WorldWideScience

Sample records for accident thermal challenge

  1. HANARO thermal hydraulic accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chul; Kim, Heon Il; Lee, Bo Yook; Lee, Sang Yong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    For the safety assessment of HANARO, accident analyses for the anticipated operational transients, accident scenarios and limiting accident scenarios were conducted. To do this, the commercial nuclear reactor system code. RELAP5/MOD2 was modified to RELAP5/KMRR; the thermal hydraulic correlations and the heat exchanger model was changed to incorporate HANARO characteristics. This report summarizes the RELAP/KMRR calculation results and the subchannel analyses results based on the RELAP/KMRR results. During the calculation, major concern was placed on the integrity of the fuel. For all the scenarios, the important accident analysis parameters, i.e., fuel centerline temperatures and the minimum critical heat flux ratio(MCHFR), satisfied safe design limits. It was verified, therefore, that the HANARO was safely designed. 21 tabs., 89 figs., 39 refs. (Author) .new.

  2. Potential steam generator tube rupture in the presence of severe accident thermal challenge and tube flaws due to foreign object wear

    International Nuclear Information System (INIS)

    Liao, Y.; Guentay, S.

    2009-01-01

    This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with counter-current natural circulating high temperature gas in the hot leg and SG tubes. The considered SG tube flaws are caused by foreign object wear, which in recent years has emerged as a major inservice degradation mechanism for the new generation tubing materials. The first step develops the statistical distributions for the flaw frequency, size, and the flaw location with respect to the tube length and the tube's tubesheet position, based on data of hundreds of flaws reported in numerous SG inservice inspection reports. The next step performs thermal-hydraulic analysis using the MELCOR code and recent CFD findings to predict the thermal challenge to the degraded tubes and the tube-to-tube difference in thermal response at the SG entrance. The final step applies the creep rupture models in the Monte Carlo random walk to test the potential for the degraded SG to rupture before the surge line. The mean and range of the SG tube rupture probability can be applied to estimate large early release frequency in probabilistic safety assessment.

  3. Thermal hydraulic features of the TMI accident

    International Nuclear Information System (INIS)

    Tolman, B.

    1985-01-01

    The TMI-2 accident resulted in extensive core damage and recent data confirms that the reactor vessel was challenged from molten core materials. A hypothesized TMI accident sencario is presented that consistently explains the TMI data and is also consistent with research findings from independent severe fuel damage experiements. The TMI data will prove useful in confirming our understanding of severe core damage accidents under realistic reactor systems conditions. This understanding will aid in addressing safety and regulatory issues related to severe core damage accidents in light water reactors

  4. Accomplishments and challenges of the severe accident research

    International Nuclear Information System (INIS)

    Sehgal, B.R.

    2001-01-01

    This paper briefly describes the progress of the severe accident research since 1980, in terms of the accomplishments made so far and the challenges that remain. Much has been accomplished: many important safety issues have been resolved and consensus is near on some others. However, some of the previously identified safety issues remain as challenges, while some new ones have arisen due to the shift in focus from containment to vessel integrity. New reactor designs have also created some new challenges. In general, the regulatory demands for new reactor designs are stricter, thereby requiring much greater attention to the safety issues concerned with the containment design of the new large reactors, and to the accident management procedures for mitigating the consequences of a severe accident. We apologize for not providing references to many fine investigations that contributed to the great progress made so far in the severe accident research

  5. Thermal hydraulics of CANDU severe accident analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents in CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D 2 O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the available data. The results are encouraging. (authors)

  6. Preventing Drowning Accidents Using Thermal Cameras

    DEFF Research Database (Denmark)

    Bonderup, Søren; Olsson, Jonas Lundgaard; Bonderup, Morten Bojesen

    2016-01-01

    detector is implemented using a virtual trip-wire in combination with an optical flow algorithm making the system able to detect 100% of all falls and only yielding a 0.08 false positive rate hourly. The entire system has been developed using 155 h of real life thermal video, hereof 56 h are manually...

  7. Accident tolerant fuel cladding development: Promise, status, and challenges

    Science.gov (United States)

    Terrani, Kurt A.

    2018-04-01

    The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

  8. Challenges within ventilation systems during accident situations

    International Nuclear Information System (INIS)

    Neuberger, M.; Bletzer, P.; Fronhoefer, M.; Hardy, T.; Kuhn, W.; Ricketts, C.I.; Schmidmeier, K.H.

    1991-01-01

    To help estimate the mechanical loading at the service location two fluid dynamic codes were investigated. FLOWMASTER shows good results in generating the air-cleaning network and in calculating steady state conditions, but fails in calculating fluid dynamic transients. PROMO is very time consuming in generating the network, but one can obtain good results in calculating fluid dynamic transients. In addition to the challenges posed by elevated pressure, temperature, and air flow, the effects of shock waves, caused by explosions or hydrogen detonations must be considered. The problem of a 'shock-shock' in the flow after the diverging channel by high shock strengths were solved. In order to guarantee the safety margins of filter units during their entire service lives, filter performance needs to be verified under standardized test conditions that take into consideration the particularly adverse effects of filter exposure to super-saturated airflows and elevated differential pressure. The detail construction for a rig to type test filter units under fog conditions and increased pressure drop was finished, and the components of the test facility were chosen. (orig./DG)

  9. Accomplishments and challenges of the severe accident research

    International Nuclear Information System (INIS)

    Sehga, B.R.

    1998-01-01

    This paper describes the progress of the severe accident research since 1980, in terms of the accomplishments made so far and the challenges that remain. Much has been accomplished: many important safety issues have been resolved and consensus is near on some others. However, some of the previously identified safety issues remain as challenges, while some new ones have arisen due to the shift in focus from containment integrity to vessel integrity. New reactor designs have also created some new challenges. In general, the regulatory demands in new reactor designs are much stricter, thereby requiring much greater attention to the safety issues concerned with the containment design of the new large reactors

  10. Challenging the assumptions for thermal sensation scales

    DEFF Research Database (Denmark)

    Schweiker, Marcel; Fuchs, Xaver; Becker, Susanne

    2016-01-01

    Scales are widely used to assess the personal experience of thermal conditions in built environments. Most commonly, thermal sensation is assessed, mainly to determine whether a particular thermal condition is comfortable for individuals. A seven-point thermal sensation scale has been used...... extensively, which is suitable for describing a one-dimensional relationship between physical parameters of indoor environments and subjective thermal sensation. However, human thermal comfort is not merely a physiological but also a psychological phenomenon. Thus, it should be investigated how scales for its...... assessment could benefit from a multidimensional conceptualization. The common assumptions related to the usage of thermal sensation scales are challenged, empirically supported by two analyses. These analyses show that the relationship between temperature and subjective thermal sensation is non...

  11. EPR design features to mitigate severe accident challenges

    International Nuclear Information System (INIS)

    Mazurkiewicz, S.M.; Fischer, M.; Bittermann, D.

    2005-01-01

    The EPR, an evolutionary pressurized water reactor (PWR), is a 4300-4500 MWth that incorporates proven technology within an optimized configuration to enhance safety. EPR was originally developed through a joint effort between Framatome ANP and Siemens by incorporating the best technological features from the French and German nuclear reactor fleets into a cost-competitive product. Commercial EPR units are currently being built in Finland at the Olkiluoto site, and planned for France at the Flamanville site. In recent months, Framatome ANP announced their intention to market the EPR units to China in response to a request for vendor bids as well as their intent to pursue design certification in the United States under 10CFR52. The EPR safety philosophy is based on a deterministic consideration of defense-in-depth complemented by probabilistic analyses. Not only is the EPR designed to prevent and mitigate design basis accidents (DBAs), it employs an extra level of safety associated with severe accident response. Therefore, as a design objective, features are included to ensure that radiological consequences are limited such that the need for stringent counter measures, such as evacuation and relocation of the nearby population, can be reasonably excluded. This paper discusses some of the innovative features of the EPR to address severe accident challenges. (author)

  12. Accident prevention ordinance 2.0 Thermal Power Plants

    International Nuclear Information System (INIS)

    Egyptien, H.H.; Fischermann, E.

    This accident prevention ordinance is to cover primarily the very section of a power station where fossil or nuclear energy is converted into thermal energy, e.g. by heating or vaporization of a heat source. In paragraph 1, 40 GJ/h are stipulated as the lower limit of capacity corresponding to about 11 MW. Therefore, the accident prevention ordinance does not only marshal the operation of steam generators in electricity supply utilities but also covers smaller industrial power stations which partly do only meet the company's own requirements. Pipes are only covered as far as they are operated in conjunction with a heat generator. The same applies to coal handling and ash removal facilities. This means that for heat release e.g. in the framework of a district heating grid, the transfer station to the distribution grid is regarded as being a border of the power station and thus a border to the area of application of the accident prevention ordinance. (orig./HP) [de

  13. Technological challenges in thermal plasma production

    International Nuclear Information System (INIS)

    Ramakrishnan, S.

    1995-01-01

    Thermal plasmas, generated by electric arc discharges, are used in a variety of industrial applications. The electric arc is a constricted electrical discharge with a high temperature in the range 6000-25,000 K. These characteristics are useful in plasma cutting, spraying, welding and specific areas of material processing. The thermal plasma technology is an enabling process technology and its status in the market depends upon its advantages over competing technologies. A few technological challenges to enhance the status of plasma technology are to improve the utilisation of the unique characteristics of the electric arc and to provide enhanced control of the process. In particular, new solutions are required for increasing the plasma-material interaction, controlling the electrode roots and controlling the thermal power generated by the arcing process. In this paper, the advantages of plasma technology, its constraints and future challenges for technology developments are highlighted. 36 refs., 14 figs

  14. Thermal-hydraulic uncertainties affecting severe accident progression

    International Nuclear Information System (INIS)

    Haskin, F.E.; Behr, V.L.

    1984-01-01

    To provide the proper technical bases for decisions regarding severe accidents, the US Nuclear Regulatory Commission (NRC) is sponsoring the following activities: (a) a variety of severe accident research programs, combined under the Severe Accident Research Plan; (b) nationwide task forces on containment loading, containment response, and fission product source terms; (c) a review by the American Physical Society of state-of-the-art methods for calculating radiological source terms; and (d) technical exchange meetings with the Industry Degraded Core (IDCOR) program. One of the means for integrating this developing array of technical information is the Severe Accident Risk Reduction Program (SARRP). One of the current SARRP objectives is to utilize insights gained from the activities listed above to characterize the relative likelihoods of competing containment failure modes for core-melt accidents

  15. An unmet challenge: consideration of heavy fouling in severe accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Robert H Leyse [CEO Inz, Inc., PO BOX 2850, Sun Valley, ID 83353 (United States)

    2005-07-01

    Full text of publication follows: The impact of heavy fouling of fuel elements in light water cooled and moderated nuclear reactors has not been considered in the analysis of severe accidents such as Reactivity Insertion Accidents and Loss of Coolant Accidents. This is the case even though operation of nuclear power reactors with significant fouling deposits is commonplace. Fouling deposits have substantial thermal resistance. This has led to fuel element failures in several instances as the zirconium alloy cladding has failed due to high temperature corrosion. Although the details of current fouling have not been disclosed, in one case the deposits have been described as, 'unusually heavy' which induced the corrosion by thermally insulating the fuel rods' and 'rods that failed had heavy crud with clumpy formations'. Such heavy clumpy fouling is complex with substantial thermal resistance. Relatively straightforward fouling at the Experimental Boiling Water Reactor was classified in terms of the thickness and the thermal conductivity.7 Thickness of the fouling was 0.013 cm, the thermal conductivity was 0.008 W/cm-C; thus the heat transfer coefficient was 0.6 W/(cm{sup 2})(C). The peak heat flux in today's large light water reactors is in the range of 150 W/cm{sup 2} and the temperature gradient for EBWR-type fouling would be 250 C. However, the effective heat transfer coefficient of the heavy, clumpy fouling in today's reactors is likely substantially less than the EBWR case. Clearly, the heat transfer characteristics are vastly degraded in contrast to clean as-built cores. The challenge for the licensees of nuclear power reactors is to produce thorough evaluations of the impact of heavy fouling on severe accidents. The findings are needed for the accurate licensing of water-cooled nuclear reactors. The United States Nuclear Regulatory Commission has evaluated two related Petitions for Rulemaking that have been initiated by the

  16. An unmet challenge: consideration of heavy fouling in severe accident analyses

    International Nuclear Information System (INIS)

    Robert H Leyse

    2005-01-01

    Full text of publication follows: The impact of heavy fouling of fuel elements in light water cooled and moderated nuclear reactors has not been considered in the analysis of severe accidents such as Reactivity Insertion Accidents and Loss of Coolant Accidents. This is the case even though operation of nuclear power reactors with significant fouling deposits is commonplace. Fouling deposits have substantial thermal resistance. This has led to fuel element failures in several instances as the zirconium alloy cladding has failed due to high temperature corrosion. Although the details of current fouling have not been disclosed, in one case the deposits have been described as, 'unusually heavy' which induced the corrosion by thermally insulating the fuel rods' and 'rods that failed had heavy crud with clumpy formations'. Such heavy clumpy fouling is complex with substantial thermal resistance. Relatively straightforward fouling at the Experimental Boiling Water Reactor was classified in terms of the thickness and the thermal conductivity.7 Thickness of the fouling was 0.013 cm, the thermal conductivity was 0.008 W/cm-C; thus the heat transfer coefficient was 0.6 W/(cm 2 )(C). The peak heat flux in today's large light water reactors is in the range of 150 W/cm 2 and the temperature gradient for EBWR-type fouling would be 250 C. However, the effective heat transfer coefficient of the heavy, clumpy fouling in today's reactors is likely substantially less than the EBWR case. Clearly, the heat transfer characteristics are vastly degraded in contrast to clean as-built cores. The challenge for the licensees of nuclear power reactors is to produce thorough evaluations of the impact of heavy fouling on severe accidents. The findings are needed for the accurate licensing of water-cooled nuclear reactors. The United States Nuclear Regulatory Commission has evaluated two related Petitions for Rulemaking that have been initiated by the author regarding these matters. Currently the

  17. Thermal Hydraulic design parameters study for severe accidents using neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Chang Hyun; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chang, Keun Sun [Sunmoon University, Asan (Korea, Republic of)

    1997-12-31

    To provide the information on severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore, was performed to investigate the effect of thermal hydraulic design parameters on severe accident progression of pressurized water reactors (PWRs). Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among nine parameters. For training, different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3 and 4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout (SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to the other six parameters. 9 refs., 5 tabs. (Author)

  18. Thermal Hydraulic design parameters study for severe accidents using neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Chang Hyun; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chang, Keun Sun [Sunmoon University, Asan (Korea, Republic of)

    1998-12-31

    To provide the information on severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore, was performed to investigate the effect of thermal hydraulic design parameters on severe accident progression of pressurized water reactors (PWRs). Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among nine parameters. For training, different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3 and 4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout (SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to the other six parameters. 9 refs., 5 tabs. (Author)

  19. Unsolved issues related to thermal-hydraulics in the suppression chamber during Fukushima Daiichi accident progressions

    International Nuclear Information System (INIS)

    Mizokami, Shinya; Yamada, Daichi; Honda, Takeshi; Yamauchi, Daisuke; Yamanaka, Yasunori

    2016-01-01

    On 11 March 2011, the Great East Japan Earthquake and Tsunami hit the Fukushima Daiichi Nuclear Power Station. The Fukushima Daiichi Units 1-3 lost all DC and AC power supplies, which set in motion a chain of events that led to releases of radioactivity to the environment. Since then, TEPCO has made many efforts to investigate the accident progressions and the status of the reactors and containment vessels. However, there still exist several tens of unsolved issues to be investigated for the fully understanding of the accident. In this paper, we introduce the unsolved issues related to thermal-hydraulics in the suppression chamber during the Fukushima Daiichi accident progressions. Especially, in Units 2 and 3, there are possibilities that thermal stratification inside their suppression chambers played an important role. It is important that these phenomena are addressed following both theoretical and experimental approaches as support to severe accident simulations. (author)

  20. Prediction of thermal hydraulic parameters in the loss of coolant accident by using artificial neural networks

    International Nuclear Information System (INIS)

    Vaziri, N.; Erfani, A.; Monsefi, M.; Hajabri, A.

    2008-01-01

    In a reactor accident like loss of coolant accident , one or more signals may not be monitored by control panel for some reasons such as interruptions and so on. Therefore a fast alternative method could guarantee the safe and reliable exploration of nuclear power planets. In this study, we used artificial neural network with Elman recurrent structure to predict six thermal hydraulic signals in a loss of coolant accident after upper plenum break. In the prediction procedure, a few previous samples are fed to the artificial neural network and the output value or next time step is estimated by the network output. The Elman recurrent network is trained with the data obtained from the benchmark simulation of loss of coolant accident in VVER. The results reveal that the predicted values follow the real trends well and artificial neural network can be used as a fast alternative prediction tool in loss of coolant accident

  1. Simulation of thermal response of the 250 MWT modular HTGR during hypothetical uncontrolled heatup accidents

    International Nuclear Information System (INIS)

    Harrington, R.M.; Ball, S.J.

    1985-01-01

    One of the central design features of the 250 MWT modular HTGR is the ability to withstand uncontrolled heatup accidents without severe consequences. This paper describes calculational studies, conducted to test this design feature. A multi-node thermal-hydraulic model of the 250 MWT modular HTGR reactor core was developed and implemented in the IBM CSMP (Continuous System Modeling Program) simulation language. Survey calculations show that the loss of forced circulation accident with loss of steam generator cooling water and with accidental depressurization is the most severe heatup accident. The peak hot-spot fuel temperature is in the neighborhood of 1600 0 C. Fuel failure and fission product releases for such accidents would be minor. Sensitivity studies show that code input assumptions for thermal properties such as the side reflector conductivity have a significant effect on the peak temperature. A computer model of the reactor vessel cavity concrete wall and its surrounding earth was developed to simulate the extremely unlikely and very slowly-developing heatup accident that would take place if the worst-case loss of forced primary coolant circulation accident were further compounded by the loss of cooling water to the reactor vessel cavity liner cooling system. Results show that the ability of the earth surrounding the cavity to act as a satisfactory long-term heat sink is very sensitive to the assumed rate of decay heat generation and on the effective thermal conductivity of the earth

  2. The accident prevention regulation 'Thermal Power Stations' and its effects in practice

    International Nuclear Information System (INIS)

    Albert, O.

    1983-01-01

    The origin of the accident prevention regulation - ''Thermal Power Stations'' is attributable mainly to two tragic accidents. It has made organizational changes and interventions in the operational process necessary in thermal power stations. Emphasis is laid upon the consistent issue of written permits-to-work on plant components carrying a heating medium and operating under pressure and on written operating licences for the operation of boilers. The paper describes additional ways in which regulation influences the daily practices of the power station operator. Brief references is made to the draft of the revised regulation. (orig./HP) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  4. 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

  5. Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code

    International Nuclear Information System (INIS)

    Meng Lin; Rui Hu; Yun Su; Ronghua Zhang; Yanhua Yang

    2005-01-01

    Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent

  6. Flexural behavior and design of steel-plate composite (SC) walls for accident thermal loading

    Energy Technology Data Exchange (ETDEWEB)

    Booth, Peter N., E-mail: boothpn@purdue.edu [Lyles School of Civil Engineering, Purdue University, West Lafayette, IN (United States); Varma, Amit H., E-mail: ahvarma@purdue.edu [Lyles School of Civil Engineering, Purdue University, West Lafayette, IN (United States); Sener, Kadir C., E-mail: ksener@purdue.edu [Lyles School of Civil Engineering, Purdue University, West Lafayette, IN (United States); Malushte, Sanjeev R. [Bechtel Corp., Frederick, MD (United States)

    2015-12-15

    Modular steel-plate composite (SC) safety-related nuclear power plant structures must be designed to resist accident thermal and mechanical loads. The design accident thermal load represents the condition where high pressure and temperature steam is released as result of a mechanical failure and applied against the surfaces of power plant structural walls. The effect of heating and pressure can have both short and long term effects on the mechanical integrity of SC structures including degradation and cracking of concrete infill, residual stresses, and out-of-plane deformations. The purpose of this research is to study the effects of thermal and mechanical loads on the out-of-plane flexural response of SC walls and to develop simplified equations that can be used to predict behavior. Four experimental beam tests are reported that represent full-scale cross-sections of SC walls subjected to combinations of mechanical and thermal loads. The study determined that thermal loads reduce the out-of-plane flexural stiffness of SC walls. For the ambient condition, the flexural stiffness closely matches a conventional elastic cracked-transformed model, and at elevated temperatures, the stiffness is reduced to a fully-cracked flexural stiffness that only takes into account the stiffness of the steel faceplates. A method is presented for estimating the thermal curvature, ϕ{sub th}, and thermal moment, M{sub th}, resulting from unequal heating of opposing faces of an SC wall. Based on the tests in this study, the application of accident thermal loads did not result in a reduction of the flexural strength of the SC section.

  7. Thermal hydraulic issues and challenges for current and new generation FBRs

    Energy Technology Data Exchange (ETDEWEB)

    Chellapandi, P.; Velusamy, K., E-mail: kvelu@igcar.gov.in

    2015-12-01

    Highlights: • We present challenges in thermal hydraulic design of sodium cooled fast reactors. • We present roadmap of Indian fast reactor program and innovative design concepts. • Analysis methodology for thermal striping and thermal stratification are highlighted. • Design solutions for gas entrainment are presented. • Experimental approaches for normal and post accident decay heat removal are highlighted. - Abstract: Pool type sodium cooled fast reactors pose several design challenges and among them, certain thermal hydraulics and structural mechanics issues are special. High frequency temperature fluctuations due to thermal striping, thermal stratifications and sodium free level fluctuations at the liquid–cover gas interfaces are to be investigated carefully to eliminate high cycle thermal fatigue of structures. Solutions to address the core thermal hydraulics call for high power computing. Innovative concepts and methods are developed to carry out plant dynamics and safety studies. Particularly, extensive numerical and experimental simulation techniques are needed for understanding and solving the gas entrainment mechanisms and its effects on core safety. Though decay heat removal through natural convection is achievable in a pool type SFR, demonstration of design solutions conceived in the reactor and performance of diverse systems under all operating conditions, especially over prolonged station blackout situations needs advanced CFD computations and should be validated by relatively large scale simulated experiments. These issues are addressed in this paper under five broad topics: special thermal hydraulic issues to be addressed in SFR, thermal hydraulic design and analysis, plant dynamics studies, safety studies and evolving thermal hydraulic studies for the future FBRs. The 500 MWe Prototype Fast Breeder Reactor (PFBR) is taken as the reference design for addressing the issues. Indian fast reactor programme is highlighted in the introduction

  8. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Jin; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures.

  9. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    International Nuclear Information System (INIS)

    Lee, Young Jin; Chung, Bub Dong

    2004-01-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures

  10. Thermal-hydraulic analysis of the Three Mile Island Unit 2 reactor accident with THALES code

    International Nuclear Information System (INIS)

    Hashimoto, Kazuichiro; Soda, Kunihisa

    1991-10-01

    The OECD Nuclear Energy Agency (NEA) has established a Task Group in the Committee on the Safety of Nuclear Installations (CSNI) to perform an analysis of Three Mile Island Unit 2 (TMI-2) accident as a standard problem to benchmark severe accident computer codes and to assess the capability of the codes. The TMI-2 Analysis Exercise was performed at the Japan Atomic Energy Research Institute (JAERI) using the THALES (Thermal-Hydraulic Analysis of Loss-of-Coolant, Emergency Core Cooling and Severe Core Damage) - PM1/TMI code. The purpose of the analysis is to verify the capability of THALES-PM1/TMI code to describe accident progression in the actual plant. The present paper describes the final result of the TMI-2 Analysis Exercise performed at JAERI. (author)

  11. Survival tactics within thermally-challenging roosts: heat tolerance ...

    African Journals Online (AJOL)

    Microclimates were thermally challenging, being very hot (>40°C) for several hours daily in summer and autumn, and cold (<10°C) for much of the night in winter Thermal preference tests revealed that the bats actively selected temperature zones (35°- 42°C) in which basal metabolic rate could be maintained, and above the ...

  12. Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code

    International Nuclear Information System (INIS)

    Kljenak, Ivo; Dapper, Maik; Dienstbier, Jiri; Herranz, Luis E.; Koch, Marco K.; Fontanet, Joan

    2010-01-01

    Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant.

  13. HEATUP: a computer program for the thermal anaysis of a LOFC accident in an HTGR

    International Nuclear Information System (INIS)

    Siman-Tov, I.I.; Turner, W.D.

    1976-11-01

    The HEATUP code, a modification of the general, time-dependent, one-, two-, and three-dimensional program HEATING5, was designed for the thermal analysis of a Loss of Forced Circulation accident in a High Temperature Gas-Cooled Reactor. This report contains a description of the computational model which includes: a description of the basic problem; a short review of preliminary results related to the choice of thermal properties, boundary conditions and initial conditions; a full description of a typical three-dimensional R-Z model and a limited one of a two-dimensional RZ model. HEATUP's additional computations are presented together with the method of input preparation. The three-dimensional model of the Fulton Generating Station Loss of Forced Circulation accident is used as a sample problem. A complete presentation of the input data is made. Also, the computer printout of the sample problem input data and results are given

  14. HEATUP: a computer program for the thermal anaysis of a LOFC accident in an HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Siman-Tov, I.I.; Turner, W.D.

    1976-11-01

    The HEATUP code, a modification of the general, time-dependent, one-, two-, and three-dimensional program HEATING5, was designed for the thermal analysis of a Loss of Forced Circulation accident in a High Temperature Gas-Cooled Reactor. This report contains a description of the computational model which includes: a description of the basic problem; a short review of preliminary results related to the choice of thermal properties, boundary conditions and initial conditions; a full description of a typical three-dimensional R-Z model and a limited one of a two-dimensional RZ model. HEATUP's additional computations are presented together with the method of input preparation. The three-dimensional model of the Fulton Generating Station Loss of Forced Circulation accident is used as a sample problem. A complete presentation of the input data is made. Also, the computer printout of the sample problem input data and results are given.

  15. Finite element analysis of thermal stresses of the reactor vessel in a severe light water reactor accident

    International Nuclear Information System (INIS)

    Borovkov, A.I.; Semenov, A.S.; Granovsky, V.S.; Kovtunova, S.V.

    1995-01-01

    The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs

  16. Finite element analysis of thermal stresses of the reactor vessel in a severe light water reactor accident

    Energy Technology Data Exchange (ETDEWEB)

    Borovkov, A.I.; Semenov, A.S. [St. Petersburg State Technical Univ. (Russian Federation); Granovsky, V.S.; Kovtunova, S.V. [Research Inst. of Technology, Sosnovy Bor (Russian Federation)

    1995-12-31

    The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs.

  17. Lessons of TEPCO's Fukushima accident from human and organizational aspects and challenge for nuclear safety reform

    International Nuclear Information System (INIS)

    Kawano, Akira

    2013-01-01

    The author participated in international experts' meeting held by IAEA on May 21, 2013 and presented the paper focusing on human and organizational aspects of the Fukushima nuclear accident. It clarified TEPCO's basic recognition: 'The cause of the accident should not be treated merely as a natural disaster due to an enormous tsunami being something difficult to anticipate and we believe it is necessary to seriously acknowledge the result that TEPCO failed to avoid an accident which might have been avoided if ample preparations had been made in advance with thorough use of human intellect' and then reconsidered the Fukushima nuclear accident: 'could we predict an enormous tsunami and take whatever countermeasures?' and 'could we respond to the accident better?' for the worldwide operators to avoid such an accident, which moved meeting's participants deeply. Presentation's contents followed 'Reassessment of the Fukushima Nuclear Accident and Nuclear Safety Reform Plan' published by TEPCO on March 29. This article described outline of the presentation. Though the only way to explore the possibility to save Unit 1 was that operators could bravely go up to the 4th floor of reactor building and open the isolation valves to start IC, it was given up without any clear communication among key decision makers for confirming the IC operational status. As for Unit 3, operators could not achieve thorough focus on ensuring core cooling such that proactive transfer from RCIC/HPCI to low pressure water injection was not challenged, mainly because of low trust on Diesel/Driven Fire Protection Pump (DDFP). During the design stage and afterward, ample consideration was not given to common cause failures originating in external events, which led to a severe situation where almost all the power supplies and safety system functions were lost. Continuous efforts to reduce risks were not ample, including the collection, analysis and utilization of information on safety enhancement

  18. IRSN-ANCCLI partnership. ANCCLI-IRSN seminar: Safety challenges after the Fukushima accident - September 2011

    International Nuclear Information System (INIS)

    Compagnat, Gilles; Laurent, Michel; Baumont, David; Rebour, Vincent; Quentin, Pascal

    2011-09-01

    Presented under the form of Power Point presentations, the contributions of this seminar proposed: a presentation by the ASN of the French context and of what is at stake with the complementary safety assessments of the French nuclear power plants, a report of the activities the three local information commissions of the Manche district within the context created by the Fukushima accident, and discussions by IRSN experts of the main safety challenges related to seismic hazard (discussion of the robustness of nuclear installations in this respect), to flooding (risk assessment), to a loss of cooling systems and loss of electric supplies, and to crisis management. Finally, a synthesis of the seminar is proposed

  19. Enhanced safety features of CHASHMA NPP UNIT-2 to encounter selected severe accidents, various challenges involved to prove the adequacy of severe accidents prevention/mitigation measures and to write management guidelines with one possible solution to these challenges

    International Nuclear Information System (INIS)

    Iqbal, Z.; Minhaj, A.

