professional drivers. Older drivers are also vulnerable to sleepiness in the mid-afternoon. Possible pathological causes of driver sleepiness are discussed, but there is little evidence that this factor contributes greatly to the accident statistics. Sleep does not occur spontaneously without warning. Drivers falling asleep are unlikely to recollect having done so, but will be aware of the precursory state of increasing sleepiness; probably reaching a state of fighting off sleep before an accident. Self awareness of sleepiness is a better method for alerting the driver than automatic sleepiness detectors in the vehicle. None of these have been proved to be reliable and most have shortcomings. Putative counter measures to sleepiness, adopted during continued driving (cold air, use of car radio) are only effective for a short time. The only safe counter measure to driver sleepiness, particularly when the driver reaches the stage of fighting sleep, is to stop driving, and--for example, take a 30 minute break encompassing a short (driving while sleepy, and driving at vulnerable times of the day.

Occupational Radiation Protection in Severe Accident Management. EG-SAM Interim Report

International Nuclear Information System (INIS)

2014-01-01

As an early response to the Fukushima NPP accident, the ISOE Bureau decided to focus on the following issues as an initial response of the joint program after having direct communications with the Japanese official participants in April 2011; - Management of high radiation area worker doses: It has been decided to make available the experience and information from the Chernobyl accident in terms of how emergency worker / responder doses were legally and practically managed, - Personal protective equipment for highly-contaminated areas: It was agreed to collect information about the types of personnel protective equipment and other equipment (e.g. air bottles, respirators, air-hoods or plastic suits, etc.), as well as high-radiation area worker dosimetry use (e.g. type, number and placement of dosimetry) for different types of emergency and high-radiation work situations. Detailed information was collected on dose criteria which are used for emergency workers/responders and their basis, dose management criteria for high dose/dose rate areas, protective equipment which is recommended for emergency workers / responders, recommended individual monitoring procedures, and any special requirement for assessment from the ISOE participating nuclear utilities and regulatory authorities and made available for Japanese utilities. With this positive response of the ISOE actors and interest in the situation in Fukushima, the Expert Group on Occupational Radiation Protection in Severe Accident Management (EG-SAM) was established by the ISOE Management Board in May 2011. The overall objective of the EG-SAM is to contribute to occupational exposure management (providing a view on management of high radiation area worker doses) within the Fukushima plant boundary with the ISOE participants and to develop a state-of-the- art ISOE report on best radiation protection management practices for proper radiation protection job coverage during severe accident initial response and recovery

Radioecological and dosimetric consequences of the Chernobyl accident in France; Consequences radioecologiques et dosimetriques de l'accident de Tchernobyl en France

Energy Technology Data Exchange (ETDEWEB)

Renaud, Ph.; Beaugelin, K.; Maubert, H.; Ledenvic, Ph. [Inst. de Protection et de Surete Nucleaire, CEA Centre d' Etudes de Fontenay-aux-Roses, 92 (France)

1997-11-01

This study has as objective a survey of the radioecological and dosimetric consequences of the Chernobyl accident in France, as well as a prognosis for the years to come. It was requested by the Direction of Nuclear Installation Safety (DSIN) in relation to different organisms which effected measurements after this accident. It is based on the use of combined results of measurements and modelling by means of the code ASTRAL developed at IPSN. Various measurements obtained from five authorities and institutions, were made available, such as: activity of air and water, soil, processed food, agricultural and natural products. However, to achieve the survey still a modelling is needed. ASTRAL is a code for evaluating the ecological consequences of an accident. It allows establishing the correspondence between the soil Remnant Surface Activities (RSA, in Bq.m{sup -2}), the activity concentration of the agricultural production and the individual and collective doses resulting from external and internal exposures (due to inhalation and ingestion of contaminated nurture). The results of principal synthesis documents on the Chernobyl accident and its consequences were also used. The report is structured in nine sections, as follows: 1.Introduction; 2.Objective and methodology; 3.Characterization of radioactive depositions; 4;Remnant surface activities; 5.Contamination of agricultural products and foods; 6.Contamination of natural, semi-natural products and of drinking water; 7.Dosimetric evaluations; 8.Proposals for the environmental surveillance; 9.Conclusion. Finally, after ten years, one concludes that at presentthe dosimetric consequences of the Chernobyl accident in France were rather limited. For the period 1986-2046 the average individual effective dose estimated for the most struck zone is lower than 1500 {mu}Sv, which represents almost 1% of the average natural exposure for the same period. At present, the cesium 137 levels are at often inferior to those recorded

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...

Dose assessment for emergency workers in early phase of Fukushima Daiichi nuclear power plant accident

Energy Technology Data Exchange (ETDEWEB)

Sadeghi, Nahid; Ahangari, Rohollah; Kasesaz, Yaser; Noori-kalkhoran, O. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research School

2017-11-15

In the case of Fukushima Daiichi nuclear power plant (FNP) accident, the radioactive material was released from reactor units 1-3 and transported to short and long distances due to the atmospheric pathways-motions. Power sources for monitoring posts were lost due to earthquake and tsunami. Based on air dose rates and other data measured by monitoring cars, the amount of radioactive material released to the atmosphere from the power station was obtained. The atmospheric dispersion and the transport model used in the RASCAL code, estimate the radionuclide concentrations downwind, both in the air and on the ground due to deposition. The calculated concentrations are then used to estimate the projected doses for workers in vicinity of the accident area in the first minutes of accident time. For dose modeling, we assumed that each worker was 15 min in vicinity of FNP in accident situation, once without and once with protective clothes or respirator. According to Tokyo Electric Power Company (TEPCO) report six workers had received doses over 250 mSv (309 to 678 mSv) apparently due to inhaling Iodine-131 fume. In this paper the calculated dose results using RASCAL code shows that, if emergency workers who work in early phase of accident had not used protective equipment, for 15 min, inhalation doses from iodine in their thyroid gland up to 12 March afternoon would have been 520 mSv. A comparison between calculation results and TEPCO report shows that dose calculated virtually is nearly equal to TEPCO measurement results.

Radionuclide concentrations in the northern part of The Netherlands after the Chernobyl reactor accident

International Nuclear Information System (INIS)

de Meijer, R.J.; Aldenkamp, F.J.; Brummelhuis, M.J.; Jansen, J.F.; Put, L.W.