    2007-01-01

    This paper describes enhanced safety features of Chashma Nuclear Power Plant Unit-2 (C-2), a 325 MWe PWR to encounter selected severe accidents and discusses various challenges involved to prove the adequacy of severe accidents encountering measures and to write severe accident management guidelines (SAMGs) in compliance with the recently introduced national regulations based on the new IAEA nuclear safety standards. C-2 is being built by China National Nuclear Corporation (CNNC) for Pakistan Atomic Energy Commission (PAEC). Its twin, Unit-1 (C-1) also a 325 MWe PWR, was commissioned in 2000. Nuclear power safety with reference to severe accidents should be treated as a global issue and therefore the developed countries should include the people of developing countries in nuclear power industry's various severe accidents based research and development programs. The implementation of this idea may also deliver few other useful and mutually beneficial byproducts. (author)

  20. Independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool

    International Nuclear Information System (INIS)

    Madni, I.K.; Eltawila, F.

    1994-01-01

    MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC called ''MELCOR Verification, Benchmarking, and Applications,'' whose aim is to provide independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool. The scope of this program is to perform quality control verification on all released versions of MELCOR, to benchmark MELCOR against more mechanistic codes and experimental data from severe fuel damage tests, and to evaluate the ability of MELCOR to simulate long-term severe accident transients in commercial LWRs, by applying the code to model both BWRs and PWRs. Under this program, BNL provided input to the NRC-sponsored MELCOR Peer Review, and is currently contributing to the MELCOR Cooperative Assessment Program (MCAP). This paper presents a summary of MELCOR assessment efforts at BNL and their contribution to NRC goals with respect to MELCOR

  1. Prediction of thermal margin for external cooling of reactor vessel lower head during a severe accident

    International Nuclear Information System (INIS)

    Yoon, Ho Jun; Suh, Kune Y.

    1998-01-01

    In the TMI-2 accident, approximately nineteen (19) tons of molten core material drained into the lower plenum. One of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 .deg. C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident management strategies. As an advanced in-vessel design concept, the COrium Attak Syndrome Immunization Structures (COASIS) are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in -vessel (COASISI) and ex-vessel (COASISO) were demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the TMI-2 and the Korean Standard Nuclear Power Plant (KSNPP) reactors. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. In studying the in-vessel severe accident phenomena, one of the main goals is to verify the cooling mechanism in the reactor vessel lower plenum and thereby to prevent the vessel failure from thermal attack by the molten debris. This paper presents the first-principle calculation results for the thermal margin for the case of external cooling of the reactor vessel lower head. Adopting the method presented by F.B. Cheung, et al., we calculated the departure from nucleate boiling ratio (DNBR) for the three cases of pool boiling, flow boiling

  2. Sensitivity analysis of thermal hydraulic response in containment at core meltdown accident

    International Nuclear Information System (INIS)

    Kobayashi, Kensuke; Ishigami, Tsutomu; Horii, Hideo; Chiba, Takemi.

    1985-01-01

    A sensitivity analysis of thermal hydraulic response in a containment during a 'station blackout' (the loss of all AC power) accident at Browns Ferry unit one plant was performed with the computer code MARCH 1.0. In the analysis, the plant station batteries were assumed to be available for 4h after the initiation of the accident. The thermal hydraulic response in the containment was calculated by varying several input data for MARCH 1.0 independently and the deviation among calculated results were investigated. The sensitivity analysis showed that (a) the containment would fail due to the overtemperature without any operator actions for plant recovery, which would be strongly dependent on the model of the debris-concrete interaction and the input parameters for specifying the containment failure modes in MARCH 1.0, (b) a core melting temperature and an amount of water left in a primary system at the end of the meltdown were identified as important parameters which influenced the time of the containment failure, and (c) experimental works regarding the parameters mentioned above could be recommended. (author)

  3. Thermal Behavior of the Reactor Vessel Penetration Under External Vessel Cooling During a Severe Accident

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Park, Rae-Joon; Kim, Jong-Tae; Min, Byung-Tae; Lee, Ki-Young; Kim, Sang-Baik

    2004-01-01

    Experimental and analytical studies on the thermal behavior of reactor vessel penetration have been performed under external vessel cooling during a severe accident in the Korean next-generation reactor APR1400. Two types of tests, SUS-EXT and SUS-DRY with and without external vessel cooling, respectively, have been performed using sustained heating by an induction heater. Three tests have been carried out varying the cooling conditions at the vessel outer surface in the SUS-EXT tests. The experimental results have been thermally estimated using the LILAC computer code. The experimental results indicate that the inner surface of the vessel was ablated by the 45-mm thickness in the SUS-DRY test. Despite the total ablation of the welding material, the penetration was not ejected outside the vessel, which could be attributed to the thermal expansion of the penetration. Unlike the SUS-DRY test, the thickness of the ablation was ∼15 to 20 mm at most, so the welding was preserved in the SUS-EXT tests. It is concluded from the experimental results that the external vessel cooling highly affected the ablation configuration and the thermal behaviors of the vessel and the penetration. An increase in coolant mass flow rate from 0.047 to 0.152 kg/s had effects on the thermal behavior of the lower head vessel and penetration in the SUS-EXT tests. The LILAC analytical results on temperature distribution and ablation depth in the lower head vessel and penetration were very similar to the experimental results

  4. A Novel Thermal-Mechanical Detection System for Reactor Pressure Vessel Bottom Failure Monitoring in Severe Accidents

    International Nuclear Information System (INIS)

    Bi, Daowei; Bu, Jiangtao; Xu, Dongling

    2013-06-01

    Following the Fukushima Daiichi nuclear accident in Japan, there is an increased need of enhanced capabilities for severe accident management (SAM) program. Among others, a reliable method for detecting reactor pressure vessel (RPV) bottom failure has been evaluated as imperative by many utility owners. Though radiation and/or temperature measurement are potential solutions by tradition, there are some limitations for them to function desirably in such severe accident as that in Japan. To provide reliable information for assessment of accident progress in SAM program, in this paper we propose a novel thermal-mechanical detection system (TMDS) for RPV bottom failure monitoring in severe accidents. The main components of TMDS include thermally sensitive element, metallic cables, tension controlled switch and main control room annunciation device. With TMDS installed, there shall be a reliable means of keeping SAM decision-makers informed whether the RPV bottom has indeed failed. Such assurance definitely guarantees enhancement of severe accident management performance and significantly improve nuclear safety and thus protect the society and people. (authors)

  5. Fuel rod thermal analysis of the Angra-1 reactor during a postulated loss of coolant accident

    International Nuclear Information System (INIS)

    Praes, J.G.L.

    1982-01-01

    A thermal analysis of a fuel element is performed, as subject to the most severe cooling conditions, such as those occurring during a postulated Loss of Coolant Accident in the Angra-I reactor. Our objective was to ascertain whether the cooling of the core is assured according to 10 CRF - 50. According to the stated purpose, sensitivity analyses are necessary, using the swelling and rupture models of the cladding, and at the same time, an updating of the FLECHT heat transfer correlations in the computing program used, which is TOODEE-2 e 1 Version(28), with the purpose of adequating it to the Angra-I core analysis. In addition, we did sensitivity studies on heat transfer coefficient calculations for the steam cooling model. From the results obtained we conclude that the maximum temperature values of the cladding and the oxidation rate due to the Z sub(r) H 2 O reaction were kept well below the maximum allowable limits. Thus, the cooling of the Angra-I core is assured for the assumed accident. (Author) [pt

  6. Challenges in coupled thermal-hydraulics and neutronics simulations for LWR safety analysis

    International Nuclear Information System (INIS)

    Ivanov, Kostadin; Avramova, Maria

    2007-01-01

    The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal-hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal-hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical

  7. Lessons Learned from the Fukushima Daiichi Accident, Actions Taken and Challenges Ahead

    International Nuclear Information System (INIS)

    Shimizu, Y.

    2016-01-01

    On 19 September, 2012, the Nuclear Regulation Authority (NRA) was established in light of lessons learned from the Fukushima Daiichi accident of 11 March 2011, to ensure that such accidents never happen again, to restore public trust in regulator both in Japan and abroad and to rebuild and foster a genuine safety culture by placing the highest priority on public safety. The NRA, an independent administrative commission of the Ministry of the Environment, is organized to separate the regulatory functions from the promotional functions of the use of nuclear energy within the government, and to independently implement its duties from the perspectives of neutrality and fairness based on its expertise. Having learned the lessons from the Fukushima Daiichi accident and with reference to IAEA safety standards, since its establishment, the NRA has endeavored to strengthen the regulatory requirements, in particular, for hazards such as tsunamis and earthquakes which may lead to common cause failures, and countermeasures against severe accidents. Under the new regulatory scheme, a back-fitting system was introduced. Emergency preparedness and response measures for nuclear facilities were also enhanced. As of end of March 2016, five reactors received NRA’s permission for changing their reactor installations based on the new regulatory requirements, and two nuclear power reactors have restarted their operations. In January 2016, at the request of Japan, the IAEA sent the IRRS mission team to Japan to assess the regulatory framework for nuclear and radiation safety. Through the self-assessment prior to the mission, the NRA has developed 22 action plans, including a) improvement of regulatory inspection, b) capacity building, and c) strengthening of safety research capability. The mission team has found that Japan’s nuclear regulator has demonstrated independence and transparency since it was set up in 2012. The team also noted that the NRA needs to improve the inspection

  8. Icare/Cathare coupling: three-dimensional thermal hydraulics of severe LWR accidents

    Energy Technology Data Exchange (ETDEWEB)

    Guillard, V.; Fichot, F. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, Dept. de Recherches en Securite, DRS, 92 (France); Boudier, P.; Parent, M. [CEA Grenoble, Dir. des Reacteurs Nucleaires, DRN, 38 (France); Roser, R. [Communication et Systemes Systemes d' Information, CS SI, 38 - Fontaine (France)

    2001-07-01

    In the phenomenology of severe LWR accidents considered in safety studies, the accidental sequences can be divided into three phases: the initial phase, where no severe damage of fuel or control rods and structures occurs; the early core degradation phase, where limited material melting and relocation takes place; and the late core degradation phase during which substantial material relocation happens, molten pools and debris beds can form and corium may fall into the lower plenum and, in case of vessel failure, come into the containment. The CATHARE2 code is a system code which has been developed by CEA for IPSN, EDF and FRAMATOME to describe the thermal-hydraulics behavior of a whole PWR circuit during the first of these three phases, with a core degradation model limited to clad rupture. The ICARE2 code, developed by IPSN, allows the complete description of early and late core degradation phases, with a thermal-hydraulics model limited to the vessel, initial and boundary conditions being provided by a system code. The aim of this paper is to present the main features of the new version of the coupling, ICARE/CATHARE V2. First, the general characteristics of ICARE2 V3mod1 and CATHARE2 V1.5 standard codes, dealing with physical models and numerical aspects, are described. Second, the technical features of the coupling between the two codes are detailed. At last, some results of ICARE/CATHARE V2 calculations are presented which demonstrate the ability of the code to simulate a severe accident in a PWR and notably to describe multi-dimensional effects occurring in the core during the LOCA and degradation phases. (authors)

  9. Preprocessor for RELAP5 code, nuclear reactor thermal hydraulics accident analysis program, using Microsoft MS-EXCEL tool

    International Nuclear Information System (INIS)

    Biaty, Patricia Andrea Paladino; Sabundjian, Gaiane

    2005-01-01

    The thermal hydraulic study in accidents and transients analyses in nuclear power plants is realized with some special tools. These programs use the best estimate analyses and have been developed to simulate accidents and transients in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code has been used as tool to licensing the nuclear facilities in our country, which is the objective of this study. The main problem when RELAP5 code is used is a lot of information necessary to simulate thermal hydraulic accidents. Moreover, there is the necessity of a reasonable amount of mathematical operations to calculation of the geometry of the components existents. Therefore, in order to facilitate the manipulation of this information, it is necessary the developing a friendly preprocessor for attainment of the mathematical calculations for RELAP5 code. One of the tools used for some of these calculations is the MS-EXCEL, which will be used in this work. (author)

  10. Post-Accident Waste Management in Ukraine : Challenges and Steps Needed to Resolve the Accident Waste Problem

    International Nuclear Information System (INIS)

    Kilochytska, T.; Zinkevich, L.; Shybetskyi, I.; Krone, J.

    2016-01-01

    Conclusions: • Solving the problem of management of “Chernobyl waste” requires: - a systematic review of the existing norms and regulations with respect to best international practice of radioactive waste management; - improvement of the existing classification / characterization of radioactive waste with a focus on disposal; - improvement of the safety assessment approaches for licensing the disposal of accident waste, taking into account relevant safety features and site specific conditions; - involving of international experience and support to plan and perform safety related activity on the Shelter Object transformation

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    International Nuclear Information System (INIS)

    Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

    1997-01-01

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

  13. Thermal analyses for the spend fuel pool of Taiwan BWR plants during the loss of cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Chen, B-Y.; Yeh, C-L.; Wei, W-C.; Chen, Y-S., E-mail: onepicemine@iner.gov.tw, E-mail: clinyeh@iner.gov.tw, E-mail: hn150456@iner.gov.tw, E-mail: yschen@iner.gov.tw [Inst. of Nuclear Energy Research, Longtan Township, Taoyuan County, Taiwan (China)

    2014-07-01

    After the Fukushima nuclear accident, the safety of the spent fuel pool has become an important concern. In this study, thermal analysis of the spent fuel pool under a loss of cooling accident is performed. The BWR spent fuel pools in Taiwan are investigated, including the Chinshan, Kuosheng, and Lungmen plants. The transient pool temperature and level behaviors are calculated based on lumped energy balance. After the pool level drops below the top of the fuel, the peak cladding temperature is predicted by the Computational Fluid Dynamics (CFD) analysis. The influence to the cladding temperature of the uniform and checkboard fuel loading patterns is also investigated. (author)

  14. Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-3 silicide core

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1997-03-01

    JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl x -Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U 3 Si 2 -Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-3, there are two operation mode. One is high power operation mode up to 20 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the flow channel blockage accident, COOLOD code was used. On the other hand, steady-state thermal hydraulic analysis for both of the high power operation mode under forced convection cooling and low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-3 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-3 LEU silicide core. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    International Nuclear Information System (INIS)

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

    1997-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Trambauer, K. [GRS, Garching (Germany)

    1997-07-01

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

  18. Thermal-hydraulic analysis of loss-of-coolant accident in the JMTR

    International Nuclear Information System (INIS)

    Sakurai, Fumio; Oyamada, Rokuro

    1985-02-01

    The reevaluation of the Loss-of-Coolant Accident (LOCA) was required through the process of a safety review for the Japan Materials Testing Reactor (JMTR) core conversion from the high-enriched uranium fuel (Enrichment : 93%) to the medium-enriched uranium fuel (Enrichment : 45%). The following were concluded by thermal-hydraulic analysis of a LOCA caused by a double-ended pipe break in the JMTR primary cooling system. (1) The fuel in the core does not burn-out as long as it is covered with water. (2) A larger siphon break valve (larger than phi60mm) should be installed instead of the present one (phi25mm) on the primary cooling system in order to prevent the core from being uncovered with water in case of a LOCA caused by a double-ended pipe break. The present siphon break valve was installed to keep the core covered with water in case of a LOCA caused by a small pipe rupture. In this analysis, the Siphon Breaker Analysis Code (SBAC) was written in order to analyse the size of the siphon break valve and its accuracy was confirmed to be within 5% through a verification experiment. (author)

  19. Development and qualification of a thermal-hydraulic nodalization for modeling station blackout accident in PSB-VVER test facility

    Energy Technology Data Exchange (ETDEWEB)

    Saghafi, Mahdi [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); Ghofrani, Mohammad Bagher, E-mail: ghofrani@sharif.edu [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); D’Auria, Francesco [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, San Piero a Grado, Pisa (Italy)

    2016-07-15

    Highlights: • A thermal-hydraulic nodalization for PSB-VVER test facility has been developed. • Station blackout accident is modeled with the developed nodalization in MELCOR code. • The developed nodalization is qualified at both steady state and transient levels. • MELCOR predictions are qualitatively and quantitatively in acceptable range. • Fast Fourier Transform Base Method is used to quantify accuracy of code predictions. - Abstract: This paper deals with the development of a qualified thermal-hydraulic nodalization for modeling Station Black-Out (SBO) accident in PSB-VVER Integral Test Facility (ITF). This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr nuclear power plant. In this regard, a nodalization has been developed for thermal-hydraulic modeling of the PSB-VVER ITF by MELCOR integrated code. The nodalization is qualitatively and quantitatively qualified at both steady-state and transient levels. The accuracy of the MELCOR predictions is quantified in the transient level using the Fast Fourier Transform Base Method (FFTBM). FFTBM provides an integral representation for quantification of the code accuracy in the frequency domain. It was observed that MELCOR predictions are qualitatively and quantitatively in the acceptable range. In addition, the influence of different nodalizations on MELCOR predictions was evaluated and quantified using FFTBM by developing 8 sensitivity cases with different numbers of control volumes and heat structures in the core region and steam generator U-tubes. The most appropriate case, which provided results with minimum deviations from the experimental data, was then considered as the qualified nodalization for analysis of SBO accident in the PSB-VVER ITF. This qualified nodalization can be used for modeling of VVER-1000 nuclear power plants when performing SBO accident analysis by MELCOR code.

  20. Thermal Load Analysis of Multilayered Corium in the Lower Head of Reactor Pressure Vessel during Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Whang, Seok Won; Park, Hyun Sun [POSTECH, Pohang (Korea, Republic of); Hwang, Tae Suk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    In-Vessel Retention (IVR) is one of the severe accident management strategies to terminate or mitigate the severe accident which is also called 'core-melt accident'. The reactor vessel would be cooled by flooding the cavity with water. The molten core mixture is divided into two or three layers due to the density difference. Light metal layer which contains Fe and Zr is on the oxide layer which is consist of UO{sub 2} and ZrO{sub 2}. Heavy metal layer which contains U, Fe and Zr is located under the oxide layer. In oxide layer, the crust which is solidified material is formed along the boundary. The assessment of IVR for nuclear power plant has been conducted with lumped parameter method by Theofanous, Rempe and Esmaili. In this paper, the numerical analysis was performed and verified with the Esmaili's work to analyze thermal load of multilayered corium in pressurized reactor vessel and also to examine the condition of in-vessel corium characteristic before the vessel failure that lead to ex-vessel severe accident progression for example, ex-vessel debris bed cooling. The in-vessel coolability analysis for several scenarios is conducted for the plant which has higher power than AP1000. Two sensitivity analyses are conducted, the first is emissivity of light metal layer and the second is the heat transfer coefficient correlations of oxide layer. The effect of three layered system also investigated. In this paper, the numerical analysis was performed and verified with Esmaili's model to analyze thermal load of multilayered corium in pressurized reactor vessel. For two layered system, thermal load was analyzed according to the severe accident scenarios, emissivity of the light metal layer and heat transfer correlations of the.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  2. Thermal Analysis Of A 9975 Package In A Facility Fire Accident

    International Nuclear Information System (INIS)

    Gupta, N.

    2011-01-01

    Surplus plutonium bearing materials in the U.S. Department of Energy (DOE) complex are stored in the 3013 containers that are designed to meet the requirements of the DOE standard DOE-STD-3013. The 3013 containers are in turn packaged inside 9975 packages that are designed to meet the NRC 10 CFR Part 71 regulatory requirements for transporting the Type B fissile materials across the DOE complex. The design requirements for the hypothetical accident conditions (HAC) involving a fire are given in 10 CFR 71.73. The 9975 packages are stored at the DOE Savannah River Site in the K-Area Material Storage (KAMS) facility for long term of up to 50 years. The design requirements for safe storage in KAMS facility containing multiple sources of combustible materials are far more challenging than the HAC requirements in 10 CFR 71.73. While the 10 CFR 71.73 postulates an HAC fire of 1475 F and 30 minutes duration, the facility fire calls for a fire of 1500 F and 86 duration. This paper describes a methodology and the analysis results that meet the design limits of the 9975 component and demonstrate the robustness of the 9975 package.

  3. Simplified methods applied to the complete thermal and mechanical behaviour of a pressure vessel during a severe accident

    International Nuclear Information System (INIS)

    Dupas, P.

    1996-01-01

    EDF has developed a software package of simplified methods (proprietary ones from literature) in order to study the thermal and mechanical behaviour of a PWR pressure vessel during a severe accident involving a corium localization in the vessel lower head. Using a part of this package, we can evaluate for instance successively: the heat flux at the inner surface of the vessel (conductive or convective pool of corium); the thermal exchange coefficient between the vessel and the outside (dry pit or flooded pit, watertight thermal insulation or not); the complete thermal evolution of the vessel (temperature profile, melting); the possible global plastic failure of the vessel; the creep behaviour in the vessel. These simplified methods are low cost alternative to finite element calculations which are yet used to validate the previous methods, waiting for experimental results to come. (authors)

  4. Simplified methods to the complete thermal and mechanical behavior of a pressure vessel during a severe accident

    International Nuclear Information System (INIS)

    Dupas, P.; Schneiter, J.R.

    1996-01-01

    EDF has developed a software package of simplified methods (proprietary ones or from literature) in order to study the thermal and mechanical behavior of a PWR pressure vessel during a severe accident involving a corium localization in the vessel lower head. Using a part of this package, the authors can evaluate for instance successively: the heat flux at the inner surface of the vessel (conductive or convective pool of corium); the thermal exchange coefficient between the vessel and the outside (dry pit or flooded pit, watertight thermal insulation or not); the complete thermal evolution of the vessel (temperature profile, melting); the possible global plastic failure of the vessel; the creep behavior in the thickness of the vessel. These simplified methods are a cost effective alternative to finite element calculations which are yet used to validate the previous methods, waiting for experimental results to come

  5. Challenging the immediate causes: A work accident investigation in an oil refinery using organizational analysis.

    Science.gov (United States)

    Beltran, Sandra Lorena; Vilela, Rodolfo Andrade de Gouveia; de Almeida, Ildeberto Muniz

    2018-01-01

    In many companies, investigations of accidents still blame the victims without exploring deeper causes. Those investigations are reactive and have no learning potential. This paper aims to debate the historical organizational aspects of a company whose policy was incubating an accident. The empirical data are analyzed as part of a qualitative study of an accident that occurred in an oil refinery in Brazil in 2014. To investigate and analyse this case we used one-to-one and group interviews, participant observation, Collective Analyses of Work and a documentary review. The analysis was conducted on the basis of concepts of the Organizational Analysis of the event and the Model for Analysis and Prevention of Work Accidents. The accident had its origin in the interaction of social and organizational factors, among them being: excessively standardized culture, management tools and outcome indicators that give a false sense of safety, the decision to speed up the project, the change of operator to facilitate this outcome and performance management that encourages getting around the usual barriers. The superficial accident analysis conducted by the company that ignored human and organizational factors reinforces the traditional safety culture and favors the occurrence of new accidents.

  6. Preliminary thermal design of a pressurized water reactor containment for handling severe accident consequences

    International Nuclear Information System (INIS)

    Abdullah, A.M.; Karameldin, A.

    1998-01-01

    A one-dimensional mathematical model has been developed for a 4250 MW(th) Advanced Pressurized Water Reactor containment analysis following a severe accident. The cooling process of the composite containment-steel shell and concrete shield- is achievable by natural circulation of atmospheric air. However, for purpose of gettering higher degrees of safety margin, the present study undertakes two objectives: (1) Installment of a diesel engine-driven air blower to force air through the annular space between the steel shell and concrete shield. The engine can be remotely operated to be effective in case of station blackout. (ii) Fixing longitudinally plate fins on the circumference of the inside and outside containment steel shell. These fins increase the heat transfer areas and hence the rate of heat removal from the containment atmosphere. In view of its importance - from the safety viewpoint - the long term behaviour of the containment which is a quasi-steady state problem, is formulated through a system of coupled nonlinear algebraic equations which describe the thermal-hydraulic and thermodynamic behaviour of the double shell containment. The calculated results revealed the following: (i) the passively air cooled containment can remove maximum heat load of 11.5 MW without failure, (ii) the effect of finned surface in the air passage tends to decrease the containment pressure by 20 to 30%, depending on the heat load, (iii) the effect of condensing fins is negligible for the proposed fin dimensions and material. However, by reducing the fin width, increasing their thickness, doubling their number, and using a higher conductive metal than the steel, it is expected that the containment pressure can be further reduced by 10% or more, (iv) the fins' dimensions and their number must be optimized via maximizing the difference or the ratio between the heat removed and pressure drop to get maximum heat flow rate

  7. FOREVER Experiments on Thermal and Mechanical Behavior of a Reactor Pressure Vessel during a Severe Accident

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Nourgaliev, R.R.; Dinh, T.N.; Karbojian, A.; Green, J.A.; Bui, V.A.

    1999-01-01

    This paper describes the FOREVER (Failure Of Reactor Vessel Retention) experimental program, which is currently underway at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS). The objectives of the FOREVER experiments are to obtain data and develop validated models (i) on the melt coolability process inside the vessel, in the presence of water (in particular, on the efficacy of the postulated gap cooling to preclude vessel failure); and (ii) on the lower head failure due to the creep process in the absence of water inside and/or outside the lower head. The facility employs 1/10.-scale carbon steel vessels of 0.4 m diameter, 15 mm thickness and 600 mm height. Up to 20 liters of binary-oxide melts with 100-300 K superheat are employed, as a simulant for the prototypic corium melt, and internal heating is provided by electrical heaters of up to 20 kW power in order to maintain the vessel wall temperatures at 1100-1200 K. Auxiliary systems are designed to provide an overpressure up to 4 MPa in the test vessel. Thus, severe accident scenarios with RCS depressurization are modeled. Creep behavior of the three-dimensional vessel, formation of the gap between the melt pool crust and the creeping vessel, and mechanisms of the gap cooling by water ingression will be the subjects of study and measurements in the FOREVER experimental program. Scaling rationale as well as pre-test analyses of the thermal and mechanical behavior of the FOREVER test vessels are presented. (authors)

  8. The post-accident nuclear issue: the new crisis expertise challenges for the IRSN

    International Nuclear Information System (INIS)

    Champion, D.