1990-01-01

Concentrations of radionuclides originating from the Chernobyl reactor accident were measured as a function of time in air, rainwater, grass, cow's milk, vegetables and dust by means of high-resolution gamma-ray spectroscopy. Special attention was paid to grass and milk originating from the same meadows. Also, milk of cows temporarily kept inside after the accident was monitored until a few days after their release from the stables. Activity ratios in various types of samples and the implication of the sheltering measures for cows are discussed

Accident Locations, MDTA Accidents, Accidents on MDTA locations, Accidents on I 95, US 50, I 695, Accident on John F Kennedy Highway, Nice Bridge, Bay Bridge locations, Published in 2011, 1:1200 (1in=100ft) scale, Maryland Transportation Authority.

Data.gov (United States)

NSGIC State | GIS Inventory — Accident Locations dataset current as of 2011. MDTA Accidents, Accidents on MDTA locations, Accidents on I 95, US 50, I 695, Accident on John F Kennedy Highway, Nice...

Post-Chernobyl accident radioactivity measurements in the Comunidad Autonoma de Valencia, Spain.

Science.gov (United States)

Ortiz, J; Ballesteros, L; Serradell, V

1992-03-01

Increased atmospheric radioactivity after the accident in Chernobyl was first detected on air filters. Measurements were begun in Valencia on May 2, 1986, with the maximum activity being observed around May 3-4, 1986. As a consequence of this accident, annual campaigns of measurements on migrating birds (several species of aquatic birds and song-thrushes) were started. The data corresponding to the campaign immediately after the accident (1986/87) show a generalized contamination (approximately 50% of the measured specimens). Significant levels of 134Cs, 137Cs and 110Agm were found. It is important to note that 110Agm is only present in Aythya ferina. In the successive campaigns in 1988/89 and 1989/91 few samples were found to be contaminated and only 137Cs was identified. Strontium-90 was measured and identified in some specimens, mainly in their bones.

Biological effects of radiation accidents on humans. September 1970-February 1990 (a Bibliography from the NTIS data base). Report for September 1970-February 1990

International Nuclear Information System (INIS)

1990-03-01

This bibliography contains citations concerning the impact of radiation accidents on humans. Radiation exposure assessment for determining appropriate medical treatment is discussed. The effects of ingesting food or inhaling air irradiated by accident fallout are considered. Follow-up studies of the survivors of specific nuclear accidents are included in an attempt to evaluate long and short term health effects of accidents. (This updated bibliography contains 224 citations, 62 of which are new entries to the previous edition.)

Arabian, Asian, western: a cross-cultural comparison of aircraft accidents from human factor perspectives.

Science.gov (United States)

Al-Wardi, Yousuf

2017-09-01

Rates of aviation accident differ in different regions; and national culture has been implicated as a factor. This invites a discussion about the role of national culture in aviation accidents. This study makes a cross-cultural comparison between Oman, Taiwan and the USA. A cross-cultural comparison was acquired using data from three studies, including this study, by applying the Human Factors Analysis and Classification System (HFACS) framework. The Taiwan study presented 523 mishaps with 1762 occurrences of human error obtained from the Republic of China Air Force. The study from the USA carried out for commercial aviation had 119 accidents with 245 instances of human error. This study carried out in Oman had a total of 40 aircraft accidents with 129 incidences. Variations were found between Oman, Taiwan and the USA at the levels of organisational influence and unsafe supervision. Seven HFACS categories showed significant differences between the three countries (p culture can have an impact on aviation safety. This study revealed that national culture plays a role in aircraft accidents related to human factors that cannot be disregarded.

On-line measurements of RuO{sub 4} during a PWR severe accident

Energy Technology Data Exchange (ETDEWEB)

Reymond-Laruinaz, S.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, CEA/Saclay, 91191 Gif sur Yvette Cedex, (France); Manceron, L. [Societe Civile Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); MONARIS, UMR 8233, Universite Pierre et Marie Curie, 4 Place Jussieu, case 49, F-75252 Paris Cedex 05, (France); Boudon, V. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS-Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Ducros, G. [CEA, DEN, Departement d' Etudes des Combustibles, CEA/Cadarache, 13108 Saint-Paul-lez-Durance cedex, (France)

2015-07-01

After the Fukushima accident, it became essential to have a way to monitor in real time the evolution of a nuclear reactor during a severe accident, in order to react efficiently and minimize the industrial, ecological and health consequences of the accident. Among gaseous fission products, the tetroxide of ruthenium RuO{sub 4} is of prime importance since it has a significant radiological impact. Ruthenium is a low volatile fission product but in case of the rupture of the vessel lower head by the molten corium, the air entering into the vessel oxidizes Ru into gaseous RuO{sub 4}, which is not trapped by the Filtered Containment Venting Systems. To monitor the presence of RuO{sub 4} allows making a diagnosis of the core degradation and quantifying the release into the atmosphere. To determine the presence of RuO{sub 4}, FTIR spectrometry was selected. To study the feasibility of the monitoring, high-resolution IR measurements were realized at the French synchrotron facility SOLEIL on the infrared beam line AILES. Thereafter, theoretical calculations were done to simulate the FTIR spectrum to describe the specific IR fingerprint of the molecule for each isotope and based on its partial pressure in the air. (authors)

Air concentration of radiocaesium in Tsukuba, Japan following the release from the Tokai waste treatment plant: comparisons of observations with predictions

International Nuclear Information System (INIS)

Igarashi, Yasuhito; Aoyama, Michio; Miyao, Takashi; Hirose, Katsumi; Komura, Kazuhisa; Yamamoto, Masayoshi

1999-01-01

On March 11, 1997 a fire and explosion accident occurred at the bituminization facility of the Power Reactor and Nuclear Fuel Development, Tokai, Japan. As a result of this accident, 134,137 Cs was detected in an air filter sample collected at the Meteorological Research Institute, Tsukuba during March 10 to 12. The 134,137 Cs air concentration was about 100 and 10 μBq m -3 , respectively. This result suggests that there was little radiation exposure of the residents in the area. The average 137 Cs air concentration during this period was about two orders of magnitude higher than 'baseline' air (sub-μBq m -3 ) during February to April, 1997, measured by ultra-low background γ-spectrometry. By a simple calculation using a Gaussian plume model with the measured data, we estimated the minimum emission of the radioactivity by the PNC accident to be in the range 60 MBq to around 600 MBq. The meteorological condition during the week of the accident are also described

Environmental and agricultural impacts of the Chernobyl NPP accident on the countries of the northern hemisphere

International Nuclear Information System (INIS)

Yang Xuexian

1990-12-01

The accident at the Chernobyl Nuclear Power Plant (NPP) on April 26, 1986 resulted in large quantities of radioactive materials being released into the atmosphere. The environmental contaminations and agricultural impacts of the accident on the countries of the northern hemisphere were reviewed. Radiological consequences of the accident were briefly assessed. The data were presented on the results of radioactivity monitoring for air, ground and water, average individual effective dose commitment for each county, and levels of contamination on plant cover, milk, meat in live animals, food, aquatic, and other agricultural products etc. The transfer coefficients of radionuclides in grass-(cow)-milk were listed. Finally, problems on radioecology were discussed

Crew Configuration, Ingress/Egress Procedures, and In-Flight Caregiving Capacity in a Space Ambulance Based on the Boeing X-37B

Science.gov (United States)

Halberg, Ephriam Etan

This study proposes that a Boeing X-37B space plane, its dimensions and performance characteristics estimated from publicly available documents, diagrams, and photographs, could be internally redesigned as a medical evacuation (ambulance) vehicle for the International Space Station. As of 2017, there is currently no spacecraft designed to accommodate a contingency medical evacuation wherein a crew member aboard the ISS is injured or ailing and must be returned to Earth for immediate medical attention. The X-37B is an unmanned vehicle with a history of success in both sub-orbital testing and all four of its long-duration orbital missions to date. Research conducted at UC Davis suggests that it is possible to retain the outer mold line of the X-37B while expanding the internal payload compartment to a volume sufficient for a crew of three--pilot, crew medical officer, and injured crew member--throughout ISS un-dock and atmospheric entry, descent, and landing. In addition to crew life support systems, this re-purposed X-37B, hereafter referred to as the X-37SA (Space Ambulance), includes medical equipment for stabilization of a patient in-transit. This study suggests an optimal, ergonomic crew configuration and berthing port location, procedures for microgravity ingress and 1G egress, a minimum medical equipment list and location within the crew cabin for the medical care and monitoring equipment. Conceptual crew configuration, ingress/egress procedures, and patient/equipment access are validated via physical simulation in a full-scale mockup of the proposed X-37SA crew cabin.

Monitoring data related to the Chernobyl accident as measured in Israel during May-July 1986 and the assessment of the radiation doses to the population

International Nuclear Information System (INIS)

Schlesinger, T.; Biran, T.I.; Even, O.; Dukhan, R.; Shamai, Y.; Koch, J.; Tal, A.; Israeli, M.

1987-07-01

Environmental monitoring was undertaken on April 30 1986, to follow the effects of the Chernobyl accident on the quality of the environment in Israel. Measurements of air radioactive contamination were continuously taken to the end of July when air radioactive declined to values below 0.01 Bq/m 3 . Along air measurements, radioactive contamination of ground, rain and drinking water, grass, vegetation and food items such as vegetables, fruits, milk, meat etc. were taken as well. Assessment of the accumulated radiation doses due to the Chernobyl accident was conducted. The effective dose equivalent is estimated to be 46 μSv (4.6 mrem)

Arrival of radionuclides released by the Fukushima accident to Tenerife (Canary Islands)

International Nuclear Information System (INIS)

López-Pérez, M.; Ramos-López, R.; Perestelo, Nayra R.; Duarte-Rodriguez, X.; Bustos, J.J.; Alonso-Pérez, S.; Cuevas, E.; Hernández-Armas, J.

2013-01-01

Two weeks after the accident at the Fukushima-Daichi nuclear power plant, 131I, 137Cs and 134Cs activities were measured in two different stations located in Tenerife (Canary Islands), situated at 300 (FIMERALL) and 2400 (IZAÑA) m.a.s.l, respectively. Peak measured activity concentrations were: 1.851 mBq/m3 (131I); 0.408 mBq/m3 (137Cs) and 0.382 mBq/m3 (134Cs). The activities measured at the FIMERALL station were always higher than at IZAÑA station, suggesting that the radioactive plume arrived to the island associated with low altitude air masses. Simulations of potential dispersion of the radioactive cloud (137Cs) after the nuclear accident in reactor Fukushima I show that radioactive pollution reached remote regions such as the Canary Islands in the Eastern subtropical North Atlantic. The corresponding effective dose to the local population was 1.17 nSv, a value less than one millionth of the annual limit for the general public. Therefore, there was no risk to public health. - Highlights: ► Arrival of radionuclides to Tenerife following the accident of Fukushima. ► The atmospheric concentration of radionuclides was always higher at low altitude. ► After reaching the peak concentration a sharp decrease of radionuclides was observed. ► Air mass forward trajectory analysis confirms the potential arrival of radionuclides to Tenerife.

Use of PSA and severe accident assessment results for the accident management

International Nuclear Information System (INIS)

Jang, S. H.; Kim, H. G.; Jang, H. S.; Moon, S. K.; Park, J. U.

1993-12-01

The objectives for this study are to investigate the basic principle or methodology which is applicable to accident management, by using the results of PSA and severe accident research, and also facilitate the preparation of accidents management program in the future. This study was performed as follows: derivation of measures for core damage prevention, derivation of measures for accident mitigation, application of computerized tool to assess severe accident management

Use of PSA and severe accident assessment results for the accident management

Energy Technology Data Exchange (ETDEWEB)

Jang, S H; Kim, H G; Jang, H S; Moon, S K; Park, J U [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

1993-12-15

The objectives for this study are to investigate the basic principle or methodology which is applicable to accident management, by using the results of PSA and severe accident research, and also facilitate the preparation of accidents management program in the future. This study was performed as follows: derivation of measures for core damage prevention, derivation of measures for accident mitigation, application of computerized tool to assess severe accident management.