    2010-01-01

    The author reports the work performed by two work groups conducted by the IRSN (the French Radioprotection and Nuclear Safety Institute), the first one on the issue of assessment of radiological and dosimetric consequences in a post-accident situation, and the second one on hypotheses to be used to perform predictive assessments of these consequences. First dealing with the end of the emergency phase, he describes how to anticipate actions of protection against immediate post-accident consequences: orientation of the expertise strategy based on the CODIRPA's doctrine, post-accident zoning based on predictive indicators, use of reasonably prudent hypotheses for the first predictive assessments, importance of initial radioactive deposits to perform predictive assessments. Then, the author presents an iterative method of assessment of post-accident consequences: organization of environment radioactivity measurement programmes, periodic update of mapping of initial deposit and of actual deposits at a given time

  9. Thermal plasma spraying for SOFCs: Applications, potential advantages, and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Hui, Rob; Wang, Zhenwei; Jankovic, Jasna; Yick, Sing; Maric, Radenka; Ghosh, Dave [National Research Council Institute for Fuel Cell Innovation, 4250 Wesbrook Mall, Vancouver, BC V6T 1W5 (Canada); Kesler, Olivera [National Research Council Institute for Fuel Cell Innovation, 4250 Wesbrook Mall, Vancouver, BC V6T 1W5 (Canada); Department of Mechanical Engineering, University of British Columbia, 2054-6250 Applied Science Lane, Vancouver, BC V6T 1Z4 (Canada); Rose, Lars [National Research Council Institute for Fuel Cell Innovation, 4250 Wesbrook Mall, Vancouver, BC V6T 1W5 (Canada); Department of Materials Engineering, University of British Columbia, 309-6350 Stores Road, Vancouver, BC V6T 1Z4 (Canada)

    2007-07-10

    In this article, the applications, potential advantages, and challenges of thermal plasma spray (PS) processing for nanopowder production and cell fabrication of solid oxide fuel cells (SOFCs) are reviewed. PS processing creates sufficiently high temperatures to melt all materials fed into the plasma. The heated material can either be quenched into oxide powders or deposited as coatings. This technique has been applied to directly deposit functional layers as well as nanopowder for SOFCs application. In particularly, low melting point and highly active electrodes can be directly fabricated on zirconia-based electrolytes. This is a simple processing technique that does not require the use of organic solvents, offering the opportunity for flexible adjustment of process parameters, and significant time saving in production of the cell and cost reduction compared with tape casting, screen printing and sintering processing steps. Most importantly, PS processing shows strong potential to enable the deposition of metal-supported SOFCs through the integrated fabrication of membrane-electrode assemblies (MEA) on porous metallic substrates with consecutive deposition steps. On the other hand, the application of PS processing to produce SOFCs faces some challenges, such as insufficient porosity of the electrodes, the difficulty of obtaining a thin (<10 {mu}m) and dense electrolyte layer. Fed with H{sub 2} as the fuel gas and oxygen as the oxidant gas, the plasma sprayed cell reached high power densities of 770 mW cm{sup -2} at 900 C and 430 mW cm{sup -2} at 800 C at a cell voltage of 0.7 V. (author)

  10. Development of Innovative Accident Tolerant High Thermal Conductivity UO2-Diamond Composite Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    Tulenko, James [Univ. of Florida, Gainesville, FL (United States); Subhash, Ghatu [Univ. of Florida, Gainesville, FL (United States)

    2016-01-01

    The University of Florida (UF) evaluated a composite fuel consisting of UO2 powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO2-Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO2 – SiC and UO2 – Carbon Nanotube fuel pins. UF is proving with the current research results that the addition of diamond micro particles to UO2 may greatly enhanced the thermal conductivity of the UO2 pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.

  11. Integral thermal model of severe accident dynamics of NPP with containment

    International Nuclear Information System (INIS)

    Arutyunyan, R.V.; Bol'shov, L.A.; Vasil'ev, A.D.; Kamennov, G.P.

    1991-01-01

    An analytical model of the interaction of reactor core remains with concrete during severe accidents at nuclear power plants is considered. Time dependences of side and radial concrete melting are plotted. Time dependences of containment atmosphere temperature and pressure during a severe accident at nuclear power plants are investigated analytically and numerically. The sensitivity of the results to the coefficient values in the problem is studied within the range of their concertainty. The Kaverna-1 is described. The results of modelling a severe NPP accident which have been obtained using the Kaverna-1 package are presented

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S.

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Rogers, J.T.

    1984-06-01

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

  15. Design features which mitigate severe accident challenges in the GE ABWR and SBWR

    International Nuclear Information System (INIS)

    Buchholz, Carol E.

    2004-01-01

    A reduction of the requirements for the emergency planning zone (EPZ) is a goal of advanced light water reactors. The technical basis for reducing the EPZ requirements is based on a very low frequency of a severe accident and high confidence that the offsite dose would be low even if a severe accident was to occur. Design features have been included in both the ABWR and SBWR to ensure that both of these goals are achieved. Probabilistic Risk Assessments (PRAs) have been performed for both plants. The PRAs indicate a core damage frequency on the order of IE-7 for both plants. The PRAs also show that the containments will not fail even if a severe accident should occur. The potential offsite is extremely low. (author)

  16. 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

  17. Coupling of 3-D core computational codes and a reactor simulation software for the computation of PWR reactivity accidents induced by thermal-hydraulic transients

    International Nuclear Information System (INIS)

    Raymond, P.; Caruge, D.; Paik, H.J.

    1994-01-01

    The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs

  18. A study of thermal stratification in the cold legs during the subcooled blowdown phase of a loss of coolant accident in the OSU APEX thermal hydraulic testing facility

    International Nuclear Information System (INIS)

    Wachs, D. M.

    1998-01-01

    Thermal stratification, which has been linked to the occurrence of pressurized thermal shock (PTS), is observed to occur during the early stages of simulated loss of coolant accidents (LOCAS) in the Oregon State University Advanced Plant Experiment (OSU APEX) Thermal Hydraulic Test Facility. The OSU APEX Test Facility is a scaled model of the Westinghouse AP600 nuclear power plant. Analysis of the OSU APEX facility data has allowed the determination of an onset criteria for thermal stratification and has provided support for the postulated mechanisms leading to thermal stratification. CFX 4.1, a computational fluid dynamics code, was used to generate a model of the cold legs and the downcomer that described the phenomena occurring within them. Some mixing phenomena were predicted that lead to non-uniformity between the two cold legs attached to the steam generator on the side of the facility containing the Passive Residual Heat Removal (PRHR) injection system. The stratification was found to be two phase and unlikely to be a factor in PTS

  19. A review of the current thermal-hydraulic modeling of the Jules Horowitz Reactor: A loss of flow accident analysis

    International Nuclear Information System (INIS)

    Pegonen, R.; Bourdon, S.; Gonnier, C.; Anglart, H.

    2014-01-01

    Highlights: • CEA methodology for thermal-hydraulic calculations in the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during LOFA using CATHARE and FLICA4. • Safety criteria, important modeling parameters and correlations are presented. • Possible improvements of the current methodology are discussed and proposed. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support existing and future nuclear reactor designs. The reactor is under construction at CEA Cadarache research center in France and is expected to start operation at the end of this decade. R and D and analytical works have already been performed to set-up the methodology for thermal-hydraulic calculations of the reactor. This paper presents the off-line coupled thermal-hydraulic modeling of the reactor using the CATHARE system code and the FLICA4 core analysis code. The main objective of the present work is to analyze the thermal-hydraulic calculations of the reactor during the loss of flow accident using CEA methodology. Possible improvements of the current methodology are shortly discussed and suggested

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

    Science.gov (United States)

    Kumagai, Daisuke; Ohsaki, Hiroyuki; Tomita, Masaru

    2016-12-01

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

  1. IRSN-ANCCLI partnership. IRSN-ANCCLI seminar - Safety challenges after the Fukushima accident - January 2012

    International Nuclear Information System (INIS)

    Compagnat, Gilles; Revol, H.; Rousselet, Yannick; Sene, Monique; Lheureux, Yves; Laurent, Michel; Lavarenne, Caroline; Jorel, M.; Houdre, Thomas; Lachaume, Jean-Luc

    2012-01-01

    After a synthesis, this document contains the contributions (Power Point presentations) of a seminar which addressed the following topics: remarks by the HCTISN on the process of complementary safety assessments, analysis and discussion by the GSIEN on reports of complementary assessment of safety of nuclear installations with respect to the Fukushima accident, opinion of the Gravelines local information commission (CLI) on the complementary safety assessment report for the Gravelines nuclear power plant, stage point of the Manche INTERCLI work-group on the safety of nuclear installations after Fukushima, presentation by the IRSN of the complementary safety assessments, and opinion of the ASN on complementary safety assessments (ECS) of priority nuclear installations

  2. Issues and challenges for pedestrian active safety systems based on real world accidents.

    Science.gov (United States)

    Hamdane, Hédi; Serre, Thierry; Masson, Catherine; Anderson, Robert

    2015-09-01

    The purpose of this study was to analyze real crashes involving pedestrians in order to evaluate the potential effectiveness of autonomous emergency braking systems (AEB) in pedestrian protection. A sample of 100 real accident cases were reconstructed providing a comprehensive set of data describing the interaction between the vehicle, the environment and the pedestrian all along the scenario of the accident. A generic AEB system based on a camera sensor for pedestrian detection was modeled in order to identify the functionality of its different attributes in the timeline of each crash scenario. These attributes were assessed to determine their impact on pedestrian safety. The influence of the detection and the activation of the AEB system were explored by varying the field of view (FOV) of the sensor and the level of deceleration. A FOV of 35° was estimated to be required to detect and react to the majority of crash scenarios. For the reaction of a system (from hazard detection to triggering the brakes), between 0.5 and 1s appears necessary. Copyright © 2015 Elsevier Ltd. All rights reserved.

  3. Safety challenges after the Fukushima accident. ANCCLI-IRSN work meeting

    International Nuclear Information System (INIS)

    Demet, Michel; Lheureux, Yves; Sene, Monique; Sene, Raymond; Eimer, Michel; Lachaume, Jean-Luc; Majnoni, Sophia; Marignac, Yves; Revol, Henri; Gilles, Compagnat; Baumont, David; Huet, Cyril; Rebour, Vincent; Besnus, Francois; Le Bars, Igor; Lizot, Marie-Therese; Carre, Christine; Dupuy, Patricia; Jorel, Martial; Lavarenne, Caroline; Rousseau, Jean-Marie; Charron, Sylvie; Gilli, Ludivine

    2011-11-01

    After a synthetic report of the meeting, this document contains Power Point presentations proposed by the different contributors. These presentations proposed: an overview on additional safety assessments (ECS) and an assessment of 'post-Fukushima' inspections performed in basic nuclear installations; the CLI's opinion on the Gravelines ECS report; an analysis and a discussion of ECS reports of nuclear installations in the perspective of the Fukushima accident; the IRSN analysis of ECS as they are performed by operators; a presentation of the IRSN analysis approach to ECS; contributions of different post-Fukushima ECS permanent groups within the IRSN (these work groups address installation condition, external flooding, and seismic risk); a presentation of the 'fusion prevention' aspects of the management of accidental situations in EDF reactors

  4. Electronic Paramagnetic Resonance of irradiated nails: challenges for a dosimetry in radiation accidents

    International Nuclear Information System (INIS)

    Giannoni, Ricardo A.; Rodrigues Junior, Orlando

    2014-01-01

    The purpose of this work is to characterize samples of human nails exposed to high doses of radiation, applying the technique of Electron Paramagnetic Resonance (EPR). The objective is to establish a dose response study that allow determine the absorbed dose by exposed individuals in situations of radiological accidents, in a retrospective form. Samples of human nails were collected and afterward irradiated with gamma radiation, and received dose of 20 Gy. The EPR measurement performed on the samples, before irradiation, permitted the signal identification of the components associated with effects caused by the mechanical stress during the fingernail cutting, the so-called mechanically induced signal (MIS). After the irradiation, different species of free radicals were identified, the so-called radiation induced signal (RIS). (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  7. Challenges in thermal and hydraulic analysis of ADS target systems

    International Nuclear Information System (INIS)

    Groetzbach, G.; Batta, A.; Lefhalm, C.-H.; Otic, I.

    2004-01-01

    The liquid metal cooled spallation targets of Accelerator Driven nuclear reactor Systems obey high thermal loads; in addition some flow and cooling conditions are of a prototypical character; in contrast the operating conditions for the engaged materials are narrow; thus, the target development requires a very careful analysis by experimental and numerical means. Especially the cooling of the steel window, which is heated by the proton beam, needs special care. Some of the main goals of the experimental and numerical analyses of the thermal dynamics of those systems are discusses. The prediction of locally detached flows and of flows with larger recirculation areas suffers from insufficient turbulence modeling; this has to be compensated by using prototypical model experiments, e.g. with water, to select the adequate models and numerical schemes. The well known problems with the Reynolds analogy in predicting the heat transfer in liquid metals requires always prototypic liquid metal experiments to select and adapt the turbulent heat flux models. The uncertainties in liquid metal experiments cannot be neglected; so it is necessary to perform CFD calculations and experiments always hand in hand and to develop improve turbulent heat flux models. One contribution to an improved 3 or 4-equation model is deduced from recent Direct Numerical Simulation (DNS) data. (author)

  8. Safety challenges after the Fukushima accident for operated installations others than EDF reactors

    International Nuclear Information System (INIS)

    Sene, Monique; Rollinger, Francois; Lheureux, Yves; Lizot, Marie-Therese; Kerdelhue, M.; Py, M.E.; Leroyer, Veronique; Pultier, Marc; Kassiotis, Christophe; Chambrette, Pierre; Devaux, Pascal; Baron, Yves; Collinet, Jacques

    2013-12-01

    This document contains Power Point presentations which, within the perspective created by the Fukushima accident, address various aspects of safety issues for installations other than currently operated EDF reactors. These contributions propose: an agenda of additional safety assessments (ECS) performed on these installations and an examination of responses made to prescriptions made on the 16 June 2012; a presentation by the IRSN of ECS performed in Areva plants; a presentation by Areva of arrangements related to these ECS; a presentation of the Manche local information commissions (CLI) and a presentation of their approach according to a white paper for the safety of civil nuclear installations located in the Manche department; a presentation by the IRSN on ECS concerning various basic nuclear installations such as laboratories, experimental reactors and stopped reactors; a presentation by the CEA of ECS of its installations (context, approach, execution and conclusions); a presentation by the ANCCLI about ASN decision and decision projects about the hard core according to ECS (example of the High flux reactor in the ILL in Grenoble)

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

    International Nuclear Information System (INIS)

    Rogers, J.T.

    1984-06-01

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

  10. Steering committee for the management of the post-accidental phase of a nuclear accident or of a radiological situation (CODIRPA) - Work-group nr 4. Response to health challenges after a radiological accident - Final report March 2011

    International Nuclear Information System (INIS)

    2011-03-01

    The first part of this report presents the context of preparation to the response to a radiological accident in France. It proposes a synthetic presentation of scenarios, of the different accident phases, of management principles based on areas and stakes as they are presented in the emergency phase exit guide. It also indicates public health challenges related to the different studied scenarios. The second part proposes a chronological synthesis of actions to be undertaken after an accident in order to face public health stakes. The third part proposes a detailed presentation of the implementation and sequence of actions to be undertaken depending on the studied scenarios: medical and psychological care, census, health risk assessment, health information

  11. Use of thermal time constant concept in the analysis of reactivity induced accidents with feedback

    International Nuclear Information System (INIS)

    Narain, R.

    1981-01-01

    A simple heat transfer model based on the thermal time constant concept which leads to significant reduction in fuel temperature computing time and gives a physical insight of the phenomena is presented. The fuel temperatures can be used to estimate the reactivity feedback using the measured or calculated Doppler coefficients. (E.G.) [pt

  12. Thermal hydraulic And RSG-Gas Core Reactivity Characteristics Due To Cold Water Insertion Accident

    International Nuclear Information System (INIS)

    Hastuti, Endiah Puji; Suparlina, Lily; Tukiran

    2000-01-01

    Under normal operating condition,the primary coolant is circulated by 2 out of the 3 primary coolant pumps. Unnecessary operation of the reserve pump would result in a temperatur decrease of the primary coolant by less than 5 o C. the corresponding increase of reactivity amounts to Δρ ≤0,1 %. The analysis was done using silicide core configuration data with 3.55 gU /cm 3 fuel loading. The calculation model was done with and without automatic control rod. The calculation results for the worst case condition, shows that reactor reached the maximum power 28.52 MW at 81.1 seconds, after the accident occurred. The maximal fuel element, cladding and outlet coolant temperatures are 148.3 o C,142.1 o C, and 75.7 o C, respectively. Safety margins for DNBR and flow instability reached 1.25 and 4.20, respectively. Comparing to the RSG-GAS safety margin at transient condition reguirement >1.48, RSG-GAS has enough safety margin if the power trip executed at 114% of 25 MW

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  14. Thermal Response of the 21-PWR Waste Package to a Fire Accident

    International Nuclear Information System (INIS)

    F.P. Faucher; H. Marr; M.J. Anderson

    2000-01-01

    The objective of this calculation is to evaluate the thermal response of the 21-PWR WP (pressurized water reactor waste package) to the regulatory fire event. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation (Attachment IV) is that of the potential design of the type of waste package considered in this calculation. The procedure AP-3.12Q.Calculations (Reference 1), and the Development Plan (Reference 24) are used to develop this calculation

  15. Security surveillance challenges and proven thermal imaging capabilities in real-world applications

    Science.gov (United States)

    Francisco, Glen L.; Roberts, Sharon

    2004-09-01

    Uncooled thermal imaging was first introduced to the public in early 1980's by Raytheon (legacy Texas Instruments Defense Segment Electronics Group) as a solution for military applications. Since the introduction of this technology, Raytheon has remained the leader in this market as well as introduced commercial versions of thermal imaging products specifically designed for security, law enforcement, fire fighting, automotive and industrial uses. Today, low cost thermal imaging for commercial use in security applications is a reality. Organizations of all types have begun to understand the advantages of using thermal imaging as a means to solve common surveillance problems where other popular technologies fall short. Thermal imaging has proven to be a successful solution for common security needs such as: ¸ vision at night where lighting is undesired and 24x7 surveillance is needed ¸ surveillance over waterways, lakes and ports where water and lighting options are impractical ¸ surveillance through challenging weather conditions where other technologies will be challenged by atmospheric particulates ¸ low maintenance requirements due to remote or difficult locations ¸ low cost over life of product Thermal imaging is now a common addition to the integrated security package. Companies are relying on thermal imaging for specific applications where no other technology can perform.

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

    International Nuclear Information System (INIS)

    Blankenship, R.W.

    1980-05-01

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

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

    International Nuclear Information System (INIS)

    Blankenship, R.W.

    1980-05-01

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

  18. 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.)

  19. Outcomes from the EURATOM–ROSATOM ERCOSAM SAMARA projects on containment thermal-hydraulics for severe accident management

    Energy Technology Data Exchange (ETDEWEB)

    Paladino, Domenico, E-mail: domenico.paladino@psi.ch [Paul Scherrer Institut (Switzerland); Andreani, Michele [Paul Scherrer Institut (Switzerland); Guentay, Salih [Innovative, Technology Development GmbH (Switzerland); Mignot, Guillaume; Kapulla, Ralf; Paranjape, Sidharth; Sharabi, Medhat [Paul Scherrer Institut (Switzerland); Kisselev, Arkadi; Yudina, Tatiana; Filippov, Aleksandr [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow 115191 (Russian Federation); Kamnev, Mikhail; Khizbullin, Akhmir; Tyurikov, Oleg [JSC “Afrikantov OKB Mechanical Engineering”, Nizhny Novgorod 603074 (Russian Federation); Liang, Zhe [CNL-2251 Speakman Drive, Mississauga, ON L5K 1B2 (Canada); Abdo, Daniele; Brinster, Jérôme; Dabbene, Frédéric [CEA, DEN, DM2S, STMF, F-91191 Gif-sur-Yvette Cedex (France); Kelm, Stephan [Forschungszentrum Juelich, 52425 Jülich (Germany); Klauck, Michael; Götz, Lasse [RWTH Aachen University (Germany); and others

    2016-11-15

    Highlights: • Hydrogen distribution in the containment of PWR was investigated for scenario leading to stratification. • The scenario was scaled from a generic PWR containment to four facilities. • Effect of spray, cooler and heat sources was investigated experimentally and with LP and CFD. • Code-to-code benchmarks aiming a scaling up the facilities to a large containment. - Abstract: ERCOSAM and SAMARA are the acronyms for two parallel projects co-financed respectively by EURATOM and ROSATOM during the period 2010–2014 with the general aim to advance the knowledge on the phenomenology associated with the hydrogen and steam spreading and stratification in the LWR containment during a postulated severe accident. The important peculiarity of the projects was in experimental and analytical investigating the impact of systems such as spray, cooler and heat sources (simulating thermal effect of PARs) on the distribution of gas mixture (e.g. hydrogen, steam, air). This paper presents the main outcomes of the ERCOSAM–SAMARA projects.

  20. Outcomes from the EURATOM–ROSATOM ERCOSAM SAMARA projects on containment thermal-hydraulics for severe accident management

    International Nuclear Information System (INIS)

    Paladino, Domenico; Andreani, Michele; Guentay, Salih; Mignot, Guillaume; Kapulla, Ralf; Paranjape, Sidharth; Sharabi, Medhat; Kisselev, Arkadi; Yudina, Tatiana; Filippov, Aleksandr; Kamnev, Mikhail; Khizbullin, Akhmir; Tyurikov, Oleg; Liang, Zhe; Abdo, Daniele; Brinster, Jérôme; Dabbene, Frédéric; Kelm, Stephan; Klauck, Michael; Götz, Lasse

    2016-01-01

    Highlights: • Hydrogen distribution in the containment of PWR was investigated for scenario leading to stratification. • The scenario was scaled from a generic PWR containment to four facilities. • Effect of spray, cooler and heat sources was investigated experimentally and with LP and CFD. • Code-to-code benchmarks aiming a scaling up the facilities to a large containment. - Abstract: ERCOSAM and SAMARA are the acronyms for two parallel projects co-financed respectively by EURATOM and ROSATOM during the period 2010–2014 with the general aim to advance the knowledge on the phenomenology associated with the hydrogen and steam spreading and stratification in the LWR containment during a postulated severe accident. The important peculiarity of the projects was in experimental and analytical investigating the impact of systems such as spray, cooler and heat sources (simulating thermal effect of PARs) on the distribution of gas mixture (e.g. hydrogen, steam, air). This paper presents the main outcomes of the ERCOSAM–SAMARA projects.

  1. ATHENA2D, Simulation Hypothetical Recriticality Accident in a Thermal Neutron Spectrum

    International Nuclear Information System (INIS)

    1999-01-01

    1 - Description of program or function: ATHENA 2 D was written to simulate a hypothetical water reflood of a highly-damaged light water reactor (such as the Three-Mile-Island Unit-2 after meltdown, with a packed debris bed near the center of the core), but with insufficiently-borated reflood water. A recriticality transient may result because of the potentially more reactive debris bed. ATHENA-2D solves the transient multigroup neutron diffusion equations in (r,z) geometry. Executing in parallel with the transient neutronics, is a single-phase computational fluid dynamics (CFD) model, driven by multichannel thermal hydraulics based on detailed pin models. Numerous PV-Wave procedure files are included on the distribution media, useful for those who already have PV-Wave from Visual Numerics. These procedures are documented in the 'README' files included on the distribution CD. Some reactor lattice computer code such as WIMS-E, CCC-576/WIMSD4, or CCC-656/WIMSD5B is required for the creation of macroscopic cross section libraries, given pin-cell geometries. WIMS-E is a commercial product available from AEA Technologies, England, WIMS is not included on the ATHENA 2 D distribution CD. Several auxiliary routines are included in the package. TFMAX: Utility that searches through ATHENA 2 D binary output to find the maximum fuel temperature over space and time. POST V EL: Utility that searches through ATHENA 2 D binary output to find maximum scalar and flow field values (over space) and outputs normalization factors as a function of time. These results are used to correctly scale animations. CONVT: If executing ATHENA 2 D on a PC under Windows, this utility converts one form of binary output (directly from ATHENA 2 D) to another, which is readable by PV-Wave for Windows (PV-Wave is data animation and visualization software from Visual Numerics, Inc.) CALC M TX: Post-processing utility for calculating the model coefficients for the calculation matrix. 2 - Methods: Both the

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

    Directory of Open Access Journals (Sweden)

    Itamar Iliuk

    2016-01-01

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

  3. Highlights from the literature on accident causation and system safety: Review of major ideas, recent contributions, and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Saleh, J.H., E-mail: jsaleh@gatech.ed [School of Aerospace Engineering, Georgia Institute of Technology (United States); Marais, K.B. [School of Aeronautics and Astronautics, Purdue University (United States); Bakolas, E.; Cowlagi, R.V. [School of Aerospace Engineering, Georgia Institute of Technology (United States)

    2010-11-15

    This work constitutes a short guide to the extensive but fragmented literature on accident causation and system safety. After briefly motivating the interest in accident causation and discussing the notion of a safety value chain, we delve into our multi-disciplinary review with discussions of Man Made Disasters, Normal Accident, and the High Reliability Organizations (HRO) paradigm. The HRO literature intersects an extensive literature on safety culture, a subject we then briefly touch upon. Following this discussion, we note that while these social and organizational contributions have significantly enriched our understanding of accident causation and system safety, they have important deficiencies and are lacking in their understanding of technical and design drivers of system safety and accident causation. These missing ingredients, we argue, were provided in part by the development of Probabilistic Risk Assessment (PRA). The idea of anticipating possible accident scenarios, based on the system design and configuration, as well as its technical and operational characteristics, constitutes an important contribution of PRA, which builds on and extends earlier contributions made by the development of Fault Tree and Event Tree Analysis. We follow the discussion of PRA with an exposition of the concept of safety barriers and the principle of defense-in-depth, both of which emphasize the functions and 'safety elements [that should be] deliberately inserted' along potential accident trajectories to prevent, contain, or mitigate accidents. Finally, we discuss two ideas that are emerging as foundational in the literature on system safety and accident causation, namely that system safety is a 'control problem', and that it requires a 'system theoretic' approach to be dealt with. We clarify these characterizations and indicate research opportunities to be pursued along these directions. We conclude this work with two general recommendations

  4. Highlights from the literature on accident causation and system safety: Review of major ideas, recent contributions, and challenges

    International Nuclear Information System (INIS)

    Saleh, J.H.; Marais, K.B.; Bakolas, E.; Cowlagi, R.V.

    2010-01-01

    This work constitutes a short guide to the extensive but fragmented literature on accident causation and system safety. After briefly motivating the interest in accident causation and discussing the notion of a safety value chain, we delve into our multi-disciplinary review with discussions of Man Made Disasters, Normal Accident, and the High Reliability Organizations (HRO) paradigm. The HRO literature intersects an extensive literature on safety culture, a subject we then briefly touch upon. Following this discussion, we note that while these social and organizational contributions have significantly enriched our understanding of accident causation and system safety, they have important deficiencies and are lacking in their understanding of technical and design drivers of system safety and accident causation. These missing ingredients, we argue, were provided in part by the development of Probabilistic Risk Assessment (PRA). The idea of anticipating possible accident scenarios, based on the system design and configuration, as well as its technical and operational characteristics, constitutes an important contribution of PRA, which builds on and extends earlier contributions made by the development of Fault Tree and Event Tree Analysis. We follow the discussion of PRA with an exposition of the concept of safety barriers and the principle of defense-in-depth, both of which emphasize the functions and 'safety elements [that should be] deliberately inserted' along potential accident trajectories to prevent, contain, or mitigate accidents. Finally, we discuss two ideas that are emerging as foundational in the literature on system safety and accident causation, namely that system safety is a 'control problem', and that it requires a 'system theoretic' approach to be dealt with. We clarify these characterizations and indicate research opportunities to be pursued along these directions. We conclude this work with two general recommendations: (1) that more fundamental

  5. Characterization of Space Shuttle Thermal Protection System (TPS) Materials for Return-to-Flight following the Shuttle Columbia Accident Investigation

    Science.gov (United States)

    Wingard, Doug

    2006-01-01

    During the Space Shuttle Columbia Accident Investigation, it was determined that a large chunk of polyurethane insulating foam (= 1.67 lbs) on the External Tank (ET) came loose during Columbia's ascent on 2-1-03. The foam piece struck some of the protective Reinforced Carbon-Carbon (RCC) panels on the leading edge of Columbia's left wing in the mid-wing area. This impact damaged Columbia to the extent that upon re-entry to Earth, superheGed air approaching 3,000 F caused the vehicle to break up, killing all seven astronauts on board. A paper after the Columbia Accident Investigation highlighted thermal analysis testing performed on External Tank TPS materials (1). These materials included BX-250 (now BX-265) rigid polyurethane foam and SLA-561 Super Lightweight Ablator (highly-filled silicone rubber). The large chunk of foam from Columbia originated fiom the left bipod ramp of the ET. The foam in this ramp area was hand-sprayed over the SLA material and various fittings, allowed to dry, and manually shaved into a ramp shape. In Return-to-Flight (RTF) efforts following Columbia, the decision was made to remove the foam in the bipod ramp areas. During RTF efforts, further thermal analysis testing was performed on BX-265 foam by DSC and DMA. Flat panels of foam about 2-in. thick were sprayed on ET tank material (aluminum alloys). The DSC testing showed that foam material very close to the metal substrate cured more slowly than bulk foam material. All of the foam used on the ET is considered fully cured about 21 days after it is sprayed. The RTF culminated in the successful launch of Space Shuttle Discovery on 7-26-05. Although the flight was a success, there was another serious incident of foam loss fiom the ET during Shuttle ascent. This time, a rather large chunk of BX-265 foam (= 0.9 lbs) came loose from the liquid hydrogen (LH2) PAL ramp, although the foam did not strike the Shuttle Orbiter containing the crew. DMA testing was performed on foam samples taken fiom

  6. Neutronics and thermal-hydraulics coupling: some contributions toward an improved methodology to simulate the initiating phase of a severe accident in a sodium fast reactor

    International Nuclear Information System (INIS)

    Guyot, Maxime

    2014-01-01

    This project is dedicated to the analysis and the quantification of bias corresponding to the computational methodology for simulating the initiating phase of severe accidents on Sodium Fast Reactors. A deterministic approach is carried out to assess the consequences of a severe accident by adopting best estimate design evaluations. An objective of this deterministic approach is to provide guidance to mitigate severe accident developments and re-criticalities through the implementation of adequate design measures. These studies are generally based on modern simulation techniques to test and verify a given design. The new approach developed in this project aims to improve the safety assessment of Sodium Fast Reactors by decreasing the bias related to the deterministic analysis of severe accident scenarios. During the initiating phase, the subassembly wrapper tubes keep their mechanical integrity. Material disruption and dispersal is primarily one-dimensional. For this reason, evaluation methodology for the initiating phase relies on a multiple-channel approach. Typically a channel represents an average pin in a subassembly or a group of similar subassemblies. In the multiple-channel approach, the core thermal-hydraulics model is composed of 1 or 2 D channels. The thermal-hydraulics model is coupled to a neutronics module to provide an estimate of the reactor power level. In this project, a new computational model has been developed to extend the initiating phase modeling. This new model is based on a multi-physics coupling. This model has been applied to obtain information unavailable up to now in regards to neutronics and thermal-hydraulics models and their coupling. (author) [fr

  7. Radiological and dosimetric consequences in case of nuclear accident: taking them into account within the security approach and protection challenges

    International Nuclear Information System (INIS)

    Cogez, E.; Herviou, K.; Isnard, O.; Cessac, B.; Reales, N.; Quentric, E.; Quelo, D.

    2010-01-01

    This report first proposes a presentation of the 'defence in depth' concept which comprises five as much as possible independent levels: preventing operation anomalies and system failures, maintaining the installation within the authorized domain, controlling accidents within design hypotheses, preventing the degradation of accidental conditions and limiting consequences of severe accidents, limiting radiological consequences for population in case of important releases. Then, after a description of a release atmospheric dispersion and of its consequences, this report describes the consequences of two accident scenarios. The first accident is a failure of steam generator tubes, and the second a loss of primary coolant. It notably indicates the main released radionuclides, exposure levels at different distance for a given set of dispersion conditions

  8. The Euratom-Rosatom ERCOSAM-SAMARA projects on containment thermal-hydraulics of current and future LWRs for severe accident management

    International Nuclear Information System (INIS)

    Paladino, D.; Guentay, S.; Andreani, M.; Tkatschenko, I.; Brinster, J.; Dabbene, F.; Kelm, S.; Allelein, H. J.; Visser, D. C.; Benz, S.; Jordan, T.; Liang, Z.; Porcheron, E.; Malet, J.; Bentaib, A.; Kiselev, A.; Yudina, T.; Filippov, A.; Khizbullin, A.; Kamnev, M.; Zaytsev, A.; Loukianov, A.

    2012-01-01

    During a postulated severe accident with core degradation, hydrogen would form in the reactor pressure vessel mainly due to high temperatures zirconium-steam reaction and flow together with steam into the containment where it will mix with the containment atmosphere (steam-air). The hydrogen transport into the containment is a safety concern because it can lead to explosive mixtures through the associated phenomena of condensation, mixing and stratification. The ERCOSAM and SAMARA projects, co-financed by the European Union and the Russia, include various experiments addressing accident scenarios scaled down from existing plant calculations to different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT). The tests sequences aim to investigate hydrogen concentration build-up and stratification during a postulated accident and the effect of the activation of Severe Accident Management systems (SAMs), e.g. sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Analytical activities, performed by the project participants, are an essential component of the projects, as they aim to improve and validate various computational methods. They accompany the projects in the various phases; plant calculations, scaling to generic containment and to the different facilities, planning pre-test and post-test simulations are performed. Code benchmark activities on the basis of conceptual near full scale HYMIX facility will finally provide a further opportunity to evaluate the applicability of the various methods to the study of scaling issues. (authors)

  9. Thermal hydraulic analysis of aggressive secondary cooldown in small break loss of coolant accident with total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Han, S. J.; Im, H. K.; Yang, J. U.

    2003-01-01

    Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). To use RIA, the present study focuses on the detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study is to evaluate the success criteria of Aggressive Secondary Cooldown (ASC) in Small Break Loss Of Coolant Accident (SBLOCA) with total loss of High Pressure Safety Injection (HPSI) and to enhance the understanding of related thermal hydraulic behavior and phenomena. The accident scenario was 2 inch coldleg break LOCA without HPSI, with 1/2 Low Pressure Safety Injection (LPSI), and performing ASC limited by 55.6 .deg. C /hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip, which successively reaches the LPSI condition for about 1.5hr after starting ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria 1204.4 .deg. C (2200 .deg. F). In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that operator should maintain the adequate ASC operation. However, it is necessary to evaluate uncertainties arisen from the related parameters of the ASC operation

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

    International Nuclear Information System (INIS)

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

    1992-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Stempniewicz, M; Marks, P; Salwa, K

    1992-06-01

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

  12. Survey of Thermal-Fluids Evaluation and Confirmatory Experimental Validation Requirements of Accident Tolerant Cladding Concepts with Focus on Boiling Heat Transfer Characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ali, Amir [Univ. of New Mexico, Albuquerque, NM (United States); Liu, Maolong [Univ. of New Mexico, Albuquerque, NM (United States); Blandford, Edward [Univ. of New Mexico, Albuquerque, NM (United States)

    2016-06-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is working closely with the nuclear industry to develop fuel and cladding candidates with potentially enhanced accident tolerance, also known as accident tolerant fuel (ATF). Thermal-fluids characteristics are a vital element of a holistic engineering evaluation of ATF concepts. One vital characteristic related to boiling heat transfer is the critical heat flux (CHF). CHF plays a vital role in determining safety margins during normal operation and also in the progression of potential transient or accident scenarios. This deliverable is a scoping survey of thermal-fluids evaluation and confirmatory experimental validation requirements of accident tolerant cladding concepts with a focus on boiling heat transfer characteristics. The key takeaway messages of this report are: 1. CHF prediction accuracy is important and the correlations may have significant uncertainty. 2. Surface conditions are important factors for CHF, primarily the wettability that is characterized by contact angle. Smaller contact angle indicates greater wettability, which increases the CHF. Surface roughness also impacts wettability. Results in the literature for pool boiling experiments indicate changes in CHF by up to 60% for several ATF cladding candidates. 3. The measured wettability of FeCrAl (i.e., contact angle and roughness) indicates that CHF should be investigated further through pool boiling and flow boiling experiments. 4. Initial measurements of static advancing contact angle and surface roughness indicate that FeCrAl is expected to have a higher CHF than Zircaloy. The measured contact angle of different FeCrAl alloy samples depends on oxide layer thickness and composition. The static advancing contact angle tends to decrease as the oxide layer thickness increases.

  13. Application of thermal hydraulic and severe accident code SOCRAT/V3 to bottom water reflood experiment QUENCH-LOCA-0

    International Nuclear Information System (INIS)

    Vasiliev, A.D.; Stuckert, J.

    2013-01-01

    Highlights: ► QLOCA-0 test simulates a design basis LOCA NPP accident with maximum temperature 1300 K. ► Deep understanding of hydraulics and thermal mechanics under accident conditions is necessary. ► We model the test QLOCA-0 with bottom flooding using the Russian code SOCRAT/V3. ► Calculated and experimental data are in a good agreement. ► Experimental procedure is determined to reach a representative LOCA scenario in future tests. -- Abstract: The thermal hydraulic and SFD (severe fuel damage) best estimate computer modeling code SOCRAT/V3 has been used for the calculation of QUENCH-LOCA-0 experiment. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario of the LOCA (loss of coolant accident) nuclear power plant accident sequence in which the overheated up to 1300 K reactor core would be reflooded from the bottom by ECCS (emergency core cooling system). The first test QUENCH-LOCA-0 was successfully conducted at the KIT, Karlsruhe, Germany, in July 22, 2010, and was performed as the commissioning test for this series. The rod claddings are identical to that used in PWRs. The bundle was electrically heated in steam from 800 K to 1340 K with the heat-up rate of approximately 2.7 K/s. After cooling in the saturated steam the bottom flooding with water flow rate of about 100 g/s was initiated. The SOCRAT calculated results are in a good agreement with experimental data taking into account additional quenching due to water condensate entrainment at the steam cooling stage. SOCRAT/V3 has been used for estimation of further steps in experimental procedure to reach a representative LOCA scenario in future tests

  14. CANDU severe accident analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    Romania is a EU member since January first 2007. This country faces now new challenges which imply also the nuclear power reactors now in operation. Romania operates since 1996 a CANDU nuclear power reactor and soon will start up a second unit. In EU PWR reactors are mostly operated, so that the Romania's reactors have to meet EU standards. Safety analysis guidelines require to model severe accidents for reactors of this type. Starting from previous studies a thermal-hydraulic model for a degraded CANDU core was developed. The initiating event is assumed to be a LOCA with simultaneous loss of moderator and coolant and the failure of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield water tank surrounding the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data. (authors)

  15. Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

    2003-03-01

    Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). The present study focuses on detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model using RIA for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study in this year is to evaluate the success cri-teria of Aggressive Secondary Cooldown (ASC) in a Small Size Loss Of Coolant Accident (SBLOCA) without HPSI and to enhance the understanding of related thermal hydraulic behavior and phenomena. An effort was made to evaluate the system success criteria and a mission time for the recovery action by an operator to prevent the core damage for that accident scenario. The accident scenario for KSNP was a 2 inch coldleg break LOCA with a total loss of High Pressure Safety Injection (HPSI) and 1/2 Low Pressure Safety Injection (LPSI) available and perform-ing ASC limited by 55.6 .deg. C/hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip. It successively reached the LPSI condition for about 1.5hr after starting the ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria of 1204.4 .deg. C (2200 .deg. F). Sensitivity studies were performed for (1) cool-ant average temperature parameters, (2) ASC operation control method, (3) operation start time, (4) 1 inch break size. The present analysis identified thermal hydraulic phenomena and parameters affecting on the behavior, which consist of coolant break flow and inventory, parameters governing secondary heat removal, ASC operation control method, and its reference temperature parameters. In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that an operator should maintain the ade-quate ASC operation. However, it is necessary to evaluate the uncertainties arisen from the

  16. Progress and Updates of Regulatory Challenges and Safety Issues in Korea during Three Years after Fukushima Accident

    International Nuclear Information System (INIS)

    Lee, Young Eal; Kim, Kyun Tae

    2014-01-01

    Before the public fear on radiation risk caused by neighboring country's severe accident disappeared, a series of nuclear safety issues last 3 years made a few reactors shut down and the public trust much lower than before. Because of these scandals such as cover-ups, forged certificated items, corruption of manager of licensee and so on, many efforts made during three year after Fukushima accident on improving the nuclear safety were invalidated and even regulators as well as operators have been sharply criticized for its responsibility and transparency. This paper shares information on the progress and uprates achieved in Korea so far in connection with the safety issues caused during last 3 years and actions taken by the regulatory body. Before the public fear on radiation risk caused by neighboring country's severe accident disappeared, a series of nuclear safety issues last 3 years made a few reactors shut down and the public trust much lower than before. Because of these scandals such as cover-ups, forged certificated items, corruption of manager of licensee and so on, many efforts made during three year after Fukushima accident on improving the nuclear safety were invalidated and even regulators as well as operators have been sharply criticized for its responsibility and transparency

  17. Progress and Updates of Regulatory Challenges and Safety Issues in Korea during Three Years after Fukushima Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Eal; Kim, Kyun Tae [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-10-15

    Before the public fear on radiation risk caused by neighboring country's severe accident disappeared, a series of nuclear safety issues last 3 years made a few reactors shut down and the public trust much lower than before. Because of these scandals such as cover-ups, forged certificated items, corruption of manager of licensee and so on, many efforts made during three year after Fukushima accident on improving the nuclear safety were invalidated and even regulators as well as operators have been sharply criticized for its responsibility and transparency. This paper shares information on the progress and uprates achieved in Korea so far in connection with the safety issues caused during last 3 years and actions taken by the regulatory body. Before the public fear on radiation risk caused by neighboring country's severe accident disappeared, a series of nuclear safety issues last 3 years made a few reactors shut down and the public trust much lower than before. Because of these scandals such as cover-ups, forged certificated items, corruption of manager of licensee and so on, many efforts made during three year after Fukushima accident on improving the nuclear safety were invalidated and even regulators as well as operators have been sharply criticized for its responsibility and transparency.

  18. Predicting Consequences of Technological Disasters from Natural Hazard Events: Challenges and Opportunities Associated with Industrial Accident Data Sources

    Science.gov (United States)

    Wood, M.

    2009-04-01

    The increased focus on the possibility of technological accidents caused by natural events (Natech) is foreseen to continue for years to come. In this case, experts in prevention, mitigation and preparation activities associated with natural events will increasingly need to borrow data and expertise traditionally associated with the technological fields to carry out the work. An important question is how useful is the data for understanding consequences from such natech events. Data and case studies provided on major industrial accidents tend to focus on lessons learned for re-engineering the process. While consequence data are reported at least nominally in most reports, their precision, quality and completeness is often lacking. Consequences that are often or sometimes available but not provided can include severity and type of injuries, distance of victims from the source, exposure measurements, volume of the release, population in potentially affected zones, and weather conditions. Yet these are precisely the type of data that will aid natural hazard experts in land-use planning and emergency response activities when a Natech event may be foreseen. This work discusses the results of a study of consequence data from accidents involving toxic releases reported in the EU's MARS accident database. The study analysed the precision, quality and completeness of three categories of consequence data reported: the description of health effects, consequence assessment and chemical risk assessment factors, and emergency response information. This work reports on the findings from this study and discusses how natural hazards experts might interact with industrial accident experts to promote more consistent and accurate reporting of the data that will be useful in consequence-based activities.

  19. 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.)

  20. ATHENA-2D: A computer code for simulation of hypothetical recriticality accidents in a thermal neutron spectrum

    International Nuclear Information System (INIS)

    Schwinkendorf, K.N.

    1995-01-01

    In a damaged light water reactor core (as in the aftermath of a Three-Mile-Island-like core meltdown), water reflood is performed to carry off decay heat. The severely degraded geometry of the fuel debris bed may increase core reactivity with water reflood. Sufficient boron poisoning of the reflood water is therefore very important. One hypothetical accident is the reintroduction of cooling water that is insufficiently borated, resulting in the damaged reactor attaining criticality in this uncontrolled configuration. The goal in simulating this accident is the prediction of the energy release from the resulting transient

  1. 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)

  2. Response of turkey muscle satellite cells to thermal challenge. I. transcriptome effects in proliferating cells.

    Science.gov (United States)

    Reed, Kent M; Mendoza, Kristelle M; Abrahante, Juan E; Barnes, Natalie E; Velleman, Sandra G; Strasburg, Gale M

    2017-05-06

    Climate change poses a multi-dimensional threat to food and agricultural systems as a result of increased risk to animal growth, development, health, and food product quality. This study was designed to characterize transcriptional changes induced in turkey muscle satellite cells cultured under cold or hot thermal challenge to better define molecular mechanisms by which thermal stress alters breast muscle ultrastructure. Satellite cells isolated from the pectoralis major muscle of 7-weeks-old male turkeys from two breeding lines (16 weeks body weight-selected and it's randombred control) were proliferated in culture at 33 °C, 38 °C or 43 °C for 72 h. Total RNA was isolated and 12 libraries subjected to RNAseq analysis. Statistically significant differences in gene expression were observed among treatments and between turkey lines with a greater number of genes altered by cold treatment than by hot and fewer differences observed between lines than between temperatures. Pathway analysis found that cold treatment resulted in an overrepresentation of genes involved in cell signaling/signal transduction and cell communication/cell signaling as compared to control (38 °C). Heat-treated muscle satellite cells showed greater tendency towards expression of genes related to muscle system development and differentiation. This study demonstrates significant transcriptome effects on turkey skeletal muscle satellite cells exposed to thermal challenge. Additional effects on gene expression could be attributed to genetic selection for 16 weeks body weight (muscle mass). New targets are identified for further research on the differential control of satellite cell proliferation in poultry.

  3. Analytical and experimental assessment of TVS-2006 fuel assembly thermal-mechanical shape deformation at temperature modeling of a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Afanasiev, A.; Semishkin, V.; Makarov, V.; Matvienko, I.; Puzanov, D.

    2015-01-01

    generated for the consideration of TVS-2006 fuel assembly thermal-mechanical stability and its component shape deformation in loss-of-coolant accidents. The calculation results have shown that in the course of a design-basis loss-of-coolant accident TVS-2006 has a two-fold margin for stability loss at the longitudinal compression considering the increase in the force of FA compression due to thermal expansion of fuel rod bundle. The increase in the width across the flats is insignificant and it does not impede the post-accident core unloading. (authors) KEYWORDS: TVS- 2006, thermal-mechanical, shape deformation, modeling, loss of coolant

  4. Endocannabinoid Catabolic Enzymes Play Differential Roles in Thermal Homeostasis in Response to Environmental or Immune Challenge.

    Science.gov (United States)

    Nass, Sara R; Long, Jonathan Z; Schlosburg, Joel E; Cravatt, Benjamin F; Lichtman, Aron H; Kinsey, Steven G

    2015-06-01

    Cannabinoid receptor agonists, such as Δ(9)-THC, the primary active constituent of Cannabis sativa, have anti-pyrogenic effects in a variety of assays. Recently, attention has turned to the endogenous cannabinoid system and how endocannabinoids, including 2-arachidonoylglycerol (2-AG) and anandamide, regulate multiple homeostatic processes, including thermoregulation. Inhibiting endocannabinoid catabolic enzymes, monoacylglycerol lipase (MAGL) or fatty acid amide hydrolase (FAAH), elevates levels of 2-AG or anandamide in vivo, respectively. The purpose of this experiment was to test the hypothesis that endocannabinoid catabolic enzymes function to maintain thermal homeostasis in response to hypothermic challenge. In separate experiments, male C57BL/6J mice were administered a MAGL or FAAH inhibitor, and then challenged with the bacterial endotoxin lipopolysaccharide (LPS; 2 mg/kg ip) or a cold (4 °C) ambient environment. Systemic LPS administration caused a significant decrease in core body temperature after 6 h, and this hypothermia persisted for at least 12 h. Similarly, cold environment induced mild hypothermia that resolved within 30 min. JZL184 exacerbated hypothermia induced by either LPS or cold challenge, both of which effects were blocked by rimonabant, but not SR144528, indicating a CB1 cannabinoid receptor mechanism of action. In contrast, the FAAH inhibitor, PF-3845, had no effect on either LPS-induced or cold-induced hypothermia. These data indicate that unlike direct acting cannabinoid receptor agonists, which elicit profound hypothermic responses on their own, neither MAGL nor FAAH inhibitors affect normal body temperature. However, these endocannabinoid catabolic enzymes play distinct roles in thermoregulation following hypothermic challenges.

  5. Break size effect on the transient thermal-hydraulic behavior during the steam generator tube rupture accident

    International Nuclear Information System (INIS)

    Kang, K.H.; Park, H.S.; Cho, S.; Choi, N.H.; Chu, I.C.; Yun, B.J.; Kim, K.D.; Kim, Y.S.; Baek, W.P.; Choi, K.Y.

    2011-01-01

    In order to simulate the SGTR accident of the APR1400, integral effect tests were performed by simulating a double-ended rupture of a single and five U-tubes. Following the reactor trip, the primary system pressure decreased and the secondary system pressure increased until the MSSVs was opened to reduce the secondary system pressure. Break area affected the timings of the major events observed in the tests. Less heat transfer to the secondary side caused by earlier actuation of the safety injection pumps had more influence on the secondary pressure of the affected steam generator than the break flow. (author)

  6. Public health activities for mitigation of radiation exposures and risk communication challenges after the Fukushima nuclear accident

    International Nuclear Information System (INIS)

    Shimura, Tsutomu; Yamaguchi, Ichiro; Terada, Hiroshi; Kunugita, Naoki; Svendsen, Erik Robert

    2015-01-01

    Herein we summarize the public health actions taken to mitigate exposure of the public to radiation after the Fukushima accident that occurred on 11 March 2011 in order to record valuable lessons learned for disaster preparedness. Evacuations from the radiation-affected areas and control of the distribution of various food products contributed to the reduction of external and internal radiation exposure resulting from the Fukushima incident. However, risk communication is also an important issue during the emergency response effort and subsequent phases of dealing with a nuclear disaster. To assist with their healing process, sound, reliable scientific information should continue to be disseminated to the radiation-affected communities via two-way communication. We will describe the essential public health actions following a nuclear disaster for the early, intermediate and late phases that will be useful for radiological preparedness planning in response to other nuclear or radiological disasters. (author)

  7. GEYSER/TONUS: a coupled multi-D lumped parameter code for reactor thermal hydraulics analysis in case of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Petit, M.; Durin, M.; Gauvain, J.

    1995-12-31

    The safety requirements for future light water reactors include accounting for severe accidents in the design process. The design must now include mitigation features to limit pressure and temperature inside the building. Hydrogen concentration is also a major issue for severe accidents. Modelling the thermal hydraulics inside the containment requires the description of complex phenomena such as condensation, stratification, transport of gases and aerosols, heat transfers. The effect of mitigation systems will increase the heterogeneities in the building, and most of those phenomena can be coupled. The GEYSER/TONUS multi-dimensional computer code is under development at CEA Saclay to model this complex situation. It allow the coupling of parts of the containment described in a lumped parameter manner, together with meshed parts. Emphasis is put on the numerical methods used to solve the transient problem, and physical models of classical lumped parameters codes will be adapted for the spatially described zones. The code is developed in the environment of the CASTEM 2000/TRIO EF system which allows to construct sophisticated applications based upon it. (J.S.). 22 refs., 1 fig.

  8. GEYSER/TONUS: a coupled multi-D lumped parameter code for reactor thermal hydraulics analysis in case of severe accidents

    International Nuclear Information System (INIS)

    Petit, M.; Durin, M.; Gauvain, J.

    1995-01-01

    The safety requirements for future light water reactors include accounting for severe accidents in the design process. The design must now include mitigation features to limit pressure and temperature inside the building. Hydrogen concentration is also a major issue for severe accidents. Modelling the thermal hydraulics inside the containment requires the description of complex phenomena such as condensation, stratification, transport of gases and aerosols, heat transfers. The effect of mitigation systems will increase the heterogeneities in the building, and most of those phenomena can be coupled. The GEYSER/TONUS multi-dimensional computer code is under development at CEA Saclay to model this complex situation. It allow the coupling of parts of the containment described in a lumped parameter manner, together with meshed parts. Emphasis is put on the numerical methods used to solve the transient problem, and physical models of classical lumped parameters codes will be adapted for the spatially described zones. The code is developed in the environment of the CASTEM 2000/TRIO EF system which allows to construct sophisticated applications based upon it. (J.S.). 22 refs., 1 fig

  9. GEYSER/TONUS: A coupled multi-D lumped parameter code for reactor thermal hydraulics analysis in case of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Petit, M.; Durin, M.; Gauvain, J. [Commissariat a l`Energie Atomique, Gif sur Yvette (France)

    1995-09-01

    In many countries, the safety requirements for future light water reactors include accounting for severe accidents in the design process. As far as the containment is concerned, the design must now include mitigation features to limit the pressure and temperature inside the building. Hydrogen concentration is also a major issue for severe accidents. In this context, new needs appear for the modeling of the thermal hydraulics inside the containment. It requires the description of complex phenomena such as condensation, stratification, transport of gases and aerosols, heat transfers. Moreover, the effect of mitigation systems will increase the heterogeneities in the building, and most of those phenomena can be coupled, as for example hydrogen stratification and condensation. To model such a complex situation, the use of multi-dimensional computer codes seems to be necessary in case of large volumes. The aim of the GEYSER/TONUS computer code is to fulfill this need. This code is currently under development at CEA in Saclay. It will allow the coupling of parts of the containment described in a lumped parameter manner, together with meshed parts. Emphasis is put on the numerical methods used to solve the transient problem, as the objective is to be able to treat complete scenarios. Physical models of classical lumped parameters codes will adapted for the spatially described zones. The code is developed in the environment of the CASTEM 2000/TRIO EF system which allows, thanks to its modular conception, to construct sophisticated applications based upon it.

  10. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1997-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  11. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1998-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  12. Response of Turkey Muscle Satellite Cells to Thermal Challenge. II. Transcriptome Effects in Differentiating Cells

    Directory of Open Access Journals (Sweden)

    Kent M. Reed

    2017-11-01

    Full Text Available Background: Exposure of poultry to extreme temperatures during the critical period of post-hatch growth can seriously affect muscle development and thus compromise subsequent meat quality. This study was designed to characterize transcriptional changes induced in turkey muscle satellite cells by thermal challenge during differentiation. Our goal is to better define how thermal stress alters breast muscle ultrastructure and subsequent development.Results: Skeletal muscle satellite cells previously isolated from the Pectoralis major muscle of 7-wk-old male turkeys (Meleagris gallopavo from two breeding lines: the F-line (16 wk body weight-selected and RBC2 (randombred control line were used in this study. Cultured cells were induced to differentiate at 38°C (control or thermal challenge temperatures of 33 or 43°C. After 48 h of differentiation, cells were harvested and total RNA was isolated for RNAseq analysis. Analysis of 39.9 Gb of sequence found 89% mapped to the turkey genome (UMD5.0, annotation 101 with average expression of 18,917 genes per library. In the cultured satellite cells, slow/cardiac muscle isoforms are generally present in greater abundance than fast skeletal isoforms. Statistically significant differences in gene expression were observed among treatments and between turkey lines, with a greater number of genes affected in the F-line cells following cold treatment whereas more differentially expressed (DE genes were observed in the RBC2 cells following heat treatment. Many of the most significant pathways involved signaling, consistent with ongoing cellular differentiation. Regulation of Ca2+ homeostasis appears to be significantly affected by temperature treatment, particularly cold treatment.Conclusions: Satellite cell differentiation is directly influenced by temperature at the level of gene transcription with greater effects attributed to selection for fast growth. At lower temperature, muscle-associated genes in the

  13. Electro-thermal injuries due to high-current accidents with special regard to the skeletal muscles

    Energy Technology Data Exchange (ETDEWEB)

    Rausch, M.

    1982-02-25

    Seven patients suffering from high-current accidents with different degrees of severity are reported on. Damages to the striped musculature are preferentially considered. When the function of the cellular membrane is affected by an electric current, the Na/sup +//K/sup +/ mechanism collapses due to an increase of membrane permeability for calcium. Depolarisation, development of contractures, changes of the fibrillar structures and decay of the stripes or swelling of the Q (A) substance were found to indicate damage. A particular effect of electrothermal muscle damages is the affection of other organs, e.g. of the kidneys, provoked by an increased chromoprotein production and a higher output of myoglobin and hemoglobin. Since in case of an accident due to electric current affecting the musculature, more acid substances enter the vascular system, an acute renal damage provoked by an alkalinisation of the urine, shall be treated by drugs. In most cases, an anuria, occurring several days after deep burn of musculature, is a so-called constipational anuria, if it is not an anuria induced by a decrease of blood pressure or by shock. The therapy of wounds due to electrical burning depends on the size of the affected skin area and down to which depth the tissue is burnt. Particular attention must be paid to the affection of bones being situated in the depth of wounds due to burning. In the last part of the study possible reconstructive and ensuing rehabilitation measures are indicated.

  14. 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

  15. Severe accident behavior

    International Nuclear Information System (INIS)

    Denning, R.S.