On possibilities to decrease influence of oxidation processes on oil yield at oil shale retorting

International Nuclear Information System (INIS)

Yefimov, V.; Loeoeper, R.; Doilov, S.; Kundel, H.

1993-01-01

At the present technical level retorting is carried out so that an increase in specific air consumption results in an increase of oxygen ingress into the semi coking shaft. At the same time a direct relationship between the degree of volatiles pyrolysis and specific air consumption was observed. This regularity enables to assume that within the reaction volume there occurs most likely the thermo oxidative pyrolysis of volatile products, not the oxidation of oil shale as it is considered traditionally. The main source of oxygen ingress ed into the semi coking shaft at processing oil shale in retorts is the process of spent shale. This process is not fully elaborated for utilization in commercial scale and can not be arranged so that the ingress of oxygen into the smacking chamber could be eliminated. In case of a slower semi coke gasification process and reduced specific air consumption for gasification the absolute amount of oxygen ingress ed into the semi coking shaft also decreases. One of the efficient methods to decrease specific air consumption is to build furnaces into the semi coking chamber to obtain additional amount of heat carrier by combusting generator gas. The maximum effect is reached when steam-and-air blow is completely replaced by recycle gas: specific air consumption is reduced whereas recycle gas is deoxygenated in the cooling zone while passing through spent shale bed which has the temperature of about 500 deg C. Another possible source of oxygen to the semi coking shaft with heat carrier is production of flue gases by combusting recycle gas in burners built in retorts. We consider the recycle gas employed upon processing oil shale in retorts hardly to be an appreciable source of the oxygen ingress into the semi coking shaft. Additional amounts of residual oxygen containing in recycle gas fed into both cooling zone and furnaces are practically totally consumed at gas combusting and passing across the bed of semi coke heated up to approximately

OECM MCCI Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-2 final data report, Rev. 0 February 12, 2003

International Nuclear Information System (INIS)

Lomperski, S.; Farmer, M.T.; Kilsdonk, D.; Aeschlimann, B.

2011-01-01

The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium (∼φ30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the second water ingression test, designated SSWICS-2. The test investigated the quench behavior of a 15 cm deep, fully

Severe accident simulation and analysis for a CAREM-like integral nuclear reactor: ex-vessel phase

International Nuclear Information System (INIS)

Caputo, M.; García, J.M.; Giménez, M.; Sánchez, S.

2013-01-01

The main phenomena and processes involved in the progression of a hypothetical nuclear severe accident in an integral type reactor like CAREM are studied, quantifying the most relevant parameters, in order to contribute to the plant design and the development of an appropriate severe accident management program. A computational plant model was developed using Melcor code, including the reactor pressure vessel and the containment. A loss of coolant accident caused by a double guillotine pipe break in the steam dome zone of the pressure vessel (1.5 inches diameter) was simulated. Along this work the analysis were focused in the containment dynamics. As a consequence of the postulated loss of coolant accident the water inventory boils off leading to the core uncovery and fuel heat-up. At high temperatures the zircaloy steam oxidation becomes relevant, with hydrogen generation as one of the reaction products. The hydrogen produced is release into the containment and the possibility of hydrogen combustion in presence of enough oxygen makes relevant the analysis of containment hydrogen distribution. It is assumed that there is not any hydrogen control system. Due to the postulated loss of coolant a big amount of steam and energy is released into the containment, with a consequent fast pressurization of the dry well which makes possible air and steam discharging into the wet well (suppression pool). At the beginning the flow discharged into the pool is mainly composed of air, a non-condensable gas that pressurizes the wet well. After most of the containment air is pushed into the atmosphere wet well the pressurization rate decreases because the flow discharge is mainly composed by steam, which condensates in the pool. Also some other containment pressure peaks were observed as a consequence of hydrogen deflagrations. (author)

Nuclear Reactor RA Safety Report, Vol. 12, Accidents during reactor operation

International Nuclear Information System (INIS)

1986-11-01

This volume includes description and analysis of typical accidents occurred during operation of RA reactor in chronological order, as follows: contamination of primary coolant circuit; leakage of heavy water from the primary coolant loop; contamination of vertical experimental channel; air contamination in the reactor building and loss of circulation of the primary coolant; failures of the vacuum pump and spent fuel packaging device; rupture of the spent fuel element cladding; dethronement's of capsule for irradiation of fuel element; rupture of the vertical experimental channel and contamination of the surroundings; swelling of a fuel element; appearance of deposits on the surface of the fuel elements cladding. The last chapter describes similar accidents occurred on nuclear reactors in the world [sr

[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.

Severe accident analysis methodology in support of accident management

International Nuclear Information System (INIS)

Boesmans, B.; Auglaire, M.; Snoeck, J.

1997-01-01

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

Comparative risk assessment of severe accidents in the energy sector

International Nuclear Information System (INIS)

Burgherr, Peter; Hirschberg, Stefan

2014-01-01

Comparative assessment of accident risks in the energy sector is a key aspect in a comprehensive evaluation of sustainability and energy security concerns. Safety performance of energy systems can have important implications on the environmental, economic and social dimensions of sustainability as well as availability, acceptability and accessibility aspects of energy security. Therefore, this study provides a broad comparison of energy technologies based on the objective expression of accident risks for complete energy chains. For fossil chains and hydropower the extensive historical experience available in PSI's Energy-related Severe Accident Database (ENSAD) is used, whereas for nuclear a simplified probabilistic safety assessment (PSA) is applied, and evaluations of new renewables are based on a combination of available data, modeling, and expert judgment. Generally, OECD and EU 27 countries perform better than non-OECD. Fatality rates are lowest for Western hydropower and nuclear as well as for new renewables. In contrast, maximum consequences can be by far highest for nuclear and hydro, intermediate for fossil, and very small for new renewables, which are less prone to severe accidents. Centralized, low-carbon technology options could generally contribute to achieve large reductions in CO 2 -emissions; however, the principal challenge for both fossil with Carbon Capture and Storage and nuclear is public acceptance. Although, external costs of severe accidents are significantly smaller than those caused by air pollution, accidents can have disastrous and long-term impacts. Overall, no technology performs best or worst in all respects, thus tradeoffs and priorities are needed to balance the conflicting objectives such as energy security, sustainability and risk aversion to support rationale decision making. - Highlights: • Accident risks are compared across a broad range of energy technologies. • Analysis of historical experience was based on the

New Solutions For Increasing Environmental Protection During Severe Accidents At Nuclear Power Plants

International Nuclear Information System (INIS)

Kulyukhin, Sergei A.; Mikheev, Nikolai B.; Falkovskii, Leo N.; Reshetov, Leo A.; Zvetkova, Marianna Ya.; Yagodkin, Ivan V.; Osipov, Viktor P.; Skvortsov, Sergei S.; Berkovich, Viktor M.; Taranov, Gennadii S.; Grigor'ev, Mikhail M.; Meshkov, Vladimir M.; Noskov, Andrei A.; Mitrofanov, Mikhail I.