    1986-01-01

    The purpose of this paper is to provide an overview of severe accident behavior. The term source term is defined and a brief history of the regulatory use of source term is presented. The processes in severe accidents in light water reactors are described with particular emphasis on the relationships between accident thermal-hydraulics and chemistry. Those factors which have the greatest impact on predicted source terms are identified. Design differences between plants that affect source term estimation are also described. The principal unresolved issues are identified that are the focus of ongoing research and debate in the technical community

  16. Acidentes químicos ampliados: um desafio para a saúde pública The increase in chemical accidents: a challenge for public health

    Directory of Open Access Journals (Sweden)

    Carlos M. de Freitas

    1995-12-01

    Full Text Available Os acidentes envolvendo substâncias perigosas nas atividades de transporte, armazenamento e produção industrial de produtos químicos constituem um sério risco à saúde e ao meio ambiente. Objetiva-se discutir, no âmbito da saúde pública, alguns dos desafios que esses tipos de acidentes colocam, principalmente para os países de economia periférica. Através da combinação de informações quantitativas e qualitativas, foram definidos e caracterizados esses tipos de acidentes e seus diversos riscos. Esses acidentes têm se apresentado com a maior gravidade nos países de economia periférica, embora a maioria deles venha ocorrendo sem o adequado registro de informações básicas para a avaliação e vigilância, como é demonstrado no caso do Rio de Janeiro (Brasil. Além da tarefa de se avaliar as conseqüências de eventos, por vezes extremamente complexos, coloca-se também, a de formular estratégias de controle e prevenção em realidades sociais que configuram um terreno fértil para a ocorrência e agravamento dos mesmos.Chemical accidents involving explosions, large fires and leakages of hazardous substances occuring during transport, storage and industrial production of chemicals constitute a real challeng to health, environmental and industrial safety professionals. The aim of this article is to discuss the main questions that this kind of accident provokes, in terms of public helth, particularly in developing countries such as Brazil. The paper defines and characterises these accidents and the various health risk they involve excluding the leakages of hazardous substances during "normal" production in industry - through the combination of quantitative and qualitative information drawn from the international literature on the subject. From some examples of chemical accidents such as occurred in Bophal (Índia, Vila Socó (Brazil, São Paulo (México and data of the World Health Organization (WHO, the authors seek to show

  17. A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong [Korea Power Engineering Company, Inc, 150 Deokjin-dong, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2006-07-01

    This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

  18. A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

    International Nuclear Information System (INIS)

    Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong

    2006-01-01

    This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

  19. 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...

  20. Transient thermal hydraulic analysis of the IAEA 10 MW MTR reactor during Loss of Flow Accident to investigate the flow inversion

    International Nuclear Information System (INIS)

    AL-Yahia, Omar S.; Albati, Mohammad A.; Park, Jonghark; Chae, Heetaek; Jo, Daeseong

    2013-01-01

    Highlights: • Transient analyses of a slow and fast LOFA were investigated. • A reactor kinetic and thermal hydraulic coupled model was developed. • Based on force balance, the flow rate during flow inversion was determined. • Flow inversion in a hot channel occurred earlier than in an average channel. • Two temperature peaks were observed during both slow and fast LOFA. - Abstract: Transient analyses of the IAEA 10 MW MTR reactor are investigated during a fast and slow Loss of Flow Accident (LOFA) with a neutron kinetic and thermal hydraulic coupling model. A spatial-dependent thermal hydraulic technique is adopted for analyzing the local thermal hydraulic parameters and hotspot location during a flow inversion. The flow rate through the channel is determined in terms of a balance between driving and preventing forces. Friction and buoyancy forces act as resistance of the flow before a flow inversion while buoyancy force becomes the driving force after a flow inversion. By taking into account the buoyancy effect to determine the flow rate, the difference in the flow inversion time between hot and average channels is investigated: a flow inversion occurs earlier in the hot channel than in an average channel. Furthermore, the movement of the hotspot location before and after a flow inversion is investigated for a slow and fast LOFA. During a flow inversion, two temperature peaks are observed: (1) the first temperature peak is at the initiation of the LOFA, and (2) the second temperature peak is when a flow inversion occurs. The maximum temperature of the cladding is found at the second temperature peak for both LOFA analyses, and is lower than the saturation temperature

  1. Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

  2. Steering committee for the management of the post-accidental phase of a nuclear accident or of a radiological situation (CODIRPA). Work group nr 4. Response to health challenges after a radiological accident - Stage report issue nr 2 of 2007 November 5; Synthesis of doctrinal elements and recommendations; Consultative meeting - February 2009; Final report March 2011; Report synthesis

    International Nuclear Information System (INIS)

    Bernier, Marie-Odile; Challeton-de Vathaire, Cecile; Catelinois, Olivier; Pirard, Philippe; Collignon, Albert; Corblet, Sibylle; Empereur Bissonnet, Pascal; Fite, Johanna; Mehl Auget, Isabelle; Fleutot, Jean-Baptiste; Gavel, Yves; Geis-Bonnemains, Nathalie; Geneau, Christian; Guagniere, Bertrand; Janin, Claire; Lang, Thierry; Marielle, Schmitt; N'Diaye, Bakhao; Raoul, Christophe; Ricoux, Christine; Schwoebel, Valerie; Telion, Caroline; Tillier, Claude; Verger, Pierre; Volant, Philippe

    2007-01-01

    A first report describes effects of radiological accidents on health, and possible scenarios (dose assessments, exposed population, health challenges), gives an overview of methods and arrangements used to count exposed populations and to assess received doses, and presents the foreseen reception centres. It describes the health management organisation (medical and psychological care of exposed population, public information), and reports an analysis of health risks associated with the accident. The next document contains a synthesis of doctrinal elements and recommendations regarding information acquisition, the management of health consequences of the accident, the assessment of the health impact. Propositions made by the work group to face health challenges are presented in a Power Point presentation. The final report addresses the context of preparation of the response to a radiological accident in France (studied accidents and scenarios, public health challenges), proposes a chronological synthesis of actions to be undertaken during a preparation phase, an emergency phase, a transition phase, and a long term phase), and a detailed presentation of main actions to be undertaken (medical and psychological care, reception centres, counting, health risk analysis, health information)

  3. Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

    2003-01-01

    To support the development of a Probabilistic Safety Assessment (PSA) model usable in Riskinformed Applications (RIA) for Korea Standard Nuclear power Plants (KSNP), we have performed a thermal hydraulic analysis of Aggressive Secondary Cooldown (ASC) in a 2-inch Small Break Loss Of Coolant Accident (SBLOCA) with a total loss of High Pressure Safety Injection (HPSI). The present study focuses on the estimation of the success criteria of ASC, and the enhanced understanding of the detailed thermal hydraulic behavior and phenomena. The results have shown that the Reactor Coolant System (RCS) pressure can be reduced to the Low Pressure Safety Injection (LPSI) operation conditions without core damage. It was also shown that more relaxed success criteria compared to those in the previous PSA models of KSNP could be used in the new PSA model. However, it was found that the results could be affected by various parameters related with ASC operation, i.e., reference temperature for the calculation of the cooldown rate and its control method

  4. Investigation of 3D spatial effect on point kinetics estimation of the thermal hydraulics code RELAP for the analysis of MSLB accident of KK-NP

    International Nuclear Information System (INIS)

    Bera, S.; Pradhan, S.K.; Dubey, S.K.; Gupta, S.K.

    2011-01-01

    In general safety analyses of design basis accident of NPPs are being carried out using system thermal hydraulics code like RELAP. In RELAP, power is calculated based on point kinetics approximation, which virtually ignores the space and energy dependence of neutron flux. To include the space and energy dependence of neutron flux, three-dimensional neutronics code TRIHEXFA has been externally coupled with RELAP through interface program, TRIHEXFA-RELAP Interface Program (TRIP). Calculation methodology of TRIP program is based on adiabatic approximation. In the adiabatic approximation the neutron flux is being factored into spatial and amplitude part. Spatial part of flux is slowly varying with time whereas amplitude part is strongly varying function. The RELAP controls the transient time steps. Transient time is divided into several major and minor time steps. Minor time step is the sub-step of major time step. Thermal hydraulics and neutronics data are exchanged at each major time step. Spatial part of neutron flux has been updated at each major time step using TRIHEXFA code. But amplitude part of the neutron flux is calculated at each minor time step using RELAP code. Convergence of results of the coupled code, TRIP has been checked through coupling time step descritization study. This study determines the minimum coupling time step. Transient concerning VVER-1000 Main Steam Line Break, MSLB has been considered to investigate the space-time effect on point kinetics. MSLB occurs as a consequence of the rupture of one steam line upstream of main steam line isolation valves. Reference design and data from Kudankulam Nuclear Power Plant (KK-NPP) are used for the analysis. From this investigation it is found that TRIP significantly overestimates the maximum reactor power against uncoupled RELAP result. The time of scram also occur six seconds earlier in TRIP calculation compared to the RELAP. This exercise has also shown a proof of principle that coupling 3D

  5. PBDOWN - a computer code for simulating core material discharge and thermal to mechanical energy conversion in LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Royl, P.

    1981-01-01

    PBDOWN is a computer code that simulates the blowdown of confined boiling materials ('pools') into a colder upper coolant plenum as time dependent ejection and expansion with consideration of a few selected exchange processes. Its application is restricted to situations resulting from hypothetical loss of flow (LOF) accidents in LMFBR's, where enough voiding has occured, that in core sodium vapor pressures become negligible. PBDOWN considers one working fluid for the discharge process (either fuel or steel) and a maximum of two working fluids (either fuel and sodium or steel and sodium) for the expansion process in the upper coolant plenum. Entrainment of sodium at the accelerated bubble liquid interfaces is mechanistically calculated by a Taylor instability entrainment model. Simulation of a hemispherical expansion form together with this mechanistic entrainment model gives a new integrated calculation of the time dependent sodium mass in the bubble. The paper summarizes the basic equations and assumptions of this computer model. Sample results compare different heat transfer and Na entrainment models during steel and fuel driven discharge processes. Mechanistic sodium entrainment simulation for SNR-type reactors coupled with a realistic heat transfer model is shown to reduce the integral mechanical work potential by a factor of 1.3 to 2.0 over the isentropic energy of the discharge working fluids. (orig.)

  6. The challenge and countermeasure for radioactivity monitor in ocean--the reflection derived from Fukushima nuclear power plant accident

    International Nuclear Information System (INIS)

    Liu Guangshan

    2014-01-01

    Environmental radioactivity monitor (ERM) follows with interest raising level, sink and transportation of radioactive materials in the environment. There are two kinds of challenges for marine radioactivity monitor (MRM), for which one concerned with facilities and methods that lead the very difficult measurement of MRM. Up to now, the bulk of sample must collected and prepared for most MRM by radiometry (counting method) and long acquisition period is needed. The long time from sampling to derived results postponed the time of the scientists and government officials learning contamination distribution information. Developing accelerator mass spectrometry (AMS) can overcome above difficulty because AMS method need one tenth of sample and one tenth measuring time needed by radiometric method. The second challenge is lack of the behavior research of radionuclide in the ocean. In the ocean, radionuclide behavior is constrained by physical and biogeochemical process, following that the radionuclides are transported from released sea area to where far from source area, from upper water to deep and bottom waters, and buried to sediment. There are some reports about radionuclide transportation with physical oceanographical process, but few paper about radionuclide biogeochemistry has been found. Lesser theory knowledge and lesser investigation data make scientists worked on MRM puzzling. The radionuclide behavior in the ocean can be studied by the method used in marine biogeochemistry. (author)

  7. Preventing accidents

    Science.gov (United States)

    2005-08-01

    As the most effective strategy for improving safety is to prevent accidents from occurring at all, the Volpe Center applies a broad range of research techniques and capabilities to determine causes and consequences of accidents and to identify, asses...

  8. From centralized hydro-thermal to decentralized renewable : the Austrian challenge

    International Nuclear Information System (INIS)

    Brauner, G.

    2007-01-01

    Austria's electricity production scheme is currently based on 67 per cent hydro power, 25 per cent fossil thermal power and 8 per cent new renewable energy, including wind, biomass, and photovoltaic conversion. As demand on centralized energy systems increases, the power generation, transmission and distribution system in the country will have to be restructured to increase the portion of regenerative generation. Decentralized micro-grid structures are a promising concept for meeting the target of efficiency improvement and renewable generation because they make it possible to use regenerative energy where it is collected. The most efficient way to meet the challenges of efficiency improvement and emission reduction is the replacement of old coal-fired power stations by new combined-cycle gas power stations with combined heat and power. This will reduce both carbon dioxide emissions as well as specific fossil energy demand. However, in just 1 decade the emission reduction achieved by fuel switching and higher efficiency of power stations will be equalized by higher demand, which is growing at an annual rate of 2 per cent. Future energy problems will be characterized by the security of supply; climatic change and the need for clean energy technologies; the lack of transmission-routes; and growing energy demand. The regenerative energy in Austria currently consists of hydraulic large run-of-river hydraulic generation as well as small scale hydro and large scale hydro storage systems. Biomass currently has a total generating capacity of 280 MW and forms 3.14 per cent of generation in Austria. Wind has reached a capacity of 950 MW and forms 3.8 per cent of generation in Austria. It was concluded that the future philosophy in energy supply will not be to try to increase regenerative generation according to growing demand, but to improve end-use efficiency and reduce demand according to the renewable regional potential. 4 refs., 4 figs

  9. Accident management for severe accidents

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  10. THERMAL-HYDRAULIC ANALYSIS OF SMR WITH NATURALLY CIRCULATING PRIMARY SYSTEM DURING LOSS OF FEED WATER ACCIDENT

    Directory of Open Access Journals (Sweden)

    Susyadi Susyadi

    2016-09-01

    ABSTRAK Reaktor daya kecil modular (SMR memiliki beberapa keunggulan dibanding reaktor daya besar konvensional. Dengan disain yang lebih sederhana dan terintegrasi, penerapan hukum alamiah untuk sistem keselamatannya dan biaya modal yang rendah, reaktor ini sangat cocok untuk dibangun di Indonesia. Salah satunya disain SMR yang sedang dikembangkan menerapkan gaya penggerak alami untuk sistim pendingin primernya. Dengan disain seperti itu, adalah sangat penting untuk memahami implikasinya terhadap aspek keselamatan pada seluruh kondisi operasi. Salah satu yang perlu diinvestigasi adalah kecelakaan kehilangan air umpan (LoFW. Pada studi ini, dilakukan analisis kinerja thermal hidrolik SMR yang menggunakan sistim pendinginan primer sirkulasi alam saat kecelakaan LoFW. Tujuannya adalah untuk menginvestigasi karakteristik aliran sistem primer saat kecelakaan LoFW dan untuk memastikan apakah aliran sirkulasi alam cukup untuk memindahkan panas dari teras guna menjaga kondisi tetap aman selama kecelakaan tersebut. Metoda yang digunakan adalah dengan merepresentasikan sistem reaktor ke dalam model-model generik program RELAP5 dan melakukan simulasi numerik. Hasil perhitungan menunjukkan bahwa setelah kejadian pemicu dan trip reaktor, pada sisi primer laju alirnya berfluktuasi secara signifikan dan temperatur pendinginnya menurun secara bertahap sedangkan  pada sisi sekunder kondisi uap berubah menjadi uap jenuh. Laju alir turun dari ~711 kg/detik menjadi ~263 kg/detik sebelum kembali naik lagi pada t=~46 detik. Saat laju alir di titik terendah, temperatur pusat bahan bakar dan fluida pendingin adalah sekitar  ~565 K dan  ~554 K, yang menujukkan bahwa temperatur bahan bakar masih jauh di bawah batas disain dan temperatur fluidanya juga berada di bawah titik saturasi. Keadaan ini menunjukkan bahwa saat transien kedua parameter utama termohidrolik reaktor tetap dalam kondisi yang dapat diterima sehingga dapat disimpulkan  bahwa saat  kecelakaan kehilangan air umpan, SMR

  11. Thermal Protection for Mars Sample Return Earth Entry Vehicle: A Grand Challenge for Design Methodology and Reliability Verification

    Science.gov (United States)

    Venkatapathy, Ethiraj; Gage, Peter; Wright, Michael J.

    2017-01-01

    Mars Sample Return is our Grand Challenge for the coming decade. TPS (Thermal Protection System) nominal performance is not the key challenge. The main difficulty for designers is the need to verify unprecedented reliability for the entry system: current guidelines for prevention of backward contamination require that the probability of spores larger than 1 micron diameter escaping into the Earth environment be lower than 1 million for the entire system, and the allocation to TPS would be more stringent than that. For reference, the reliability allocation for Orion TPS is closer to 11000, and the demonstrated reliability for previous human Earth return systems was closer to 1100. Improving reliability by more than 3 orders of magnitude is a grand challenge indeed. The TPS community must embrace the possibility of new architectures that are focused on reliability above thermal performance and mass efficiency. MSR (Mars Sample Return) EEV (Earth Entry Vehicle) will be hit with MMOD (Micrometeoroid and Orbital Debris) prior to reentry. A chute-less aero-shell design which allows for self-righting shape was baselined in prior MSR studies, with the assumption that a passive system will maximize EEV robustness. Hence the aero-shell along with the TPS has to take ground impact and not break apart. System verification will require testing to establish ablative performance and thermal failure but also testing of damage from MMOD, and structural performance at ground impact. Mission requirements will demand analysis, testing and verification that are focused on establishing reliability of the design. In this proposed talk, we will focus on the grand challenge of MSR EEV TPS and the need for innovative approaches to address challenges in modeling, testing, manufacturing and verification.

  12. Cernavoda CANDU severe accident evaluation

    International Nuclear Information System (INIS)

    Negut, G.; Marin, A.

    1997-01-01

    The papers present the activities dedicated to Romania Cernavoda Nuclear Power Plant first CANDU Unit severe accident evaluation. This activity is part of more general PSA assessment activities. CANDU specific safety features are calandria moderator and calandria vault water capabilities to remove the residual heat in the case of severe accidents, when the conventional heat sinks are no more available. Severe accidents evaluation, that is a deterministic thermal hydraulic analysis, assesses the accidents progression and gives the milestones when important events take place. This kind of assessment is important to evaluate to recovery time for the reactor operators that can lead to the accident mitigation. The Cernavoda CANDU unit is modeled for the of all heat sinks accident and results compared with the AECL CANDU 600 assessment. (orig.)

  13. 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

  14. Investigation of break location effects on thermal-hydraulics during intermediate break loss-of-coolant accident experiments at ROSA-III

    International Nuclear Information System (INIS)

    Koizumi, Yasuo; Tasaka, Kanji

    1986-01-01

    The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25 % main recirculation pump suction line break (MRPS-B) experiments, the 21 % single-ended jet pump drive line break (JPD-B) experiment and the 15 % main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests. In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop. (author)

  15. Challenges in thermal design of industrial single-phase power inverter

    Directory of Open Access Journals (Sweden)

    Ninković Predrag

    2016-01-01

    Full Text Available This paper presents the influence of thermal aspects in design process of an industrial single-phase inverter, choice of its topology and components. Stringent design inputs like very high overload level, demand for natural cooling and very wide input voltage range have made conventional circuit topology inappropriate therefore asking for alternative solution. Different power losses calculations in semiconductors are performed and compared, outlining the guidelines how to choose the final topology. Some recommendations in power magnetic components design are given. Based on the final project, a 20kVA single-phase inverter for thermal power plant supervisory and control system is designed and commissioned.

  16. Thermal management of hybrid vehicles. Challenges and solutions; Thermomanagement von Hybridfahrzeugen. Herausforderungen und Loesungen

    Energy Technology Data Exchange (ETDEWEB)

    Heckenberger, T. [Behr GmbH und Co. KG, Stuttgart (Germany)

    2007-10-15

    Higher thermal management demands are placed on hybrid vehicles than on conventional vehicles. The thermal comfort of the passengers must be assured when the combustion engine shuts down, i.e. during start-stop phases and during purely electric driving. Likewise, the particular hybrid modules and components must be cooled to temperature levels that are too low for conventional cooling circuits. Behr is presently developing cooling and climate control systems for all three types of hybrid vehicles: micro-, mild, and full hybrids. (orig.)

  17. 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

  18. 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

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

    International Nuclear Information System (INIS)

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

    1987-02-01

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

  20. Accident Statistics

    Data.gov (United States)

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

  1. 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)

  2. Sports Accidents

    CERN Multimedia

    Kiebel

    1972-01-01

    Le Docteur Kiebel, chirurgien à Genève, est aussi un grand ami de sport et de temps en temps médecin des classes genevoises de ski et également médecin de l'équipe de hockey sur glace de Genève Servette. Il est bien qualifié pour nous parler d'accidents de sport et surtout d'accidents de ski.

  3. Radiation accidents

    International Nuclear Information System (INIS)

    Poplavskij, K.K.; Smorodintseva, G.I.

    1978-01-01

    On the basis of a critical analysis of the available data on causes and consequences of radiation accidents (RA), a classification of RA by severity (five groups of accidents) according to biomedical consequences and categories of exposed personnel is proposed. A RA is defined and its main characteristics are described. Methods of RA prevention are proposed, as is a plan of specific measures to deal with RA in accordance with the proposed classification

  4. Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

    International Nuclear Information System (INIS)

    Sun, K.; Chenu, A.; Mikityuk, K.; Krepel, J.; Chawla, R.

    2012-01-01

    The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

  5. Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Chenu, A. [Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Mikityuk, K.; Krepel, J. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland)

    2012-07-01

    The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

  6. The post-accident nuclear issue: the new crisis expertise challenges for the IRSN; Post-accidentel nucleaire: les nouveaux challenges de l'expertise de crise a L'IRSN

    Energy Technology Data Exchange (ETDEWEB)

    Champion, D. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Direction de l' environnement et de l' intervention, 78 - Le Vesinet (France)

    2010-07-01

    The author reports the work performed by two work groups conducted by the IRSN (the French Radioprotection and Nuclear Safety Institute), the first one on the issue of assessment of radiological and dosimetric consequences in a post-accident situation, and the second one on hypotheses to be used to perform predictive assessments of these consequences. First dealing with the end of the emergency phase, he describes how to anticipate actions of protection against immediate post-accident consequences: orientation of the expertise strategy based on the CODIRPA's doctrine, post-accident zoning based on predictive indicators, use of reasonably prudent hypotheses for the first predictive assessments, importance of initial radioactive deposits to perform predictive assessments. Then, the author presents an iterative method of assessment of post-accident consequences: organization of environment radioactivity measurement programmes, periodic update of mapping of initial deposit and of actual deposits at a given time

  7. Scientific and Technical Challenges in Thermal Transport and Thermoelectric Materials and Devices

    KAUST Repository

    O'Dwyer, Colm

    2017-01-19

    This paper considers the state-of-the-art and open scientific and technological questions in thermoelectric materials and devices, from phonon engineering and scattering methods, to new and complex materials and their thermoelectric behavior. The paper also describes recent approaches to create structural and compositional material systems designed to enhance the thermoelectric figure of merit and power factors. We also summarize and contextualize recent advances in the use of superlattice structures and porosity or roughness to influence phonon scattering mechanisms and detail some advances in integrated thermoelectric materials for generators and coolers for thermally stable photonic devices.

  8. Scientific and Technical Challenges in Thermal Transport and Thermoelectric Materials and Devices

    KAUST Repository

    O'Dwyer, Colm; Chen, Renkun; He, Jr-Hau; Lee, Jaeho; Razeeb, Kafil M.

    2017-01-01

    This paper considers the state-of-the-art and open scientific and technological questions in thermoelectric materials and devices, from phonon engineering and scattering methods, to new and complex materials and their thermoelectric behavior. The paper also describes recent approaches to create structural and compositional material systems designed to enhance the thermoelectric figure of merit and power factors. We also summarize and contextualize recent advances in the use of superlattice structures and porosity or roughness to influence phonon scattering mechanisms and detail some advances in integrated thermoelectric materials for generators and coolers for thermally stable photonic devices.

  9. The promise and challenges of cermet fueled nuclear thermal propulsion reactors

    International Nuclear Information System (INIS)

    Brengle, R.G.; Harty, R.B.; Bhattacharyya, S.K.

    1993-06-01

    The use of cermet fuels in nuclear thermal propulsion systems was examined and the characteristics of systems using these fuel forms is discussed in terms of current mission and safety requirements. For use at high temperatures cermet fueled reactors utilize ceramic fuels with refractory metals as the matrix material. Cermet fueled reactors tend to be heavy when compared to concepts that utilize graphite as the fuel matrix because of the high density of the refractory metal matrix which makes up 20-40 percent of the total volume. On the positive side the metal matrix is strong and more resistant to loads from either the launch or flow induced vibration. The compatibility of the tungsten cermet with hydrogen is excellent and lifetimes of several hours is certainly achievable. Probably the biggest drawback to cermet nuclear thermal propulsion concepts is that the amount of actual data to support the theoretical conclusions is small. In fact there is no data under representative conditions of temperature, propellant and flux for the required fuel burnup. Although cermet systems appear to be attractive, the lack of fuel data at representative conditions does not allow reliable comparisons of cermet systems to systems where fuel data is available. 10 refs

  10. Nanofluid two-phase flow and thermal physics: a new research frontier of nanotechnology and its challenges.

    Science.gov (United States)

    Cheng, Lixin; Bandarra Filho, Enio P; Thome, John R

    2008-07-01

    Nanofluids are a new class of fluids engineered by dispersing nanometer-size solid particles in base fluids. As a new research frontier, nanofluid two-phase flow and thermal physics have the potential to improve heat transfer and energy efficiency in thermal management systems for many applications, such as microelectronics, power electronics, transportation, nuclear engineering, heat pipes, refrigeration, air-conditioning and heat pump systems. So far, the study of nanofluid two-phase flow and thermal physics is still in its infancy. This field of research provides many opportunities to study new frontiers but also poses great challenges. To summarize the current status of research in this newly developing interdisciplinary field and to identify the future research needs as well, this paper focuses on presenting a comprehensive review of nucleate pool boiling, flow boiling, critical heat flux, condensation and two-phase flow of nanofluids. Even for the limited studies done so far, there are some controversies. Conclusions and contradictions on the available nanofluid studies on physical properties, two-phase flow, heat transfer and critical heat flux (CHF) are presented. Based on a comprehensive analysis, it has been realized that the physical properties of nanofluids such as surface tension, liquid thermal conductivity, viscosity and density have significant effects on the nanofluid two-phase flow and heat transfer characteristics but the lack of the accurate knowledge of these physical properties has greatly limited the study in this interdisciplinary field. Therefore, effort should be made to contribute to the physical property database of nanofluids as a first priority. Secondly, in particular, research on nanofluid two-phase flow and heat transfer in microchannels should be emphasized in the future.

  11. Modelling and analysis of severe accidents for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Tusheva, Polina

    2012-01-01

    Accident conditions involving significant core degradation are termed severe accidents /IAEA: NS-G-2.15/. Despite the low probability of occurrence of such events, the investigation of severe accident scenarios is an important part of the nuclear safety research. Considering a hypothetical core melt down scenario in a VVER-1000 light water reactor, the early in-vessel phase focusing on the thermal-hydraulic phenomena, and the late in-vessel phase focusing on the melt relocation into the reactor pressure vessel (RPV) lower head, are investigated. The objective of this work is the assessment of severe accident management procedures for VVER-1000 reactors, i.e. the estimation of the maximum period of time available for taking appropriate measures and particular decisions by the plant personnel. During high pressure severe accident sequences it is of prime importance to depressurize the primary circuit in order to allow for effective injection from the emergency core cooling systems and to avoid reactor pressure vessel failure at high pressure that could cause direct containment heating and subsequent challenge to the containment structure. Therefore different accident management measures were investigated for the in-vessel phase of a hypothetical station blackout accident using the severe accident code ASTEC, the mechanistic code ATHLET and the multi-purpose code system ANSYS. The analyses performed on the PHEBUS ISP-46 experiment, as well as simulations of small break loss of coolant accident and station blackout scenarios were used to contribute to the validation and improvement of the integral severe accident code ASTEC. Investigations on the applicability and the effectiveness of accident management procedures in the preventive domain, as well as detailed analyses on the thermal-hydraulic phenomena during the early in-vessel phase of a station blackout accident have been performed with the mechanistic code ATHLET. The results of the simulations show, that the

  12. Thermal use of challenging biomass fuels; Thermische Nutzung von anspruchsvollen Biomassebrennstoffen. Versuche Herbst 2006

    Energy Technology Data Exchange (ETDEWEB)

    Buehler, R. [Umwelt und Energie, Maschwanden (Switzerland); Hersener, J.-L. [Ingenieurbuero Hersener, Wiesendangen (Switzerland); Jenni, A. [Ardens GmbH, Liestal (Switzerland); Klippel, N. [Verenum, Zuerich (Switzerland)

    2007-10-15

    This final report for the Swiss Federal Office of Energy (SFOE) reports on experiments made in autumn 2006 on the thermal use of biomass fuels such as agricultural wastes. As a continuation of the tests performed in 2005, further tests were planned for 2006. The authors quote that for various reasons, only part of the planned test program could be carried out. Tests made with the fuel mixtures sedge and chipped wood as well as horse manure, sedge and chipped wood are reported on. The tests showed that, under optimal conditions, these fuel mixtures can be used as biomass fuel, leading to low emissions. Stable combustion conditions were, however, very difficult to achieve. Details on the tests performed and their results are presented and knowledge gained is discussed.

  13. Criticality accident:

    International Nuclear Information System (INIS)

    Canavese, Susana I.

    2000-01-01

    A criticality accident occurred at 10:35 on September 30, 1999. It occurred in a precipitation tank in a Conversion Test Building at the JCO Tokai Works site in Tokaimura (Tokai Village) in the Ibaraki Prefecture of Japan. STA provisionally rated this accident a 4 on the seven-level, logarithmic International Nuclear Event Scale (INES). The September 30, 1999 criticality accident at the JCO Tokai Works Site in Tokaimura, Japan in described in preliminary, technical detail. Information is based on preliminary presentations to technical groups by Japanese scientists and spokespersons, translations by technical and non-technical persons of technical web postings by various nuclear authorities, and English-language non-technical reports from various news media and nuclear-interest groups. (author)

  14. Effects of GnRHa treatment during vitellogenesis on the reproductive physiology of thermally challenged female Atlantic salmon (Salmo salar

    Directory of Open Access Journals (Sweden)

    Kelli Anderson

    2017-10-01

    Full Text Available Tasmanian Atlantic salmon (S. salar broodstock can experience temperatures above 20 °C, which impairs reproductive development and inhibits ovulation. The present study investigated the prolonged use of gonadotropin releasing hormone analogue (GnRHa during vitellogenesis as a means of maintaining endocrine function and promoting egg quality at elevated temperature in maiden and repeat spawning S. salar. GnRHa-treatment during vitellogenesis did not compensate for the negative effects of thermal challenge on the timing of ovulation, egg size, egg fertility or embryo survival in any fish maintained at 22 °C relative to 14 °C. The lack of effectiveness was reflected by the endocrine data, as plasma follicle stimulating hormone and luteinising hormone levels were not different between treated and untreated groups at 22 °C. Furthermore, plasma testosterone and E2 levels were unchanged in GnRHa-treated fish at 22 °C, and plasma levels were generally lower in both groups maintained at 22 °C relative to 14 °C. Transcription of vitellogenin, and zona pellucida B and C was not enhanced in GnRHa-treated fish relative to untreated fish at 22 °C, presumably due to observed suppression of plasma E2. These results indicate that thermal impairment of reproduction is likely to occur on multiple levels, and is difficult to overcome via hormonal manipulation.

  15. 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

  16. A novel challenge test incorporating irradiation (60Co) of compost sub-samples to validate thermal lethality towards pathogenic bacteria.

    Science.gov (United States)

    Moore, John E; Watabe, Miyuki; Stewart, Andrew; Cherie Millar, B; Rao, Juluri R

    2009-01-01

    Maturing compost heaps normally attaining temperatures ranging from 55 to 65 degrees C is generally regarded to conform to recommended biological risks and sanitation standards for composts stipulated by either EU or US-EPA. Composted products derived from animal sources are further required by EU biohazard safety regulatory legislation that such composts either attain 70 degrees C for over 3h during maturation or via treatment at 70 degrees C for 1h before being considered for dispensation on land. The setting of the upper limit of thermal lethality at 70 degrees C/1h for achieving biosecurity of the animal waste composted products (e.g. pelleted fertilizer formulations) is not properly substantiated by specific validation tests, comprising a 'wipe-out' step (usually via autoclaving) followed by inoculation of a prescribed bacterium, exposure to 70 degrees C/1h and the lethality determined. Pelleted formulations of composts are not amenable for wet methods (autoclaving) for wipe-out sterilization step as this is detrimental to the pellet and compromises sample integrity. This study describes a laboratory method involving the employment of ((60)Co) irradiation 'wipe-out' step to: (a) compost sub-samples drawn from compost formulation heaps and (b) pelleted products derived from composted animal products while determining the thermal lethality of a given time/temperature (70 degrees C/1h) treatment process and by challenging the irradiated sample (not just with one bacterium but), out with 10 potential food-poisoning organisms from the bacterial genera (Campylobacter, Escherichia, Listeria, Salmonella, Yersinia) frequently detected in pig and poultry farm wastes. This challenge test on compost sub-samples can be a useful intervention ploy for 'inspection and validation' technique for composters during the compost maturity process, whose attainment of temperatures of 55-65 degrees C is presumed sufficient for attainment of sanitation. Stringent measures are further

  17. Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

    International Nuclear Information System (INIS)

    De Rosa, Felice

    2006-01-01

    In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when ΔTsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its

  18. Tchernobyl accident

    International Nuclear Information System (INIS)

    1986-06-01

    First, R.M.B.K type reactors are described. Then, safety problems are dealt with reactor control, behavior during transients, normal loss of power and behavior of the reactor in case of leak. A possible scenario of the accident of Tchernobyl is proposed: events before the explosion, possible initiators, possible scenario and events subsequent to the core meltdown (corium-concrete interaction, interaction with the groundwater table). An estimation of the source term is proposed first from the installation characteristics and the supposed scenario of the accident, and from the measurements in Europe; radiological consequences are also estimated. Radioactivity measurements (Europe, Scandinavia, Western Europe, France) are given in tables (meteorological maps and fallouts in Europe). Finally, a description of the site is given [fr

  19. Accident: Reminder

    CERN Multimedia

    2003-01-01

    There is no left turn to Point 1 from the customs, direction CERN. A terrible accident happened last week on the Route de Meyrin just outside Entrance B because traffic regulations were not respected. You are reminded that when travelling from the customs, direction CERN, turning left to Point 1 is forbidden. Access to Point 1 from the customs is only via entering CERN, going down to the roundabout and coming back up to the traffic lights at Entrance B

  20. Challenges and Enhancements to Defence-in-Depth (DiD) in Light of the Fukushima Daiichi NPP Accident. Proceedings of a Joint CNRA/CSNI Workshop, Paris, France, 5 June 2013

    International Nuclear Information System (INIS)

    Niel, Jean-Christophe; Autorite de surete nucleaire - ASN; ); Echavarri, Luis; ); Schneider, Thierry; ); Borchardt, Bill; Wanner, Hans; Park, Youn Won; Glenn, Tracy; Regaldo, Jacques; Fuketa, Toyoshi; Repussard, Jacques; Bolshov, Leonid; Chevet, Pierre Frank

    2014-01-01

    An international Workshop on Challenges and Enhancements to Defence in Depth (DiD) in Light of the Fukushima Daiichi Accident was jointly organised by the NEA Committee on Nuclear Regulatory Activities (CNRA) and the NEA Committee on the Safety on Nuclear Installations (CSNI), with input from the NEA Committee on Radiation Protection and Public Health (CRPPH), on 5 June 2013 in Paris. About 100 participants from NEA member countries, India, the International Atomic Energy Agency (IAEA), the World Association of Nuclear Operators (WANO) and Eurelectric held in-depth discussions on the defence-in-depth concept and its implementation in the post-Fukushima context. They also considered additional steps to be taken at the national and international levels to address the challenges identified and to make further enhancements to nuclear safety, along with future NEA activities in support of these processes. The concept of DiD - of multiple levels of protection - has been developed and refined by the nuclear safety community over many years. The concept is based on the experience and practice of high hazard industry in general as well as developments within the nuclear industry. The accident at the Fukushima Daiichi Nuclear Power Station in Japan on the 11 March 2011 demonstrated the importance of the concept of DiD, how these multiple levels of protection can operate and how some can be challenged. Just as with the nuclear accidents at Three Mile Island and Chernobyl and accidents from other industries (e.g. chemical, aerospace, oil and gas), the NEA felt it important that lessons learned are used to further develop the concept and implementation of Defence-in-Depth to help ensure and enhance the safe operation of nuclear power plants around the world. The NEA has therefore looked at the concept and implementation of Defence-in-Depth and possible challenges and enhancements. It was noted that a great deal of interest has been shown in this area, particularly in the

  1. Early life thermal stress: Impact on future thermotolerance, stress response, behavior, and intestinal morphology in piglets exposed to a heat stress challenge during simulated transport

    Science.gov (United States)

    Study objectives were to evaluate the impact of early life thermal stress (ELTS) on thermoregulation, stress, and intestinal health of piglets subjected to a future heat stress (HS) challenge during simulated transport. Approximately 7 d after farrowing, 12 first parity gilts and their litters were ...

  2. Severe accident training simulator APROS SA

    International Nuclear Information System (INIS)

    Raiko, Eerikki; Salminen, Kai; Lundstroem, Petra; Harti, Mika; Routamo, Tomi

    2003-01-01

    APROS SA is a severe accident training simulator based on the APROS simulation environment. APROS SA has been developed in Fortum Nuclear Services Ltd to serve as a training tool for the personnel of the Loviisa NPP. Training with APROS SA gives the personnel a deeper understanding of the severe accident phenomena and thus it is an important part of the implementation of the severe accident management strategy. APROS SA consists of two parts, a comprehensive Loviisa plant model and an external severe accident model. The external model is an extension to the Loviisa plant model, which allows the simulation to proceed into the severe accident phase. The severe accident model has three submodels: the core melting and relocation model, corium pool model and fission product model. In addition to these, a new thermal-hydraulic solver is introduced to the core region of the Loviisa plant model to replace the more limited APROS thermal-hydraulic solver. The full APROS SA training simulator has a graphical user interface with visualizations of both severe accident management panels at the operator room and the important physical phenomena during the accident. This paper describes the background of the APROS SA training simulator, the severe accident submodels and the graphical user interface. A short description how APROS SA will be used as a training tool at the Loviisa NPP is also given

  3. 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...

  4. Severe accident development modeling and evaluation for CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Negut, Gheorghe [National Agency for Radioactive Waste, 1, Campului Str., 115400 Mioveni (Romania)], E-mail: gheorghe.negut@andrad.ro; Catana, Alexandru [Institute for Nuclear Research Pitesti, 1, Campului Str., Mioveni P.O. Box 78, 0300 Pitesti (Romania); Prisecaru, Ilie; Dupleac, Daniel [Politehnica University Bucharest, 313, Splaiul Independentei, Sect. 6, 060042 Bucharest (Romania)

    2009-09-15

    Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents. Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.

  5. Severe accident development modeling and evaluation for CANDU

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2009-01-01

    Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents. Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.

  6. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Estimation of trigger condition for vapor explosion. JAERI's nuclear research promotion program, H10-027-1. Contract research

    International Nuclear Information System (INIS)

    Nariai, Hideki

    2002-03-01

    The experimental and analytical researches were conducted to study melted core material and coolant interaction including solidification and vapor explosion which is one of the most unidentified thermal hydraulic phenomena during sever accident of nuclear reactor. At first, the effect of the material properties on vapor explosion and solidification was examined to clarify the dominant factors for the spontaneous vapor explosion. Next, the interfacial phenomena of the high temperature melt material and violent boiling behavior of water at the interface was visually observed in the experiment. The interfacial phenomena were physically modeled. Finally, trigger phenomena from liquid-liquid contact to atomization were clarified through the forced collapse experiment of vapor film around a molten droplet by using pressure wave generation device. It is indicated by applying the results obtained in the present study to the actual reactor conditions that the possibility of the vapor explosion is extremely unlikely in the actual reactor accident sequence, since the surface of the molten uranium oxide is solidified in the water and the liquid-liquid contact can not be achieved. It should be noted that the decrease of the solidified temperature by metal compounds and the increase of the molten core temperature. (author)

  7. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Estimation of trigger condition for vapor explosion. JAERI's nuclear research promotion program, H10-027-1. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Nariai, Hideki [Tsukuba Univ., Institute of Engineering Mechanics and Systems, Tsukuba, Ibaraki (Japan)

    2002-03-01

    The experimental and analytical researches were conducted to study melted core material and coolant interaction including solidification and vapor explosion which is one of the most unidentified thermal hydraulic phenomena during sever accident of nuclear reactor. At first, the effect of the material properties on vapor explosion and solidification was examined to clarify the dominant factors for the spontaneous vapor explosion. Next, the interfacial phenomena of the high temperature melt material and violent boiling behavior of water at the interface was visually observed in the experiment. The interfacial phenomena were physically modeled. Finally, trigger phenomena from liquid-liquid contact to atomization were clarified through the forced collapse experiment of vapor film around a molten droplet by using pressure wave generation device. It is indicated by applying the results obtained in the present study to the actual reactor conditions that the possibility of the vapor explosion is extremely unlikely in the actual reactor accident sequence, since the surface of the molten uranium oxide is solidified in the water and the liquid-liquid contact can not be achieved. It should be noted that the decrease of the solidified temperature by metal compounds and the increase of the molten core temperature. (author)

  8. Neutronic, thermal-hydraulics and accident analysis calculations for an irradiation device to be used in the qualification process of dispersion fuels in the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas Borges; Silva, Antonio Teixeira e; Umbehaun, Pedro Ernesto; Silva, Jose Eduardo Rosa da; Conti, Thadeu das Neves; Yamaguchi, Mitsuo [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, SP (Brazil)], e-mail: douglasborgesdomingos@yahoo.com.br

    2009-07-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of an irradiation device placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U{sub 3}O{sub 8}-Al e U{sub 3}Si{sub 2}-Al dispersion fuels, LEU type (19.9% of {sup 235}U), with uranium densities of, respectively, 3.0 gU/cm{sup 3} and 4.8gU/cm{sup 3}. The fuel miniplates will be irradiated to nominal {sup 235}U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor, now in the conception phase. For the neutronic calculation, the computer code CITATION was utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation of the fuel miniplates will happen without any adverse consequence in the IEA-R1 reactor. (author)

  9. Electronic Paramagnetic Resonance of irradiated nails: challenges for a dosimetry in radiation accidents; Ressonancia Paramagnetica Eletronica de unhas irradiadas: desafios para uma dosimetria em acidentes radiologicos

    Energy Technology Data Exchange (ETDEWEB)

    Giannoni, Ricardo A., E-mail: giannoni@ird.gov.br [Instituto de Radioprotecao e Dosimetria, (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Rodrigues Junior, Orlando [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2014-07-01

    The purpose of this work is to characterize samples of human nails exposed to high doses of radiation, applying the technique of Electron Paramagnetic Resonance (EPR). The objective is to establish a dose response study that allow determine the absorbed dose by exposed individuals in situations of radiological accidents, in a retrospective form. Samples of human nails were collected and afterward irradiated with gamma radiation, and received dose of 20 Gy. The EPR measurement performed on the samples, before irradiation, permitted the signal identification of the components associated with effects caused by the mechanical stress during the fingernail cutting, the so-called mechanically induced signal (MIS). After the irradiation, different species of free radicals were identified, the so-called radiation induced signal (RIS). (author)

  10. The handling of radiation accidents

    International Nuclear Information System (INIS)

    1977-01-01

    The symposium was attended by 204 participants from 39 countries and 5 international organizations. Forty-two papers were presented in 8 sessions. The purpose of the meeting was to foster an exchange of experiences gained in establishing and exercising plans for mitigating the effects of radiation accidents and in the handling of actual accident situations. Only a small number of accidents were reported at the symposium, and this reflects the very high standards of safety that has been achieved by the nuclear industry. No accidents of radiological significance were reported to have occurred at commercial nuclear power plants. Of the accidents reported, industrial radiography continues to be the area in which most of the radiation accidents occur. The experience gained in the reported accident situations served to confirm the crucial importance of the prompt availability of medical and radiological services, particularly in the case of uptake of radioactive material, and emphasized the importance of detailed investigation into the causes of the accident in order to improve preventative measures. One of the principal themes of the symposium involved emergency procedures related to nuclear power plant accidents, and several papers defining the scope, progression and consequences of design base accidents for both thermal and fast reactor systems were presented. These were complemented by papers defining the resultant protection requirements that should be satisfied in the establishment of plans designed to mitigate the effects of the postulated accident situations. Several papers were presented describing existing emergency organizational arrangements relating both to specific nuclear power plants and to comprehensive national schemes, and a particularly informative session was devoted to the topic of training of personnel in the practical conduct of emergency arrangements. The general feeling of the participants was one of studied confidence in the competence and

  11. Determination of rare earth elements, thorium and uranium by inductively coupled plasma mass spectrometry and strontium isotopes by thermal ionization mass spectrometry in soil samples of Bryansk region contaminated due to Chernobyl accident

    International Nuclear Information System (INIS)

    Sahoo, S.K.; Yonehara, H.; Kurotaki, K.; Shiraishi, K.; Ramzaev, V.; Barkovski, A.

    2001-01-01

    Inductively coupled plasma mass spectrometric (ICP-MS) determination of rare earth elements (REEs), thorium and uranium in forest, pasture, field and kitchen garden soils from a Russian territory and in certified reference materials (JLK-1, JSD-2 and BCR-1) is described. In addition to concentration data, strontium isotopic composition of the soil samples were measured by thermal ionization mass spectrometry. The measurements contributed to the understanding of the background levels of these elements in an area contaminated due to Chernobyl accident. There was not a significant variation in the concentration of REEs at different depth levels in forest soil samples, however, the ratio of Th/U varied from 3.32 to 3.60. Though concentration of U and Th varied to some extent, the ratio did not show much variation. The value of 87 Sr/ 86 Sr ratio, was in the top layer soil sample relatively higher than in the lower layers. (author)

  12. Comparison between MARCH-3 and MAAP-3 thermal-hydraulic results for a severe accident in a BWR system with MARK-III containment

    International Nuclear Information System (INIS)

    Barbucci, P.; Guidi, L.; Mariotti, G.

    1988-01-01

    A comparison between results provided by the Source Term Code Package and by the MAAP-3 code for a PWR with full pressure containment was presented. Thereafter the same two methodologies were used to analyse a severe accident sequence in a typical BWR power plant equipped with a General Electric BWR 6 reactor, rated at 2894 MWt, and a MARK-III type containment. As a reference sequence the TQUV was chosen. This sequence is characterized by a transient (T) with loss of feedwater (Q) and loss of all Emergency Core Cooling Systems, both at high pressure (U) and, after the intervention of the Automatic Depressurization System (ADS), at low pressure (V). After the vessel, failure two basic scenarios for the containment response were analysed: in the first one the pedestal is always dry, in the second one it is fully flooded. Typical limestone/common sand and basaltic concrete compositions were considered. In the following sections the obtained results will be shown with the main purpose to point out the different phenomenological models of the two codes rather than to look for the true plant response to such a severe accident. After the presentation of the most important physical models and of the main assumptions for the analyses (sects. 2 and 4), the comparison will be performed for the in-vessel phase, in section 3, and for the ex-vessel phase, in section 5

  13. Radiation accidents

    International Nuclear Information System (INIS)

    Saenger, E.L.

    1986-01-01

    It is essential that emergency physicians understand ways to manage patients contaminated by radioactive materials and/or exposed to external radiation sources. Contamination accidents require careful surveys to identify the metabolic pathway of the radionuclides to guide prognosis and treatment. The level of treatment required will depend on careful surveys and meticulous decontamination. There is no specific therapy for the acute radiation syndrome. Prophylactic antibodies are desirable. For severely exposed patients treatment is similar to the supportive care given to patients undergoing organ transplantation. For high-dose extremity injury, no methods have been developed to reverse the fibrosing endarteritis that eventually leads to tissue death so frequently found with this type of injury. Although the Three Mile Island episode of March 1979 created tremendous public concern, there were no radiation injuries. The contamination outside the reactor building and the release of radioiodine were negligible. The accidental fuel element meltdown at Chernobyl, USSR, resulted in many cases of acute radiation syndrome. More than 100,000 people were exposed to high levels of radioactive fallout. The general principles outlined here are applicable to accidents of that degree of severity

  14. Chernobyl accident

    International Nuclear Information System (INIS)

    Bar'yakhtar, V.G.

    1995-01-01

    The monograph contains the catastrophe's events chronology, the efficiency assessed of those measures assumed for their localization as well as their environmental and socio-economic impact. Among materials of the monograph the results are presented of research on the radioactive contamination field forming as well as those concerning the investigation of biogeochemical properties of Chernobyl radionuclides and their migration process in the environment of the Ukraine. The data dealing with biological effects of the continued combined internal and external radioactive influence on plants, animals and human health under the circumstances of Chernobyl accident are of the special interest. In order to provide the scientific generalizing information on the medical aspects of Chernobyl catastrophe, the great part of the monograph is allotted to appraise those factors affecting the health of different population groups as well as to depict clinic aspects of Chernobyl events and medico-sanitarian help system. The National Programme of Ukraine for the accident consequences elimination and population social protection assuring for the years 1986-1993 and this Programme concept for the period up to the year 2000 with a special regard of the world community participation there

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

    OpenAIRE

    Itamar Iliuk; José Manoel Balthazar; Ângelo Marcelo Tusset; José Roberto Castilho Piqueira

    2016-01-01

    Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was prop...

  16. A study on the effect of fluidic device installed in a safety injection tank on thermal-hydraulic phenomena of large break loss of coolant accident

    International Nuclear Information System (INIS)

    Chung, Young Jong; Bae, Kyoo Hwan; Song, Jin Ho; Sim, Suk Ku; Park, Jong Kyun

    1999-03-01

    The performance of the Safety Injection Tank (SIT) with fluidic device (advanced SIT) is analyzed for the large break loss of coolant accident (LBLOCA) using RELAP5/MOD3.1-KREM. First the case is analyzed using the conventional SIT. Among various cases the case with 4-split downcomer, discharge coefficient Cd=0.6, MCP trip with reactor trip and break location of cold leg discharge side with the pressurizer is found to be the most limiting case. For the same condition, the advanced SIT results the similar PCT, however it can maintain adequately the liquid level in the downcomer. By changing the ECCS location from the current injection to the cold leg elevations, PCT is improved by 75 K. (Author). 6 refs., 4 tabs., 54 figs

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

    International Nuclear Information System (INIS)

    Milovanovic, D.; Zlvkovic, Z.

    1996-01-01

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

  18. Containment loading during severe core damage accidents

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

    1984-11-01

    The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code

  19. Some computational challenges of developing efficient parallel algorithms for data-dependent computations in thermal-hydraulics supercomputer applications

    International Nuclear Information System (INIS)

    Woodruff, S.B.

    1994-01-01

    The Transient Reactor Analysis Code (TRAC), which features a two-fluid treatment of thermal-hydraulics, is designed to model transients in water reactors and related facilities. One of the major computational costs associated with TRAC and similar codes is calculating constitutive coefficients. Although the formulations for these coefficients are local, the costs are flow-regime- or data-dependent; i.e., the computations needed for a given spatial node often vary widely as a function of time. Consequently, a fixed, uniform assignment of nodes to prallel processors will result in degraded computational efficiency due to the poor load balancing. A standard method for treating data-dependent models on vector architectures has been to use gather operations (or indirect adressing) to sort the nodes into subsets that (temporarily) share a common computational model. However, this method is not effective on distributed memory data parallel architectures, where indirect adressing involves expensive communication overhead. Another serious problem with this method involves software engineering challenges in the areas of maintainability and extensibility. For example, an implementation that was hand-tuned to achieve good computational efficiency would have to be rewritten whenever the decision tree governing the sorting was modified. Using an example based on the calculation of the wall-to-liquid and wall-to-vapor heat-transfer coefficients for three nonboiling flow regimes, we describe how the use of the Fortran 90 WHERE construct and automatic inlining of functions can be used to ameliorate this problem while improving both efficiency and software engineering. Unfortunately, a general automatic solution to the load-balancing problem associated with data-dependent computations is not yet available for massively parallel architectures. We discuss why developers should either wait for such solutions or consider alternative numerical algorithms, such as a neural network

  20. Chernobyl accident

    International Nuclear Information System (INIS)

    Capra, D.; Facchini, U.; Gianelle, V.; Ravasini, G.; Bacci, P.

    1988-01-01

    The radioactive cloud released during the Chernobyl accident reached the Padana plain and Lombardy in the night of April 30th 1986; the cloud remained in the northern Italian skies for a few days and then disappeared either dispersed by winds and washed by rains. The evidence in atmosphere of radionuclides as Tellurium, Iodine, Cesium, was promptly observed. The intense rain, in first week of may, washed the radioactivity and fall-out contamined the land, soil, grass. The present work concerns the overall contamination of the Northern Italy territory and in particular the radioactive fall-out in the Lakes region. Samples of soil have been measured at the gamma spectroscope; a correlation is found between the radionuclides concentration in soil samples and the rain intensity, when appropriate deposition models are considered. A number of measurements has been done on the Como'lake ecosystem: sediments, plankton, fishes and the overall fall-out in the area has been investigated

  1. 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....

  2. Accident Assessment

    International Nuclear Information System (INIS)

    Tripputi, Ivo; Lund, Ingemar

    2002-01-01

    There is a general feeling that decommissioning is an activity involving limited risks, compared to NPP operation, and in particular risks involving the general public. This is technically confirmed by licensing analysis and evaluations, where, once the spent fuel has been removed from the plant, the radioactivity inventory available to be released to the environment is very limited. Decommissioning activities performed so far in the world have also confirmed the first assumptions and no specific issue has been identified, in this field, to justify a completely new approach. Commercial interests in international harmonization, which could drive an in-depth discussion about the bases of this approach, are weak at the moment. However, there are several reasons why a discussion in an international framework about the Safety Case for decommissioning (and, in particular, about Accident Assessment) may be considered necessary and important, and why it may show some specific and peculiar aspects. An effort for a comprehensive and systematic D and D accident safety assessment of the decommissioning process is justified. It is necessary also to explore in a holistic way the aspects of industrial safety, and develop tools for the decision-making process optimization. The expected results are the implementation of appropriate and optimized protective measures in any event and of adequate on/off-site emergency plans for optimal public and workers protection. The experience from other decommissioning projects and large-scale industrial activities is essential to balance provisions and an Operating Experience review process (specific for decommissioning) should help to focus on real issues

  3. 1. Safety in operation: The challenges and the fields of action. 2. Incident and accident management in PWR plants ''The French approach''

    International Nuclear Information System (INIS)

    Debes, M.

    1992-01-01

    The French standardization pressurized water reactor facilities currently consist of 51 standardized units, 34 of 900 MW and 17 of 1300 MW. For EDF, safety and quality in operation constitute a unique challenge: An economic challenge: the nuclear facilities account for 75% of the power generated in France and ensure the competitiveness of its price, both for domestic consumers and on the international markets, an environmental challenge: both to limit actual and potential radioactive releases associated with operation of the nuclear facilities, and to avoid releases of combustion gases associated with the use of fossil fuels, a public relations challenge: to improve the image and the confidence of the public associated with nuclear generation. Safety in operation dependence primarily on the quality of the design and construction of the nuclear units. But it also depends on technical, human and organizational factors specific to the operation of the units, which play an essential role in maintaining and increasing operating safety. After a brief review of the safety results of the facilities and of the main themes which contribute to safety in operation, a description is given of the principal fields involved and of the main current orientations

  4. PWR pressure vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.

    1981-01-01

    Pressurized water reactors are susceptible to certain types of hypothetical accidents that under some circumstances, including operation of the reactor beyond a critical time in its life, could result in failure of the pressure vessel as a result of propagation of crack-like defects in the vessel wall. The accidents of concern are those that result in thermal shock to the vessel while the vessel is subjected to internal pressure. Such accidents, referred to as pressurized thermal shock or overcooling accidents (OCA), include a steamline break, small-break LOCA, turbine trip followed by stuck-open bypass valves, the 1978 Rancho Seco and the TMI accidents and many other postulated and actual accidents. The source of cold water for the thermal shock is either emergency core coolant or the normal primary-system coolant. ORNL performed fracture-mechanics calculations for a steamline break in 1978 and for a turbine-trip case in 1980 and concluded on the basis of the results that many more such calculations would be required. To meet the expected demand in a realistic way a computer code, OCA-I, was developed that accepts primary-system temperature and pressure transients as input and then performs one-dimensional thermal and stress analyses for the wall and a corresponding fracture-mechanics analysis for a long axial flaw. The code is briefly described, and its use in both generic and specific plant analyses is discussed

  5. Thermal hydraulic analyses of accidents associated with coolant leak from the primary circuit through a hole 10 mm equivalent diameter for the needs of PTS

    International Nuclear Information System (INIS)

    Krhounkova, J.; Kral, P.; Parduba, Z.