2008-01-01

This paper reports new solutions for increasing environmental protection during severe accidents at NPPs. For NPPs with two protective shells and pressure release system such as WWER-1000 we suggest a new comprehensive, passive-mode environmental protection system of decontamination of the radioactive air-steam mixture from the containment and the inter-containment area, which includes the 'wet' stage (scrubbers, etc.), the 'dry' stage (sorption module), and also an ejector, which in a passive mode is capable of solving the multi-purpose task of decontamination of the air-steam mixture. For Russian WWER-440/230 NPPs we suggest three protection levels: 1) a jet-vortex condenser; 2) the spray system; 3) a sorption module. For modern designs of new generation NPPs, which do not provide for pressure release systems, we proposed a new passive filtering system together with the passive heat-removal system, which can be used during severe accidents in case all power supply units become unavailable. (authors)

New Solutions For Increasing Environmental Protection During Severe Accidents At Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Kulyukhin, Sergei A.; Mikheev, Nikolai B. [Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow (Russian Federation); Falkovskii, Leo N.; Reshetov, Leo A.; Zvetkova, Marianna Ya. [All-Russian Research Institute of Atomic Machine-Building, Moscow, Russia (Russian Federation); Yagodkin, Ivan V.; Osipov, Viktor P.; Skvortsov, Sergei S. [Institute of Physics and Power Engineering, Obninsk (Russian Federation); Berkovich, Viktor M.; Taranov, Gennadii S.; Grigor' ev, Mikhail M. [Institute ' Atomenergoproekt' , Moscow (Russian Federation); Meshkov, Vladimir M.; Noskov, Andrei A.; Mitrofanov, Mikhail I. [ROSENERGOATOM Concern, Moscow (Russian Federation)

2008-07-01

This paper reports new solutions for increasing environmental protection during severe accidents at NPPs. For NPPs with two protective shells and pressure release system such as WWER-1000 we suggest a new comprehensive, passive-mode environmental protection system of decontamination of the radioactive air-steam mixture from the containment and the inter-containment area, which includes the 'wet' stage (scrubbers, etc.), the 'dry' stage (sorption module), and also an ejector, which in a passive mode is capable of solving the multi-purpose task of decontamination of the air-steam mixture. For Russian WWER-440/230 NPPs we suggest three protection levels: 1) a jet-vortex condenser; 2) the spray system; 3) a sorption module. For modern designs of new generation NPPs, which do not provide for pressure release systems, we proposed a new passive filtering system together with the passive heat-removal system, which can be used during severe accidents in case all power supply units become unavailable. (authors)

Estimation Of 137Cs Using Atmospheric Dispersion Models After A Nuclear Reactor Accident

Science.gov (United States)

Simsek, V.; Kindap, T.; Unal, A.; Pozzoli, L.; Karaca, M.

2012-04-01

Nuclear energy will continue to have an important role in the production of electricity in the world as the need of energy grows up. But the safety of power plants will always be a question mark for people because of the accidents happened in the past. Chernobyl nuclear reactor accident which happened in 26 April 1986 was the biggest nuclear accident ever. Because of explosion and fire large quantities of radioactive material was released to the atmosphere. The release of the radioactive particles because of accident affected not only its region but the entire Northern hemisphere. But much of the radioactive material was spread over west USSR and Europe. There are many studies about distribution of radioactive particles and the deposition of radionuclides all over Europe. But this was not true for Turkey especially for the deposition of radionuclides released after Chernobyl nuclear reactor accident and the radiation doses received by people. The aim of this study is to determine the radiation doses received by people living in Turkish territory after Chernobyl nuclear reactor accident and use this method in case of an emergency. For this purpose The Weather Research and Forecasting (WRF) Model was used to simulate meteorological conditions after the accident. The results of WRF which were for the 12 days after accident were used as input data for the HYSPLIT model. NOAA-ARL's (National Oceanic and Atmospheric Administration Air Resources Laboratory) dispersion model HYSPLIT was used to simulate the 137Cs distrubition. The deposition values of 137Cs in our domain after Chernobyl Nuclear Reactor Accident were between 1.2E-37 Bq/m2 and 3.5E+08 Bq/m2. The results showed that Turkey was affected because of the accident especially the Black Sea Region. And the doses were calculated by using GENII-LIN which is multipurpose health physics code.

Biomass accident investigations – missed opportunities for learning and accident prevention

DEFF Research Database (Denmark)

Hedlund, Frank Huess

2017-01-01

The past decade has seen a major increase in the production of energy from biomass. The growth has been mirrored in an increase of serious biomass related accidents involving fires, gas explosions, combustible dust explosions and the release of toxic gasses. There are indications that the number...... of bioenergy related accidents is growing faster than the energy production. This paper argues that biomass accidents, if properly investigated and lessons shared widely, provide ample opportunities for improving general hazard awareness and safety performance of the biomass industry. The paper examines...... selected serious accidents involving biogas and wood pellets in Denmark and argues that such opportunities for learning were missed because accident investigations were superficial, follow-up incomplete and information sharing absent. In one particularly distressing case, a facility saw a repeat accident...