    1999-10-01

    The conservative assumptions of the analyses were oriented towards a worsening of the process with respect to the pressurized thermal shock (PTS). Four variants were treated, viz. leaks from the cold or hot leg, each at the rated power or zero power. Since the temperature of water supplied to the primary circuit by the emergency core cooling system is an important parameter with respect to a PTS, the calculations were performed by the iterative procedure: the basic thermal hydraulic calculation was performed by the RELAP5/MOD3.2.1 code which calculates the behaviour of the primary and secondary circuits, whereas the MELCOR code was used to calculate the behaviour of the parameters in the hermetic rooms. The calculation by the RELAP code was then repeated using data from the MELCOR calculations. Interventions by the reactor operators were also considered. (P.A.)

  6. Analysis framework to calibrate a numerical model to simulate the thermal test of a 1:2 scale dual purpose cask under accident conditions

    International Nuclear Information System (INIS)

    Miranda, Carlos A.J.; Libardi, Rosani M.P.; Marcelino, Sergio; Oliveira, Carlos Alberto de; Mattar Neto, Miguel

    2013-01-01

    This work describes thermal analysis framework including a 3D model and some 2D models to be performed in a 1:2 scale model of a dual-purpose cask to transport and to store spent fuel elements from research reactors to assess the behavior of the cask structure and materials when submitted to heating and drop tests. The analyses should consider all non-linearities involved like the lead phase change and thermal contacts, beside the variation of material properties with the temperature, the air inside it and the heat transfer phenomena (conduction, convection and irradiation) to reproduce the experimental results already obtained in a 1:2 model. A full 3D finite element model takes several hours to run just one analysis. To speed up the analyses to evaluate the significance of some parameters like the emissivity, contact resistance and heat transfer phenomena, among others, two 2D models are planned: one simulating a vertical cut by a diametral plane and another one simulating a horizontal cut by a plane at the cask half height. These 2D models are predicted to run fast enough to allow several analyses in a short period of time and to define options and the best parameters values to match the already obtained experimental results. As this thermal test can not be extrapolated to an 1:1 scale, these parameter values will be used in the final 3D model analysis and also in the full scale model. (author)

  7. Accident prevention in power plants

    International Nuclear Information System (INIS)

    Steyrer, H.

    Large thermal power plants are insured to a great extent at the Industrial Injuries Insurance Institute of Instrument and Electric Engineering. Approximately 4800 employees are registered. The accident frequency according to an evaluation over 12 months lies around 79.8 per year and 1000 employees in fossil-fired power plants, around 34.1 per year and 1000 employees in nuclear power plants, as in nuclear power plants coal handling and ash removal are excluded. Injuries due to radiation were not registered. The crucial points of accidents are mechanical injuries received on solid, sharp-edged and pointed objects (fossil-fired power plants 28.6%, nuclear power plants 41.5%), stumbling, twisting or slipping (fossil-fired power plants 21.8%, nuclear power plants 19.5%) and injuries due to moving machine parts (only nuclear power plants 12.2%). However, accidents due to burns or scalds obtain with 4.2% and less a lower portion than expected. The accident statistics can explain this fact in a way that the typical power plant accident does not exist. (orig./GL) [de

  8. Thermal Reactor Safety

    International Nuclear Information System (INIS)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

  9. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  10. 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

    accident is very hazardous. If the operator initiates F and B operation properly under the combined accident including TLOFW accident, the operators can prevent the core damage. Since F and B operation is last resort to prevent core damage and necessary conditions of F and B operation are very complicated, the consequence of these events should be considered in PSA model to improve emergency response capabilities under the rare events. Dynamic PSA modeling is better to estimate the effects of heading order and timing issues. Especially, dynamic PSA can model accident sequences and estimate their probabilities through integrated, time-dependent, probabilistic and deterministic models of NPPs, based on the thermal-hydraulic processes and operator behavior in accident conditions. We will develop the dynamic PSA model for the combined accident including TLOFW accident in the further study.

  11. Structural and containment response to LMFBR accidents

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Fistedis, S.H.; Baker, L. Jr.; Stepnewski, D.D.; Peak, R.D.; Gluekler, E.L.

    1978-01-01

    The results of current developments in analysing the response of reactor structures and containment to LMFBR accidents are presented. The current status of analysis of the structural response of LMFBR's to core disruptive accidents, including head response, potential missile generation and the effects of internal structures are presented. The results of recent experiments to help clarify the thermal response of reactor structures to molten core debris are summarized, including the use of this data to calculate the response of the secondary containment. (author)

  12. 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

  13. 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.

  14. 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.

  15. 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.)

  16. Modeling of the thermal transfer inside a porous environment: application to nuclear reactors in accident situation; Modelisation du transfert thermique dans un milieu poreux: application aux reacteurs nucleaires en situation accidentelle

    Energy Technology Data Exchange (ETDEWEB)

    Rubiolo, P.R

    2000-03-01

    The purpose of this report is to simulate heat exchanges occurring by conduction, by convection and by radiating in a porous medium made up of opaque particles in a semi-transparent fluid. Usually the determination of the macroscopic equations is based on homogenization techniques, but in the case of a major accident, the complexity of the problem is so overwhelming that semi-empirical methods are used to determine macroscopic coefficients. The author develops a new method to determine these coefficients, this method is based on the calculation of different tensors: the equivalent conductivity tensor, the radiative conductivity tensor, the thermal conductivity tensor and the heat exchange coefficient (h{sub sf}) between the solid phase and the fluid one. The first chapter briefly describes energy, impulse and mass balances. In the case of the energy balance the solid phase is not supposed to be in thermal equilibrium with the liquid phase. The second chapter presents an application of the porous media method to a one-dimensional and stationary problem, this application to a simple problem gives an idea of the performance of the method. The model allowing the calculation of h{sub sf} is developed, it is a wide range model. The second chapter ends with the presentation of the model allowing the computing of the effective conductivity of fuel rods. A comparison between results given by this new method and other numeric calculations or experimental data coming from benchmarks is presented in the third chapter. This chapter ends with the simulation of a reactor core in accidental situation, 2 cases are presented: with and without the presence of water steam. (A.C.)

  17. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-15

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

  18. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    International Nuclear Information System (INIS)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-01

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis

  19. Chemical phenomena under severe accident conditions

    International Nuclear Information System (INIS)

    Powers, D.A.

    1988-01-01

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

  20. An analysis of uncertainty and of dependence on season of year of ingestion population dose arising from design basis accidents in advanced thermal reactors

    International Nuclear Information System (INIS)

    Nair, S.; Ponting, A.C.

    1985-03-01

    The results of a detailed study of ingestion collective dose from five limiting PWR design basis releases are presented, the PWR being chosen as being typical of an advanced thermal reactor for which source terms are readily available. The ingestion collective dose was calculated for a range of wind direction/weather scenarios for releases from a typical U.K. rural and a U.K. semi-urban site and scenarios identified where the ingestion pathway was of potential significance. The dependence of the ingestion collective dose for these cases on the season of year when the release occurs was investigated. An analysis was carried out of the uncertainty in the ''worst case'' ingestion calculations arising from uncertainties in foodchain input parameters. An efficient but comprehensive set of dynamic foodchain computer models was produced and the literature surveyed to produce probability distribution functions (PDF's) for all relevant independent input data items. These were used to produce output PDF's for food contamination levels and for ingestion collective dose from the five releases. Finally, the study has highlighted several areas central to ingestion collective dose assessments where the available data are inadequate. This led to the formulation of a set of future research requirements which will need to be met both to obtain a better fundamental understanding of foodchain transfer and to reduce uncertainties in ingestion collective dose assessments. (author)

  1. The effect of system modeling on the Fukushima accident evolution

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L.E.; Fontanet, J.; López, C.; Fernández, E.

    2015-07-01

    The Fukushima accident is becoming both a unique opportunity and a huge challenge for severe accident analysis. The OECD-BSAF project has articulated a good part of the modeling efforts conducted so far. Inside this project, CIEMAT has conducted forensic analyses of the Fukushima accident in units 1 through 3 with MELCOR 2.1 and it has postulated a set of accident scenarios consistent with data. Beyond specific results, sensitivity analyses on safety systems performance and prevailing boundary conditions have highlighted the need of conducting uncertainty analyses when modeling NPPs severe accident scenarios. (Author)

  2. On The Cusp of the New Spatial Challenges - The Thermal Waste Processing Plant as an Element of Urban Space

    Science.gov (United States)

    Wójtowicz-Wróbel, Agnieszka

    2017-10-01

    The goal of this paper is to answer the question about the current importance of structures associated with the thermal processing of waste within the space of Polish cities and what status can they have in the functional and spatial structure of Polish cities in the future. The construction of thermal waste processing plants in Poland is currently a new and important problem, with numerous structures of this type being built due to increasing care for the natural environment, with the introduction of legal regulations, as well as due to the possibility of obtaining large external funding for the purposes of undertaking pro-environmental spatial initiatives, etc. For this reason, the paper contains research on the increase in the number of thermal waste processing plants in Poland in recent years. The abovementioned data was compared with similar information from other European Union member states. In the group containing Polish thermal waste processing plants, research was performed regarding the stage of the construction of a plant (operating plant, plant under construction, design in a construction phase, etc.). The paper also contains a listing of the functions other than the basic form of use, which is the incineration of waste - similarly to numerous foreign examples - that the environmentally friendly waste incineration plants fulfil in Poland, dividing the additional forms of use into "hard" elements (at the design level, requiring the expansion of a building featuring new elements that are not directly associated with the basic purpose of waste processing) and soft (social, educational, promotional actions, as well as other endeavours that require human involvement, but that do not entail significant design work on the buildings itself, expanding its form of use, etc.) as well as mixed activity, which required design work, but on a relatively small scale. Research was also conducted regarding the placement of thermal waste processing plants within the

  3. Application of FFTBM to severe accidents

    International Nuclear Information System (INIS)

    Prosek, A.; Leskovar, M.

    2005-01-01

    In Europe an initiative for the reduction of uncertainties in severe accident safety issues was initiated. Generally, the error made in predicting plant behaviour is called uncertainty, while the discrepancies between measured and calculated trends related to experimental facilities are called the accuracy of the prediction. The purpose of the work is to assess the accuracy of the calculations of the severe accident International Standard Problem ISP-46 (Phebus FPT1), performed with two versions of MELCOR 1.8.5 for validation purposes. For the quantitative assessment of calculations the improved fast Fourier transform based method (FFTBM) was used with the capability to calculate time dependent code accuracy. In addition, a new measure for the indication of the time shift between the experimental and the calculated signal was proposed. The quantitative results obtained with FFTBM confirm the qualitative conclusions made during the Jozef Stefan Institute participation in ISP-46. In general good agreement of thermal-hydraulic variables and satisfactory agreement of total releases for most radionuclide classes was obtained. The quantitative FFTBM results showed that for the Phebus FPT1 severe accident experiment the accuracy of thermal-hydraulic variables calculated with the MELCOR severe accident code is close to the accuracy of thermal-hydraulic variables for design basis accident experiments calculated with best-estimate system codes. (author)

  4. 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

  5. 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

  6. Supervisor's accident investigation handbook

    International Nuclear Information System (INIS)

    1980-02-01

    This pamphlet was prepared by the Environmental Health and Safety Department (EH and S) of Lawrence Berkeley Laboratory (LBL) to provide LBL supervisors with a handy reference to LBL's accident investigation program. The publication supplements the Accident and Emergencies section of LBL's Regulations and Procedures Manual, Pub. 201. The present guide discusses only accidents that are to be investigated by the supervisor. These accidents are classified as Type C by the Department of Energy (DOE) and include most occupational injuries and illnesses, government motor-vehicle accidents, and property damages of less than $50,000

  7. Actions taken in response to Fukushima : approaches and challenges

    International Nuclear Information System (INIS)

    Dermarkar, F.

    2012-01-01

    This paper discusses actions taken to Mitigating a Beyond Design Basis (BDB) external event and approaches for preventing a BDB Accident. It outlines the challenges in mitigating a BDB Accident and approaches for preventing fuel failure, protecting containment, and integrating BDB Accident Management Guides into existing Accident Management Procedures. Other challenges in BDB External Events are seismic and flooding risks, high wind assessment BDB Accidents (BDBA) and effectiveness of human response under BDBA conditions.

  8. HTR-10 severe accident management

    International Nuclear Information System (INIS)

    Xu Yuanhui; Sun Yuliang

    1997-01-01

    The High Temperature Gas-cooled Reactor (HTR-10) is under construction at the Institute of Nuclear Energy Technology site northwest of Beijing. This 10 MW thermal plant utilizes a pebble bed high temperature gas cooled reactor for a large range of applications such as electricity generation, steam and district heat generation, gas turbine and steam turbine combined cycle and process heat for methane reforming. The HTR-10 is the first high temperature gas cooled reactor to be licensed in China. This paper describes the safety characteristics and design criteria for the HTR-10 as well as the accident management and analysis required for the licensing process. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-15

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

  10. Characterization of SiC–SiC composites for accident tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Deck, C.P., E-mail: Christian.Deck@ga.com; Jacobsen, G.M.; Sheeder, J.; Gutierrez, O.; Zhang, J.; Stone, J.; Khalifa, H.E.; Back, C.A.

    2015-11-15

    Silicon carbide (SiC) is being investigated for accident tolerant fuel cladding applications due to its high temperature strength, exceptional stability under irradiation, and reduced oxidation compared to Zircaloy under accident conditions. An engineered cladding design combining monolithic SiC and SiC–SiC composite layers could offer a tough, hermetic structure to provide improved performance and safety, with a failure rate comparable to current Zircaloy cladding. Modeling and design efforts require a thorough understanding of the properties and structure of SiC-based cladding. Furthermore, both fabrication and characterization of long, thin-walled SiC–SiC tubes to meet application requirements are challenging. In this work, mechanical and thermal properties of unirradiated, as-fabricated SiC-based cladding structures were measured, and permeability and dimensional control were assessed. In order to account for the tubular geometry of the cladding designs, development and modification of several characterization methods were required.

  11. Some computational challenges of developing efficient parallel algorithms for data-dependent computations in thermal-hydraulics supercomputer applications

    International Nuclear Information System (INIS)

    Woodruff, S.B.

    1992-01-01

    The Transient Reactor Analysis Code (TRAC), which features a two- fluid treatment of thermal-hydraulics, is designed to model transients in water reactors and related facilities. One of the major computational costs associated with TRAC and similar codes is calculating constitutive coefficients. Although the formulations for these coefficients are local the costs are flow-regime- or data-dependent; i.e., the computations needed for a given spatial node often vary widely as a function of time. Consequently, poor load balancing will degrade efficiency on either vector or data parallel architectures when the data are organized according to spatial location. Unfortunately, a general automatic solution to the load-balancing problem associated with data-dependent computations is not yet available for massively parallel architectures. This document discusses why developers algorithms, such as a neural net representation, that do not exhibit algorithms, such as a neural net representation, that do not exhibit load-balancing problems

  12. Visualization of Traffic Accidents

    Science.gov (United States)

    Wang, Jie; Shen, Yuzhong; Khattak, Asad

    2010-01-01

    Traffic accidents have tremendous impact on society. Annually approximately 6.4 million vehicle accidents are reported by police in the US and nearly half of them result in catastrophic injuries. Visualizations of traffic accidents using geographic information systems (GIS) greatly facilitate handling and analysis of traffic accidents in many aspects. Environmental Systems Research Institute (ESRI), Inc. is the world leader in GIS research and development. ArcGIS, a software package developed by ESRI, has the capabilities to display events associated with a road network, such as accident locations, and pavement quality. But when event locations related to a road network are processed, the existing algorithm used by ArcGIS does not utilize all the information related to the routes of the road network and produces erroneous visualization results of event locations. This software bug causes serious problems for applications in which accurate location information is critical for emergency responses, such as traffic accidents. This paper aims to address this problem and proposes an improved method that utilizes all relevant information of traffic accidents, namely, route number, direction, and mile post, and extracts correct event locations for accurate traffic accident visualization and analysis. The proposed method generates a new shape file for traffic accidents and displays them on top of the existing road network in ArcGIS. Visualization of traffic accidents along Hampton Roads Bridge Tunnel is included to demonstrate the effectiveness of the proposed method.

  13. 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

  14. 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

  15. 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

  16. Bayes classifiers for imbalanced traffic accidents datasets.

    Science.gov (United States)

    Mujalli, Randa Oqab; López, Griselda; Garach, Laura

    2016-03-01

    Traffic accidents data sets are usually imbalanced, where the number of instances classified under the killed or severe injuries class (minority) is much lower than those classified under the slight injuries class (majority). This, however, supposes a challenging problem for classification algorithms and may cause obtaining a model that well cover the slight injuries instances whereas the killed or severe injuries instances are misclassified frequently. Based on traffic accidents data collected on urban and suburban roads in Jordan for three years (2009-2011); three different data balancing techniques were used: under-sampling which removes some instances of the majority class, oversampling which creates new instances of the minority class and a mix technique that combines both. In addition, different Bayes classifiers were compared for the different imbalanced and balanced data sets: Averaged One-Dependence Estimators, Weightily Average One-Dependence Estimators, and Bayesian networks in order to identify factors that affect the severity of an accident. The results indicated that using the balanced data sets, especially those created using oversampling techniques, with Bayesian networks improved classifying a traffic accident according to its severity and reduced the misclassification of killed and severe injuries instances. On the other hand, the following variables were found to contribute to the occurrence of a killed causality or a severe injury in a traffic accident: number of vehicles involved, accident pattern, number of directions, accident type, lighting, surface condition, and speed limit. This work, to the knowledge of the authors, is the first that aims at analyzing historical data records for traffic accidents occurring in Jordan and the first to apply balancing techniques to analyze injury severity of traffic accidents. Copyright © 2015 Elsevier Ltd. All rights reserved.

  17. The Chernobyl accident consequences

    International Nuclear Information System (INIS)

    2001-04-01

    Five teen years later, Tchernobyl remains the symbol of the greater industrial nuclear accident. To take stock on this accident, this paper proposes a chronology of the events and presents the opinion of many international and national organizations. It provides also web sites references concerning the environmental and sanitary consequences of the Tchernobyl accident, the economic actions and propositions for the nuclear safety improvement in the East Europe. (A.L.B.)

  18. Radiation, accidents, society

    International Nuclear Information System (INIS)

    1988-01-01

    This book is meant to be used as a reference book for information officers at the event of a nuclear accident. The main part is edited in alphabetical order to facilitate use under stress. The book gives a short review of the health risks of radiation, and descriptions of accidents that have occured. The index words that have been chosen for the main part of the book have been selected due to experiences in connection with incidents and accidents. (L.E.)

  19. Road traffic accident: The neglected health problem in Amhara ...

    African Journals Online (AJOL)

    Background: Road traffic accident is a major but neglected public health challenge. There is a paucity of published data on road traffic crashes in Amhara National Regional State. Objective: This study attempts to describe the main causes and consequences of road traffic accidents in the Amhara Region. Methods: This ...

  20. Reactivity insertion accident analysis

    International Nuclear Information System (INIS)

    Moreira, J.M.L.; Nakata, H.; Yorihaz, H.

    1990-04-01

    The correct prediction of postulated accidents is the fundamental requirement for the reactor licensing procedures. Accident sequences and severity of their consequences depend upon the analysis which rely on analytical tools which must be validated against known experimental results. Present work presents a systematic approach to analyse and estimate the reactivity insertion accident sequences. The methodology is based on the CINETHICA code which solves the point-kinetics/thermohydraulic coupled equations with weighted temperature feedback. Comparison against SPERT experimental results shows good agreement for the step insertion accidents. (author) [pt

  1. Nuclear accidents and epidemiology

    International Nuclear Information System (INIS)

    1987-01-01

    A consultation on epidemiology related to the Chernobyl accident was held in Copenhagen in May 1987 as a basis for concerted action. This was followed by a joint IAEA/WHO workshop in Vienna, which reviewed appropriate methodologies for possible long-term effects of radiation following nuclear accidents. The reports of these two meetings are included in this volume, and cover the subjects: 1) Epidemiology related to the Chernobyl nuclear accident. 2) Appropriate methodologies for studying possible long-term effects of radiation on individuals exposed in a nuclear accident. Figs and tabs

  2. Accidents (FARS) (National)

    Data.gov (United States)

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

  3. Monitoring and surveillance in accident situations

    International Nuclear Information System (INIS)

    Chadwick, K.; Menzel, H.

    1993-01-01

    The Chernobyl accident, which occurred on 26 April 1986, presented major challenges to the European Community with respect to the practical and regulatory aspects of radiation protection, public information, trade -particularly in food - and international politics. The Chernobyl accident was also a major challenge to the international scientific community which had to evaluate rapidly the radiological consequences of the accident and advise on the introduction at Chernobyl, countermeasures to reduce the consequences of radioactive contamination had been conceived largely in the context of relatively small accidental releases and for application over relatively small areas. Less consideration had been given to the practical implications of applying such measures in the case of a large source and a spread over a very large area

  4. Severe accident testing of electrical penetration assemblies

    International Nuclear Information System (INIS)

    Clauss, D.B.

    1989-11-01

    This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs

  5. Strategies for the prevention and mitigation of severe accidents

    International Nuclear Information System (INIS)

    Ader, C.; Heusener, G.; Snell, V.G.

    1999-01-01

    The currently operating nuclear power plants have, in general, achieved a high level of safety, as a result of design philosophies that have emphasized concepts such as defense-in-depth. This type of an approach has resulted in plants that have robust designs and strong containments. These designs were later found to have capabilities to protect the public from severe accidents (accidents more severe than traditional design basis in which substantial damage is done to the reactor core). In spite of this high level of safety, it has also been recognized that future plants need to be designed to achieve an enhanced level of safety, in particular with respect to severe accidents. This has led both regulatory authorities and utilities to develop guidance and/or requirements to guide plant designers in achieving improved severe accident performance through prevention and mitigation. The considerable research programs initiated after the TMI-2 accident have provided a large body of technical data, analytical methods, and the expertise necessary to provide for an understanding of a range of severe accident phenomena. This understanding of the ways severe accidents can progress and challenge containments, combined with the wide use of probabilistic safety assessments, have provided designers of evolutionary water cooled reactors opportunities to develop designs that minimize the challenges to the plant and to the public from severe accidents, including the development of accident management strategies intended to further reduce the risk of severe accidents. This paper describes some of the recent progress made in the understanding of severe accidents and related safety assessment methodology and how this knowledge has supported the incorporation of features into representative evolutionary designs that will prevent or mitigate many of the severe accident challenges present in current plants. (author)

  6. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.

  7. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

    International Nuclear Information System (INIS)

    2004-01-01

    More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

  8. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.

  9. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

    International Nuclear Information System (INIS)

    2004-01-01

    More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

  10. 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...

  11. An overview of selected severe accident research and applications

    International Nuclear Information System (INIS)

    Hammersley, R.J.; Henry, R.E.

    2004-01-01

    Severe accident research is being conducted world wide by industry organizations, utilities, and regulatory agencies. As this research is disseminated, it is being applied by utilities when they perform their Individual Plant Examinations (IPEs) and consider the preparation of Accident Management programs. The research is associated with phenomenological assessments of containment challenges and associated uncertainties, severe accident codes and analysis tools, systematic evaluation processes, and accident management planning. The continued advancement of this research and its applications will significantly contribute to the enhanced safety and operation of nuclear power plants. (author)

  12. Chernobyl accident and Denmark

    International Nuclear Information System (INIS)

    1986-12-01

    The report describes the Chernobyl accident and its consequences for Denmark in particular. It was commissioned by The Secretary of State for the Environment. Volume 2 contains copies of original documents issued by Danish authorities during the first accident phase and afterwards. Evaluations, monitoring data, press releases, legislation acts etc. are included. (author)

  13. Criticality accident in Argentina

    International Nuclear Information System (INIS)

    Oliveira, A.R. de.

    1984-01-01

    A recent criticality type accident, ocurred in Argetina, is commented. Considerations about the nature of the facility where this accident took place, its genesis, type of operation carried out on the day of the event, and the medical aspects involved are done. (Author) [pt

  14. Chernobyl accident and Danmark

    International Nuclear Information System (INIS)

    1986-12-01

    The report describes the Chernobyl accident and its consequences for Denmark in particular. It was commissioned by the Secretary of State for the Environment. Volume 1 contains copies of original documents issued by Danish authorities during the first accident phase and afterwards. Evaluations, monitoring data, press releases, legislation acts etc. are included. (author)

  15. Communication and industrial accidents

    NARCIS (Netherlands)

    As, Sicco van

    2001-01-01

    This paper deals with the influence of organizational communication on safety. Accidents are actually caused by individual mistakes. However the underlying causes of accidents are often organizational. As a link between these two levels - the organizational failures and mistakes - I suggest the

  16. Chapter 6: Accidents

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2018-04-01

    Th chapter 6 presents the accidents of: 1) Stimos (Italy - May, 1975); 2) San Salvador (El Salvador - February 5, 1989); 3) Soreq (Israel - June 21, 1990); 4) Nesvizh (Belarus - October 26, 1991); 5) Illinois (USA - February, 1965); 6)Maryland (EUA - December 11, 1991); 7)Hanoi (Vietnam -November 17, 1992); 8)Fleurus (Belgium - March 11, 2006) and final remarks on accidents.

  17. Uncertainties and severe-accident management

    International Nuclear Information System (INIS)

    Kastenberg, W.E.

    1991-01-01

    Severe-accident management can be defined as the use of existing and or alternative resources, systems, and actions to prevent or mitigate a core-melt accident. Together with risk management (e.g., changes in plant operation and/or addition of equipment) and emergency planning (off-site actions), accident management provides an extension of the defense-indepth safety philosophy for severe accidents. A significant number of probabilistic safety assessments have been completed, which yield the principal plant vulnerabilities, and can be categorized as (a) dominant sequences with respect to core-melt frequency, (b) dominant sequences with respect to various risk measures, (c) dominant threats that challenge safety functions, and (d) dominant threats with respect to failure of safety systems. Severe-accident management strategies can be generically classified as (a) use of alternative resources, (b) use of alternative equipment, and (c) use of alternative actions. For each sequence/threat and each combination of strategy, there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These include (a) uncertainty in key phenomena, (b) uncertainty in operator behavior, (c) uncertainty in system availability and behavior, and (d) uncertainty in information availability (i.e., instrumentation). This paper focuses on phenomenological uncertainties associated with severe-accident management strategies

  18. Radiological accidents in medical practice

    International Nuclear Information System (INIS)

    Cardenas Herrera, Juan

    2012-01-01

    Different radiological accidents that may occur in medical practice are shown. The following topics are focused: accident statistics for medical exposure, accidental medical exposures, radiotherapy accidents and potential accidental scenarios [es

  19. Nuclear fuel in a reactor accident.

    Science.gov (United States)

    Burns, Peter C; Ewing, Rodney C; Navrotsky, Alexandra

    2012-03-09

    Nuclear accidents that lead to melting of a reactor core create heterogeneous materials containing hundreds of radionuclides, many with short half-lives. The long-lived fission products and transuranium elements within damaged fuel remain a concern for millennia. Currently, accurate fundamental models for the prediction of release rates of radionuclides from fuel, especially in contact with water, after an accident remain limited. Relatively little is known about fuel corrosion and radionuclide release under the extreme chemical, radiation, and thermal conditions during and subsequent to a nuclear accident. We review the current understanding of nuclear fuel interactions with the environment, including studies over the relatively narrow range of geochemical, hydrological, and radiation environments relevant to geological repository performance, and discuss priorities for research needed to develop future predictive models.

  20. The role of chemistry in nuclear accidents

    International Nuclear Information System (INIS)

    Johnson, C.E.; Johnson, I.