Radiological environment within an NPP after a severe nuclear accident

Science.gov (United States)

Andgren, Karin; Fritioff, Karin; Buhr, Anna Maria Blixt; Huutoniemi, Tommi

2017-09-01

The radiological environment following a severe nuclear accident can be visualised on building layouts. The direct radiation in an area (or room) can be visualized on the layout by a colouring scheme depending on the dose rate level (for example orange for high gamma dose rate level and purple for an intermediate gamma dose rate level). Following the Fukushima accident, a need for update of these layouts has been identified at the Swedish nuclear power plant of Forsmark. Shielding calculations for areas where access is desired for severe accident management have been performed. Many different sources of radiation together with different types of shielding material contribute to the dose that would be received by a person entering the area. External radiation from radioactivity within e.g. pipes and components is considered and also external radiation from radioactivity in the air (originating from diffuse leakage of the containment atmosphere). Results are presented as dose rates for relevant dose points together with a method for estimating the dose rate levels for each of the rooms of the reactor building.

Radiological environment within an NPP after a severe nuclear accident

Directory of Open Access Journals (Sweden)

Andgren Karin

2017-01-01

Full Text Available The radiological environment following a severe nuclear accident can be visualised on building layouts. The direct radiation in an area (or room can be visualized on the layout by a colouring scheme depending on the dose rate level (for example orange for high gamma dose rate level and purple for an intermediate gamma dose rate level. Following the Fukushima accident, a need for update of these layouts has been identified at the Swedish nuclear power plant of Forsmark. Shielding calculations for areas where access is desired for severe accident management have been performed. Many different sources of radiation together with different types of shielding material contribute to the dose that would be received by a person entering the area. External radiation from radioactivity within e.g. pipes and components is considered and also external radiation from radioactivity in the air (originating from diffuse leakage of the containment atmosphere. Results are presented as dose rates for relevant dose points together with a method for estimating the dose rate levels for each of the rooms of the reactor building.

Aftermath of the Fukushima Daiichi nuclear accident in March 2011 - Situation review in March 2016

International Nuclear Information System (INIS)

2016-03-01

The first part of this detailed report addresses the consequences of the accident regarding nuclear safety. It proposes a situation review of site damaged installations, of radioactive water management, and of underground water management. It presents and comments lessons learned from this accident for French nuclear installations, gives an overview of researches performed by the IRSN in the field of nuclear safety. The second part addresses health consequences of the accident. It discusses an assessment of epidemiologic studies performed on inhabitants of the Fukushima Prefecture, and comments the situation of workers involved in operations performed in the Fukushima Daiichi nuclear power plant. The third part addresses environmental consequences. It discusses values of radionuclide concentrations in Japanese air five years after the accident, measurements of caesium activities, assessments of contamination of Japanese food products, decontamination actions and waste management, the status of marine contamination in 2015, the evolution of evacuation areas between 2011 and 2016, the first returns and wills to return of evacuated populations, the update of knowledge related to the dispersion and depositions of atmospheric releases of the accident, and the modelling of atmospheric transport and fallouts of releases emitted during the accident. The last part proposes a comparison between the Chernobyl accident and the Fukushima accident in terms of distribution of radioactive depositions within river basins, of knowledge drawn from ecologic studies on fauna and flora performed on the long term in contaminated areas, and of management of forest environments after a nuclear accident

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

NARCIS (Netherlands)

Oppe, S.

1982-01-01

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

Regulation Plans on Severe Accidents developed by KINS Severe Accident Regulation Preparation TFT

International Nuclear Information System (INIS)

Kim, Kyun Tae; Chung, Ku Young; Na, Han Bee

2016-01-01

Some nuclear power plants in Fukushima Daiichi site had lost their emergency reactor cooling function for long-time so the fuels inside the reactors were molten, and the integrity of containment was damaged. Therefore, large amount of radioactive material was released to environment. Because the social and economic effects of severe accidents are enormous, Korean Government already issued 'Severe Accident Policy' in 2001 which requires nuclear power plant operators to set up 'Quantitative Safety Goal', to do 'Probabilistic Safety Analysis', to install 'Severe Accident Countermeasures' and to make 'Severe Accident Management Plan'. After the Fukushima disaster, a Special Safety Inspection was performed for all operating nuclear power plants of Korea. The inspection team from industry, academia, and research institutes assessed Korean NPPs capabilities to cope with or respond to severe accidents and emergency situation caused by natural disasters such as a large earthquake or tsunami. As a result of the special inspection, about 50 action items were identified to increase the capability to cope with natural disaster and severe accidents. Nuclear Safety Act has been amended to require NPP operators to submit Accident Management Plant as part of operating license application. The KINS Severe Accident Regulation Preparation TFT had first investigated oversea severe accident regulation trend before and after the Fukushima accident. Then, the TFT has developed regulation draft for severe accidents such as Severe accident Management Plans, the required design features for new NPPs to prevent severe accident against multiple failures and beyond-design external events, countermeasures to mitigate severe accident and to keep the integrity of containment, and assessment methodology on safety assessment plan and probabilistic safety assessment

Underreporting of maritime accidents to vessel accident databases.

Science.gov (United States)

Hassel, Martin; Asbjørnslett, Bjørn Egil; Hole, Lars Petter

2011-11-01

Underreporting of maritime accidents is a problem not only for authorities trying to improve maritime safety through legislation, but also to risk management companies and other entities using maritime casualty statistics in risk and accident analysis. This study collected and compared casualty data from 01.01.2005 to 31.12.2009, from IHS Fairplay and the maritime authorities from a set of nations. The data was compared to find common records, and estimation of the true number of occurred accidents was performed using conditional probability given positive dependency between data sources, several variations of the capture-recapture method, calculation of best case scenario assuming perfect reporting, and scaling up a subset of casualty information from a marine insurance statistics database. The estimated upper limit reporting performance for the selected flag states ranged from 14% to 74%, while the corresponding estimated coverage of IHS Fairplay ranges from 4% to 62%. On average the study results document that the number of unreported accidents makes up roughly 50% of all occurred accidents. Even in a best case scenario, only a few flag states come close to perfect reporting (94%). The considerable scope of underreporting uncovered in the study, indicates that users of statistical vessel accident data should assume a certain degree of underreporting, and adjust their analyses accordingly. Whether to use correction factors, a safety margin, or rely on expert judgment, should be decided on a case by case basis. Copyright © 2011 Elsevier Ltd. All rights reserved.

PAT-2 (Plutonium Air Transportable Model 2)

International Nuclear Information System (INIS)

Anderson, J.

1981-01-01

The PAT-2 (Plutonium Air Transportable Model 2) package is designed for the safe transport of plutonium and/or uranium in small quantities, especially as used in international safeguards activities, and especially as transported by air. The PAT-2 package is resistant to severe accidents, including that of a high-speed jet aircraft crash, and is designed to withstand such environments as extreme impact, crushing, puncturing and slashing loads, severe hydrocarbon-fueled fires, and deep underwater immersion, with no escape of contents. The accident environments may be imposed upon the package singly or seqentially. The package meets the requirements of 10 CFR 71 for Fissile Class I packages with a cargo of 15 grams of Pu-239, or other isotopic forms described herein, not to exceed 2 watts of thermal activity. Packaging, operational features, and contents of package, are discussed

Early measurements in urban areas after the Chernobyl accident

International Nuclear Information System (INIS)

Likhtarev, I.

2000-01-01

This paper summarises the experience on the radioactive monitoring of the environment and population dose assessment provided in urban areas, mainly in Kiev, after the Chernobyl accident. It emphasises the need of several radiological teams, of the support from several institutions and of preparedness for a consistent database, dose assessment and criteria for decision making. Main results of measurements of gamma exposure rates, air, grass and food radioactive contamination are presented. (author)

OECD MCCI project Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-1 test data report: thermal hydraulic results. Rev. 0 September 20, 2002

International Nuclear Information System (INIS)

Lomperski, S.; Farmer, M.T.; Kilsdonk, D.J.; Aeschlimann, R.W.; Basu, S.

2011-01-01

The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium (∼φ30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the first water ingression test, designated SSWICS-1. The report includes a description of the test apparatus, the

ARGX-87: Accident Response Group Exercise, 1987: A Broken Arrow mini exercise

International Nuclear Information System (INIS)

Schuld, E.P.; Cruff, D.F.

1987-07-01

A Broken Arrow mini exercise dubbed ''Accident Response Group Exercise - 1987'' (ARGX-87) was conducted on June 1, 1987 at the Lawrence Livermore National Laboratory (LLNL) and Sandia National Laboratories, Livermore (SNLL). The exercise started at 0445 PDT with a call from the Department of Energy (DOE) - EOC in Washington, DC, to the Albuquerque Operations (AL - ) - EOC. AL, in turn, called the Laboratory off-hour emergency number (Fire Dispatcher), who called the Laboratory Emergency Duty Officer (LEDO). The LEDO then contacted the Accident Response Group (ARG) Senior Scientific Advisor. Calls were placed to assemble appropriate members of the ARG in the ALERT Center. No phone number for SNLL was available at the Albuquerque Operations EOC, so a controller injected a message to SNLL to get them involved in the exercise. The messages received at the Laboratory identified the Air Force line item weapon system involved in the accident and the accident location. As people arrived at the ALERT Center they began discussing the details of the accident. They also started working the deployment logistics and other issues. Travel arrangements for the HOT SPOT equipment and ARG personnel were made for immediate deployment to the accident site in North Dakota. The exercise was terminated at 0840 as planned. While certain procedural deficiencies were noted, the exercise was considered a valuable learning experience. The results and observations from this experience will be used to refine the operating procedures and the training program

Magnitude and reactivity consequences of accidental moisture ingress into the Modular High-Temperature Gas-Cooled Reactor core

International Nuclear Information System (INIS)

Smith, O.L.

1992-01-01

Accidental admission of moisture into the primary system of a Modular High-Temperature Gas-Cooled Reactor (MHTGR) has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moistureingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. Rate and magnitude of steam buildup are found to be dominated by major system features such as break size in comparison with safety valve capacity and reliability, while being less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions

Regulation Plans on Severe Accidents developed by KINS Severe Accident Regulation Preparation TFT

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyun Tae; Chung, Ku Young; Na, Han Bee [KINS, Daejeon (Korea, Republic of)

2016-05-15

Some nuclear power plants in Fukushima Daiichi site had lost their emergency reactor cooling function for long-time so the fuels inside the reactors were molten, and the integrity of containment was damaged. Therefore, large amount of radioactive material was released to environment. Because the social and economic effects of severe accidents are enormous, Korean Government already issued 'Severe Accident Policy' in 2001 which requires nuclear power plant operators to set up 'Quantitative Safety Goal', to do 'Probabilistic Safety Analysis', to install 'Severe Accident Countermeasures' and to make 'Severe Accident Management Plan'. After the Fukushima disaster, a Special Safety Inspection was performed for all operating nuclear power plants of Korea. The inspection team from industry, academia, and research institutes assessed Korean NPPs capabilities to cope with or respond to severe accidents and emergency situation caused by natural disasters such as a large earthquake or tsunami. As a result of the special inspection, about 50 action items were identified to increase the capability to cope with natural disaster and severe accidents. Nuclear Safety Act has been amended to require NPP operators to submit Accident Management Plant as part of operating license application. The KINS Severe Accident Regulation Preparation TFT had first investigated oversea severe accident regulation trend before and after the Fukushima accident. Then, the TFT has developed regulation draft for severe accidents such as Severe accident Management Plans, the required design features for new NPPs to prevent severe accident against multiple failures and beyond-design external events, countermeasures to mitigate severe accident and to keep the integrity of containment, and assessment methodology on safety assessment plan and probabilistic safety assessment.

Specific features of RBMK severe accidents progression and approach to the accident management

International Nuclear Information System (INIS)

Vasilevskij, V.P.; Nikitin, Yu.M.; Petrov, A.A.; Potapov, A.A.; Cherkashov, Yu.M.

2001-01-01

Fundamental construction features of the LWGR facilities (absence of common external containment shell, disintegrated circulation circuit and multichannel reactor core, positive vapor reactivity coefficient, high mass of thermally capacious graphite moderator) predetermining development of assumed heavy non-projected accidents and handling them are treated. Rating the categories of the reactor core damages for non-projected accidents and accident types producing specific grope of damages is given. Passing standard non-projected accidents, possible methods of attack accident consequences, as well as methods of calculated analysis of non-projected accidents are demonstrated [ru

Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

International Nuclear Information System (INIS)

Sdouz, Gert

2006-01-01

The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the un-tightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the 'Station Blackout'- sequence and the 'Large Break LOCA'. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was

SEVERE ACCIDENT ISSUES RAISED BY THE FUKUSHIMA ACCIDENT AND IMPROVEMENTS SUGGESTED

OpenAIRE

SONG, JIN HO; KIM, TAE WOON

2014-01-01

This paper revisits the Fukushima accident to draw lessons in the aspect of nuclear safety considering the fact that the Fukushima accident resulted in core damage for three nuclear power plants simultaneously and that there is a high possibility of a failure of the integrity of reactor vessel and primary containment vessel. A brief review on the accident progression at Fukushima nuclear power plants is discussed to highlight the nature and characteristic of the event. As the severe accide...

Experience of air transport of nuclear fuel material as type A package

International Nuclear Information System (INIS)

Kawasaki, Masashi; Kageyama, Tomio; Suzuki, Toru

2004-01-01

Special law on nuclear disaster countermeasures (hereafter called as to nuclear disaster countermeasures low) that is domestic law for dealing with measures for nuclear disaster, was enforced in June, 2000. Therefore, nuclear enterprise was obliged to report accidents as required by nuclear disaster countermeasures law, besides meeting the technical requirement of existent transport regulation. For overseas procurement of plutonium reference materials that are needed for material accountability, A Type package must be transported by air. Therefore, concept of air transport of nuclear fuel materials according to the nuclear disaster countermeasures law was discussed, and the manual including measures against accident in air transport was prepared for the oversea procurement. In this presentation, the concept of air transport of A Type package containing nuclear fuel materials according to the nuclear disaster countermeasures law, and the experience of a transportation of plutonium solution from France are shown. (author)

Radioactivity of people in Finland after the Chernobyl accident in 1986

International Nuclear Information System (INIS)

Rahola, T.; Suomela, M.; Illukka, E.; Puhakainen, M.; Pusa, S.

1987-06-01

After the reactor accident at Chernobyl on April 26, 1986 radioactive fallout was carried by air currents to most parts of Europe. The radioactive air currents reached Finland on April 27. Immediately after the arrival of such air in Finland, contamination of people by radioactive nuclides began via inhalation of this air. The ingestion route become important later, when radionuclides were transported via different foodchains to man. To determine the level of radionuclides in the body and to estimate the internal radiation doses caused by the Chernobyl accident, whole-body counting measurements were performed. The results of whole-body counting of six different groups of Finnish people measured during 1986 after the accident at Chernobyl are reported. Three were reference groups measured routinely once or twice annually, two groups were comprised of workers at nuclear power stations and one group consisted of 262 persons not belonging to any other group. The total number of whole-body counting measurements of persons in these groups in 1986 from the end of April to the end of December was 624. In April and May small amounts of 131 I were detected in the thyroid. In June the first signs of 134 Cs in the body were noticed. At the end of 1986 the mean 134 Cs body burden for women and men in the Helsinki reference group was 730 Bq. The mean 137 Cs body burden in women and men increased from 150 Bq in June to 1500 Bq in December. The differences in the deposition levels in the five fallout regions into which Finland was divided were reflected in the activity levels of the measured people. The weighted annual mean body burden for people in Finland at the end of 1986 was 370 Bq 134 Cs and 820 Bq 137 Cs. The maximum body burdens of 134 Cs and 137 Cs found in Finns in 1986 were 6300 Bq and 13000 Bq, respectively. The mean internal committed effective dose equivalent 0.06 mSv from 134 Cs and 137 Cs in the body of Finnish people in 1986 was calculated using the whole

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

Feedback from practical experience with large sodium fire accidents

International Nuclear Information System (INIS)

Luster, V.P.; Freudenstein, K.F.

1996-01-01

contact with the concrete floor and walls. Sodium concrete reactions and concrete destruction were the consequence. The Almeria accident had previously drawn attention to the potential problem of Na aerosol release into the environment. As consequence the local environment protection group took samples of the air around the Bensberg facility during the accident. No noxious fall out has been found although post accident analyses have shown that the amount of released aerosols was about 1000 kg. The accident confirmed also that severe damage was limited to the immediate fire zone. Neither the Almeria spray fire accident nor the ILONA sodium concrete reactions would have endangered the nuclear safety if they would have occurred at a reactor plant (not speaking of investment losses). The ILONA building was cleaned. After repair the building was ready for another use. (author)

Nuclear accident dosimetry

International Nuclear Information System (INIS)

1982-01-01

The film presents statistical data on criticality accidents. It outlines past IAEA activities on criticality accident dosimetry and the technical documents that resulted from this work. The film furthermore illustrates an international comparison study on nuclear accident dosimetry conducted at the Atomic Energy Research Establishment, Harwell, United Kingdom

Nuclear accident dosimetry

Energy Technology Data Exchange (ETDEWEB)

NONE

1983-12-31

The film presents statistical data on criticality accidents. It outlines past IAEA activities on criticality accident dosimetry and the technical documents that resulted from this work. The film furthermore illustrates an international comparison study on nuclear accident dosimetry conducted at the Atomic Energy Research Establishment, Harwell, United Kingdom

Planning for large-scale accidents: learning from the Three Mile Island accident

International Nuclear Information System (INIS)

Fischer, D.W.

1981-01-01

Decision-making issues raised at the Three Mile Island nuclear accident in Pennsylvania are explored. The organizations involved, their interconnections, and decisions are described. The underlying issues bearing on allocation of effort to pre-accident planning and actual accident responses are also noted. Finally, a framework from this effort is used for guiding the planning of operations for future accidents. (author)

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)

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.)

Weather Information Communications (WINCOMM) Project: Dissemination of Weather Information for the Reduction of Aviation Weather-Related Accident Causal Factors

Science.gov (United States)

Jarrell, Michael; Tanger, Thomas

2004-01-01

Weather Information Communications (WINCOMM) is part of the Weather Accident Prevention (WxAP) Project, which is part of the NASA's Aviation Safety and Security Program. The goals of WINCOMM are to facilitate the exchange of tactical and strategic weather information between air and ground. This viewgraph presentation provides information on data link decision factors, architectures, validation goals. WINCOMM is capable of providing en-route communication air-to-ground, ground-to-air, and air-to-air, even on international or intercontinental flights. The presentation also includes information on the capacity, cost, and development of data links.