    1986-01-01

    An accurate description of the chemical state of fission products is required for quick response in assessing the impact of nuclide release during a nuclear accident. The chemical state of the fission products is certain to change in response to their local environment. More specifically, fission products released from fuel will change their composition on contact with high-temperature steam, and these changes will determine their behavior with regard to either transport, deposition, aerosol formation, or reaction with structural components. The local oxygen potential is a key parameter in establishing the chemical state of the fission products and their release and transport mechanisms. Knowledge of the relationship of this parameter and thermal hydraulics is needed for prediction of fission product behavior in degraded core accidents. The behavior of key fission products in various stages of an accident, based on experimental results and appropriate calculations founded on fundamental thermodynamic information, will be discussed

  1. Thermal-spectrum recriticality energetics

    International Nuclear Information System (INIS)

    Schwinkendorf, K.N.

    1993-12-01

    Large computer codes have been created in the past to predict the energy release in hypothetical core disruptive accidents (CDA), postulated to occur in liquid metal reactors (LMR). These codes, such as SIMMER, are highly specific to LMR designs. More recent attention has focused on thermal-spectrum criticality accidents, such as for fuel storage basins and waste tanks containing fissile material. This paper resents results from recent one-dimensional kinetics simulations, performed for a recriticality accident in a thermal spectrum. Reactivity insertion rates generally are smaller than in LMR CDAs, and the energetics generally are more benign. Parametric variation of input was performed, including reactivity insertion and initial temperature

  2. [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.

  3. Database on aircraft accidents

    International Nuclear Information System (INIS)

    Nishio, Masahide; Koriyama, Tamio

    2012-09-01

    The Reactor Safety Subcommittee in the Nuclear Safety and Preservation Committee published the report 'The criteria on assessment of probability of aircraft crash into light water reactor facilities' as the standard method for evaluating probability of aircraft crash into nuclear reactor facilities in July 2002. In response to the report, Japan Nuclear Energy Safety Organization has been collecting open information on aircraft accidents of commercial airplanes, self-defense force (SDF) airplanes and US force airplanes every year since 2003, sorting out them and developing the database of aircraft accidents for latest 20 years to evaluate probability of aircraft crash into nuclear reactor facilities. This year, the database was revised by adding aircraft accidents in 2010 to the existing database and deleting aircraft accidents in 1991 from it, resulting in development of the revised 2011 database for latest 20 years from 1991 to 2010. Furthermore, the flight information on commercial aircrafts was also collected to develop the flight database for latest 20 years from 1991 to 2010 to evaluate probability of aircraft crash into reactor facilities. The method for developing the database of aircraft accidents to evaluate probability of aircraft crash into reactor facilities is based on the report 'The criteria on assessment of probability of aircraft crash into light water reactor facilities' described above. The 2011 revised database for latest 20 years from 1991 to 2010 shows the followings. The trend of the 2011 database changes little as compared to the last year's one. (1) The data of commercial aircraft accidents is based on 'Aircraft accident investigation reports of Japan transport safety board' of Ministry of Land, Infrastructure, Transport and Tourism. 4 large fixed-wing aircraft accidents, 58 small fixed-wing aircraft accidents, 5 large bladed aircraft accidents and 114 small bladed aircraft accidents occurred. The relevant accidents for evaluating

  4. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1987-01-01

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

  5. Accidents with sulfuric acid

    Directory of Open Access Journals (Sweden)

    Rajković Miloš B.

    2006-01-01

    Full Text Available Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eighteen years (from 1988 till the beginning of 2006 are analyzed in this paper. It is very alarming data that, according to all the recorded accidents, over 1.6 million tons of sulfuric acid were exuded. Although water transport is the safest (only 16.38% of the total amount of accidents in that way 98.88% of the total amount of sulfuric acid was exuded into the environment. Human factor was the common factor in all the accidents, whether there was enough control of the production process, of reservoirs or transportation tanks or the transport was done by inadequate (old tanks, or the accidents arose from human factor (inadequate speed, lock of caution etc. The fact is that huge energy, sacrifice and courage were involved in the recovery from accidents where rescue teams and fire brigades showed great courage to prevent real environmental catastrophes and very often they lost their lives during the events. So, the phrase that sulfuric acid is a real "environmental bomb" has become clearer.

  6. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1988-01-01

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

  7. 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.

  8. Physiological benefits of being small in a changing world: responses of Coho salmon (Oncorhynchus kisutch to an acute thermal challenge and a simulated capture event.

    Directory of Open Access Journals (Sweden)

    Timothy D Clark

    Full Text Available Evidence is building to suggest that both chronic and acute warm temperature exposure, as well as other anthropogenic perturbations, may select for small adult fish within a species. To shed light on this phenomenon, we investigated physiological and anatomical attributes associated with size-specific responses to an acute thermal challenge and a fisheries capture simulation (exercise+air exposure in maturing male coho salmon (Oncorhynchus kisutch. Full-size females were included for a sex-specific comparison. A size-specific response in haematology to an acute thermal challenge (from 7 to 20 °C at 3 °C h(-1 was apparent only for plasma potassium, whereby full-size males exhibited a significant increase in comparison with smaller males ('jacks'. Full-size females exhibited an elevated blood stress response in comparison with full-size males. Metabolic recovery following exhaustive exercise at 7 °C was size-specific, with jacks regaining resting levels of metabolism at 9.3 ± 0.5 h post-exercise in comparison with 12.3 ± 0.4 h for full-size fish of both sexes. Excess post-exercise oxygen consumption scaled with body mass in male fish with an exponent of b = 1.20 ± 0.08. Jacks appeared to regain osmoregulatory homeostasis faster than full-size males, and they had higher ventilation rates at 1 h post-exercise. Peak metabolic rate during post-exercise recovery scaled with body mass with an exponent of b~1, suggesting that the slower metabolic recovery in large fish was not due to limitations in diffusive or convective oxygen transport, but that large fish simply accumulated a greater 'oxygen debt' that took longer to pay back at the size-independent peak metabolic rate of ~6 mg min(-1 kg(-1. Post-exercise recovery of plasma testosterone was faster in jacks compared with full-size males, suggesting less impairment of the maturation trajectory of smaller fish. Supporting previous studies, these findings suggest that environmental change and non

  9. 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

  10. Thermal-hydraulic criteria for the APT tungsten neutron source design

    International Nuclear Information System (INIS)

    Pasamehmetoglu, K.

    1998-03-01

    This report presents the thermal-hydraulic design criteria (THDC) developed for the tungsten neutron source (TNS). The THDC are developed for the normal operations, operational transients, and design-basis accidents. The requirements of the safety analyses are incorporated into the design criteria, consistent with the integrated safety management and the safety-by-design philosophy implemented throughout the APT design process. The phenomenology limiting the thermal-hydraulic design and the confidence level requirements for each limit are discussed. The overall philosophy of the uncertainty analyses and the confidence level requirements also are presented. Different sets of criteria are developed for normal operations, operational transients, anticipated accidents, unlikely accidents, extremely unlikely accidents, and accidents during TNS replacement. In general, the philosophy is to use the strictest criteria for the high-frequency events. The criteria is relaxed as the event frequencies become smaller. The THDC must be considered as a guide for the design philosophy and not as a hard limit. When achievable, design margins greater than those required by the THDC must be used. However, if a specific event sequence cannot meet the THDC, expensive design changes are not necessary if the single event sequence results in sufficient margin to safety criteria and does not challenge the plant availability or investment protection considerations

  11. Social impact of accidents

    International Nuclear Information System (INIS)

    Kuroda, Isao

    1997-01-01

    There is the quite big difference between technological risk and social risk feeling. Various biases of social and sensational factors on accidents must be considered to recognize this difference. 'How safe is safe enough' is the perpetual thema concerning with not only technology but also sociology. The safety goal in aircraft design and how making effort to improve the present safety status in civil jet aircrafts is discussed as an example of social risk allowance. INSAG under IAEA started to discuss the safety culture after Chernobyl nuclear power plant accident on 1986. Safety culture and risk communication are the most important procedures to relieve the social impact for accidents. (author)

  12. Management of accident risks

    International Nuclear Information System (INIS)

    Compes, P.C.

    1987-01-01

    The example of the Chernobyl accident and the statistics of the occurrence of accidents make clear the threat to humanity, if one cannot guarantee successful accident prevention in the use and distribution of the projects aimed at. The science of safety, as it is known in the Wuppertal model, makes its contribution to this vital task for the human community. It makes it necessary to create the essential dates and concepts, the methods, principles and techniques based on them and the associated instrumentation. (DG) [de

  13. Chernobyl accident and Denmark

    International Nuclear Information System (INIS)

    1986-12-01

    The report describes the Chernobyl accident and its consequences for Denmark in particular. It was commissioned by The Secretary of State for the Environment. The event at the accident site, the release and dispersal of radioactive substances into the atmosphere and over Europe, is described. A discussion of the Danish organisation for nuclear emergencies, how it was activated and adapted to the actual situation, is given. A comprehensive description of the radiological contamination in Denmark following the accident and the estimated health effects, is presented. The situation in other European countries is mentioned. (author)

  14. Management of Radioactive Waste after a Nuclear Power Plant Accident

    International Nuclear Information System (INIS)

    Strand, Per; Laurent, Gerard; Rindo, Hiroshi; Georges, Christine; Ito, Eiichiro; Yamada, Norikazu; Iablokov, Iuri; Kilochytska, Tatiana; Jefferies, Nick; Byrne, Jim; Siemann, Michael; Koganeya, Toshiyuki; Aoki, Hiroomi

    2016-01-01

    The NEA Expert Group on Fukushima Waste Management and Decommissioning R and D (EGFWMD) was established in 2014 to offer advice to the authorities in Japan on the management of large quantities of on-site waste with complex properties and to share experiences with the international community and NEA member countries on ongoing work at the Fukushima Daiichi site. The group was formed with specialists from around the world who had gained experience in waste management, radiological contamination or decommissioning and waste management R and D after the Three Mile Island and Chernobyl accidents. This report provides technical opinions and ideas from these experts on post-accident waste management and R and D at the Fukushima Daiichi site, as well as information on decommissioning challenges. Chapter 1 provides general descriptions and a short introduction to nuclear accidents or radiological contaminations; for instance the Chernobyl NPP accident, the Three Mile Island Unit 2 accident and the Windscale fire accident. Chapter 2 provides experiences on regulator-implementer interaction in both normal and abnormal situations, including after a nuclear accident. Chapter 3 provides experiences on stakeholder involvement after accidents. These two chapters focus on human aspects after an accident and provide recommendations on how to improve communication between stakeholders so as to resolve issues arising after unexpected nuclear accidents. Chapters 4, 5 and 6 provide information on technical issues related to waste management after accidents. Chapter 4 focuses on the physical and chemical nature of the waste, Chapter 5 on radiological characterisation, and Chapter 6 on waste classification and categorisation. The persons involved in waste management after an accident should address these issues as soon as possible after the accident. Chapters 7 and 8 also focus on technical issues but with a long-term perspective of the waste direction in the future. Chapter 7 relates

  15. The development of a model to study the thermal behaviour of the coolant in the blind elements of a fast sodium-cooled breeder in the case of a severe hypothetical accident during the initial phase

    International Nuclear Information System (INIS)

    Genter, G.

    1981-03-01

    The enthalpy level of the coolant is studied in the interior of gaps and special elements of a fast sodium coded breeder reactor during the initial and the final stages of a hypothetical accident. For this purpose numerical models are presented to calculate the heat transport in the special element on the basis of heat conduction and axial convection. (orig./RW) [de

  16. Severe Accident Recriticality Analyses (SARA)

    Energy Technology Data Exchange (ETDEWEB)

    Frid, W. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hoejerup, F. [Risoe National Lab. (Denmark); Lindholm, I.; Miettinen, J.; Puska, E.K. [VTT Energy, Helsinki (Finland); Nilsson, Lars [Studsvik Eco and Safety AB, Nykoeping (Sweden); Sjoevall, H. [Teoliisuuden Voima Oy (Finland)

    1999-11-01

    rate of 2000 kg/s. In most cases, however, the predicted energy deposition was much smaller, below the regulatory limits for fuel failure, but close or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated quasi steady-state power following initial power excursion was in most cases about 20 % of the nominal reactor power, according to SIMULATE-3K and APROS. RECRIT predictions were in general different in this respect with either oscillating power or power increase approaching 50 % of nominal power which in both cases resulted in fuel temperatures above the melting point as a result of insufficient cooling. Long-term containment response to recriticality was assessed through MELCOR calculations for Olkiluoto 1 plant. At stabilised reactor power of 19 % of nominal power the containment failure due to overpressurization was predicted to occur 1.3 h after recriticality, if the accident is not mitigated. The SARA studies have clearly shown the sensitivity of recriticality phenomena to thermal-hydraulic modelling, the specifics of accident scenario, such as distribution of boron-carbide, and importance of multi-dimensional kinetics for determination of local power distribution in the core. The results of the project have pointed out the importance of adequate accident management procedures to be used by reactor operators and emergency staff during recovery actions. Recommendations in this area are given in the report.

  17. Severe accident recriticality analyses (SARA)

    Energy Technology Data Exchange (ETDEWEB)

    Frid, W. E-mail: wiktor.frid@ski.se; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Nilsson, L.; Puska, E.K.; Sjoevall, H

    2001-11-01

    quasi steady-state power following initial power excursion was in most cases approximately 20% of the nominal reactor power, according to SIMULATE-3K and APROS. However, in some RECRIT cases higher power levels, approaching 50% of the nominal power, were predicted leading to fuel temperatures exceeding the melting point, as a result of insufficient cooling of the fuel. Long-term containment response to recriticality was assessed through MELCOR calculations for the Olkiluoto 1 plant. At a stabilised reactor power of 19% of nominal power, the containment failure due to overpressurisation was predicted to occur 1.3 h after recriticality, if the accident is not mitigated. The SARA studies have clearly shown the sensitivity of recriticality phenomena to thermal-hydraulic modelling, the specifics of accident scenario, such as distribution of boron-carbide, and importance of multi-dimensional kinetics for determination of local power distribution in the core. The results of the project have pointed out the importance of adequate accident management strategies to be used by reactor operators and emergency staff during recovery actions. Recommendations in this area are given in the paper.

  18. Severe accident recriticality analyses (SARA)

    International Nuclear Information System (INIS)

    Frid, W.; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Nilsson, L.; Puska, E.K.; Sjoevall, H.

    2001-01-01

    -state power following initial power excursion was in most cases approximately 20% of the nominal reactor power, according to SIMULATE-3K and APROS. However, in some RECRIT cases higher power levels, approaching 50% of the nominal power, were predicted leading to fuel temperatures exceeding the melting point, as a result of insufficient cooling of the fuel. Long-term containment response to recriticality was assessed through MELCOR calculations for the Olkiluoto 1 plant. At a stabilised reactor power of 19% of nominal power, the containment failure due to overpressurisation was predicted to occur 1.3 h after recriticality, if the accident is not mitigated. The SARA studies have clearly shown the sensitivity of recriticality phenomena to thermal-hydraulic modelling, the specifics of accident scenario, such as distribution of boron-carbide, and importance of multi-dimensional kinetics for determination of local power distribution in the core. The results of the project have pointed out the importance of adequate accident management strategies to be used by reactor operators and emergency staff during recovery actions. Recommendations in this area are given in the paper

  19. Severe Accident Recriticality Analyses (SARA)

    International Nuclear Information System (INIS)

    Frid, W.; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Puska, E.K.; Nilsson, Lars; Sjoevall, H.

    1999-11-01

    rate of 2000 kg/s. In most cases, however, the predicted energy deposition was much smaller, below the regulatory limits for fuel failure, but close or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated quasi steady-state power following initial power excursion was in most cases about 20 % of the nominal reactor power, according to SIMULATE-3K and APROS. RECRIT predictions were in general different in this respect with either oscillating power or power increase approaching 50 % of nominal power which in both cases resulted in fuel temperatures above the melting point as a result of insufficient cooling. Long-term containment response to recriticality was assessed through MELCOR calculations for Olkiluoto 1 plant. At stabilised reactor power of 19 % of nominal power the containment failure due to overpressurization was predicted to occur 1.3 h after recriticality, if the accident is not mitigated. The SARA studies have clearly shown the sensitivity of recriticality phenomena to thermal-hydraulic modelling, the specifics of accident scenario, such as distribution of boron-carbide, and importance of multi-dimensional kinetics for determination of local power distribution in the core. The results of the project have pointed out the importance of adequate accident management procedures to be used by reactor operators and emergency staff during recovery actions. Recommendations in this area are given in the report

  20. Review of progress on enhanced accident tolerant fuel

    International Nuclear Information System (INIS)

    McCoy, K.; Dunn, B.; Kochendarfer, R.

    2015-01-01

    The accident at Fukushima has resulted in renewed interest in understanding the performance of nuclear power plants under accident conditions. Part of that interest is directed toward determining how to improve the performance of fuel during an accident that involves long exposures of the fuel to high temperatures. This paper describes the method being used by AREVA to select and evaluate approaches for improving the accident tolerance of nuclear fuel. The method involves starting with a large number of approaches that might enhance accident tolerance, and reviewing how well each approach satisfies a set of engineering requirements and goals. Among the approaches investigated we have the development of fuel pellets that contain a second phase to improve thermal conductivity, the use of molybdenum alloy tubing as fuel cladding, the use of oxidation-resistant coatings to zirconium cladding, and the use of nanoparticles in the coolant to improve heat transfer

  1. Immediate medical consequences of nuclear accidents: lessons from Chernobyl

    International Nuclear Information System (INIS)

    Gale, R.P.

    1987-01-01

    The immediate medical response to the nuclear accident at the Chernobyl nuclear power station involved containment of the radioactivity and evacuation of the nearby population. The next step consisted of assessment of the radiation dose received by individuals, based on biological dosimetry, and treatment of those exposed. Medical care involved treatment of skin burns; measures to support bone marrow failure, gastrointestinal tract injury, and other organ damage (i.e., infection prophylaxis and transfusions) for those with lower radiation dose exposure; and bone marrow transplantation for those exposed to a high dose of radiation. At Chernobyl, two victims died immediately and 29 died of radiation or thermal injuries in the next three months. The remaining victims of the accident are currently well. A nuclear accident anywhere is a nuclear accident everywhere. Prevention and cooperation in response to these accidents are essential goals

  2. 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.

  3. 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

  4. Big nuclear accidents

    International Nuclear Information System (INIS)

    Marshall, W.; Billingon, D.E.; Cameron, R.F.; Curl, S.J.

    1983-09-01

    Much of the debate on the safety of nuclear power focuses on the large number of fatalities that could, in theory, be caused by extremely unlikely but just imaginable reactor accidents. This, along with the nuclear industry's inappropriate use of vocabulary during public debate, has given the general public a distorted impression of the risks of nuclear power. The paper reviews the way in which the probability and consequences of big nuclear accidents have been presented in the past and makes recommendations for the future, including the presentation of the long-term consequences of such accidents in terms of 'loss of life expectancy', 'increased chance of fatal cancer' and 'equivalent pattern of compulsory cigarette smoking'. The paper presents mathematical arguments, which show the derivation and validity of the proposed methods of presenting the consequences of imaginable big nuclear accidents. (author)

  5. Accidents in perspective

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1989-01-01

    The nuclear industry perspective and the public perspective on big nuclear accidents and leukaemia near nuclear sites are discussed. The industry perspective is that big accidents are so unlikely as to be virtually impossible and that leukaemia is not specifically associated with nuclear installations. Clusters of cancer with statistical significance occur in major cities. The public perspective is coloured by a prejudice and myth: the fear of radiation. The big nuclear accident is seen therefore as much more unacceptable than any other big accident. Risks associated with Sizewell-B nuclear station and the liquid gas depot at Canvey Island are discussed. The facts and figures are presented as tables and graphs. Given conflicting interpretations of the leukaemia problem the public inclines towards the more pessimistic view. (author)

  6. Boating Accident Statistics

    Data.gov (United States)

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

  7. Improving performance with accident tolerant-fuels

    International Nuclear Information System (INIS)

    Gomes, Daniel S.; Muniz, Rafael O.R.; Giovedi, Claudia; Universidade de São Paulo

    2017-01-01

    After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO 2 - Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density - above that supported by UO 2 - and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U 3 Si 2 , UN, and UC contain a higher uranium density and thermal conductivity than UO 2 , providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U 3 Si 2 , UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO 2 . These ATFs also have thermal conductivities approximately four times higher than that of UO 2 . A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO 2 - Zr as the fuel system of choice. (author)

  8. Occupational Accidents And Preventive Measures

    CERN Document Server

    Fassnacht, V

    2006-01-01

    This report presents the 2005 statistics concerning occupational accidents involving members of the CERN personnel and contractors' personnel. It sets out the accident frequency and severity rates and provides a breakdown of accidents by cause and injury. It also contains a summary analysis of the most serious accidents and the associated recommendations.

  9. Accidents with sulfuric acid

    OpenAIRE

    Rajković Miloš B.

    2006-01-01

    Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eigh...

  10. 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

  11. Accident and emergency management

    International Nuclear Information System (INIS)

    Andersen, V.; Moellenbach, K.; Heinonen, R.; Jakobsson, S.; Kukko, T.; Berg, Oe.; Larsen, J.S.; Westgaard, T.; Magnusson, B.; Andersson, H.; Holmstroem, C.; Brehmer, B.; Allard, R.

    1988-06-01

    There is an increasing potential for severe accidents as the industrial development tends towards large, centralised production units. In several industries this has led to the formation of large organisations which are prepared for accidents fighting and for emergency management. The functioning of these organisations critically depends upon efficient decision making and exchange of information. This project is aimed at securing and possibly improving the functionality and efficiency of the accident and emergency management by verifying, demonstrating, and validating the possible use of advanced information technology in the organisations mentioned above. With the nuclear industry in focus the project consists of five main activities: 1) The study and detailed analysis of accident and emergency scenarios based on records from incidents and rills in nuclear installations. 2) Development of a conceptual understanding of accident and emergency management with emphasis on distributed decision making, information flow, and control structure sthat are involved. 3) Development of a general experimental methodology for evaluating the effects of different kinds of decision aids and forms of organisation for emergency management systems with distributed decision making. 4) Development and test of a prototype system for a limited part of an accident and emergency organisation to demonstrate the potential use of computer and communication systems, data-base and knowledge base technology, and applications of expert systems and methods used in artificial intelligence. 5) Production of guidelines for the introduction of advanced information technology in the organisations based on evaluation and validation of the prototype system. (author)

  12. Accident management information needs

    International Nuclear Information System (INIS)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R.

    1990-04-01

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

  13. Accident management information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-04-01

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

  14. IRSN-Ancli seminar on the post-accident context

    International Nuclear Information System (INIS)

    Didier, Damien; Leroyer, Veronique; Gariel, Jean-Christophe; Meier, Christine; Petitfrere, Michael; Meraux-Netillard, Isabelle; Lerouxel, Roland; Gandouen, Gael; Boutin, Dominique; Charre, Jean-Pierre; Noe, Maite; Quenneville, Celine; Farandeau, Sebastien; Mouchet, Chantal; Pineau, Coralie; Rollinger, Francois; GARIEL, Jean-Christophe; Ando, Ryoko; Nishida, Shoshi; Miazaki, Makoto; Hayano, Ryugo; Lheureux, Yves; Lochard, Jacques; Boilley, David; Godet, Jean-Luc

    2014-10-01

    The first session addressed the context of post-accident management: main challenges of radiation protection in case of nuclear accident, management of energy situations (specific intervention plans of nuclear plants), elements of doctrine for the post-accident management of an accident. The second session addressed the preparedness of territories to post-accident management: preparation to post-accident management in the Montbeliard district, emergency and post-accidental situation (preparedness at the district scale, example of Loiret), and return on experience from the post-accident exercise in Cattenom. The third session addressed the action undertaken by the ANCCLI and IRSN for the awareness of post-accidental problematic (experiments in Saclay, Marcoule, Gravelines and Golfech, lessons learned from the pilot phase and perspectives). The last session addressed the post-accidental management of the Fukushima accident: approach of the IRSN to learn lessons from the dialogue initiative in Fukushima, round table on challenges on the long term of post-accidental management, Japanese witnesses

  15. Report of Task Group on Ex-Vessel Thermal-Hydraulics Corium/concrete interactions and combustible gas distribution in large dry containments

    International Nuclear Information System (INIS)

    1987-11-01

    The Task Group on Ex-Vessel Thermal-Hydraulics was established by the PWG 2 to address the physical processes that occur in the ex-vessel phase of severe accidents, to study their impact on containment loading and failure, and to assess the available calculation methods. This effort is part of an overall CSNI effort to come to an international understanding of the issues involved. The Task Group decided to focus its initial efforts on the Large Dry Containment used extensively to contain the consequences of postulated (design basis) accidents in Light Water Reactors (LWR). Although such containments have not been designed with explicit consideration of severe accidents, recent assessments indicate a substantial inherent capability for these accidents. The Task Group has examined the loads likely to challenge the integrity of the containment, and considered the calculation of the containment's response. This report is the outcome of this effort

  16. Database on aircraft accidents

    International Nuclear Information System (INIS)

    Nishio, Masahide; Koriyama, Tamio

    2013-11-01

    The Reactor Safety Subcommittee in the Nuclear Safety and Preservation Committee published 'The criteria on assessment of probability of aircraft crash into light water reactor facilities' as the standard method for evaluating probability of aircraft crash into nuclear reactor facilities in July 2002. In response to this issue, Japan Nuclear Energy Safety Organization has been collecting open information on aircraft accidents of commercial airplanes, self-defense force (SDF) airplanes and US force airplanes every year since 2003, sorting out them and developing the database of aircraft accidents for the latest 20 years to evaluate probability of aircraft crash into nuclear reactor facilities. In this report the database was revised by adding aircraft accidents in 2011 to the existing database and deleting aircraft accidents in 1991 from it, resulting in development of the revised 2012 database for the latest 20 years from 1992 to 2011. Furthermore, the flight information on commercial aircrafts was also collected to develop the flight database for the latest 20 years from 1992 to 2011 to evaluate probability of aircraft crash into reactor facilities. The method for developing the database of aircraft accidents to evaluate probability of aircraft crash into reactor facilities is based on the report 'The criteria on assessment of probability of aircraft crash into light water reactor facilities' described above. The 2012 revised database for the latest 20 years from 1992 to 2011 shows the followings. The trend of the 2012 database changes little as compared to the last year's report. (1) The data of commercial aircraft accidents is based on 'Aircraft accident investigation reports of Japan transport safety board' of Ministry of Land, Infrastructure, Transport and Tourism. The number of commercial aircraft accidents is 4 for large fixed-wing aircraft, 58 for small fixed-wing aircraft, 5 for large bladed aircraft and 99 for small bladed aircraft. The relevant accidents

  17. High enrichment to low enrichment core's conversion. Accidents analysis

    International Nuclear Information System (INIS)

    Abbate, P.; Rubio, R.; Doval, A.; Lovotti, O.

    1990-01-01

    This work analyzes the different accidents that may occur in the reactor's facility after the 20% high-enriched uranium core's conversion. The reactor (of 5 thermal Mw), built in the 50's and 60's, is of the 'swimming pool' type, with light water and fuel elements of the curve plates MTR type, enriched at 93.15 %. This analysis includes: a) accidents by reactivity insertion; b) accidents by coolant loss; c) analysis by flow loss and d) fission products release. (Author) [es

  18. Severe accident management. Optimized guidelines and strategies

    International Nuclear Information System (INIS)

    Braun, Matthias; Löffler, Micha; Plank, Hermann; Asse, Dietmar; Dimmelmeier, Harald

    2014-01-01

    The highest priority for mitigating the consequences of a severe accident with core melt lies in securing containment integrity, as this represents the last barrier against fission product release to the environment. Containment integrity is endangered by several physical phenomena, especially highly transient phenomena following high-pressure reactor pressure vessel failure (like direct containment heating or steam explosions which can lead to early containment failure), hydrogen combustion, quasi-static over-pressure, temperature failure of penetrations, and basemat penetration by core melt. Each of these challenges can be counteracted by dedicated severe accident mitigation hardware, like dedicated primary circuit depressurization valves, hydrogen recombiners or igniters, filtered containment venting, containment cooling systems, and core melt stabilization systems (if available). However, besides their main safety function these systems often have also secondary effects that need to be considered. Filtered containment venting causes (though limited) fission product release into the environment, primary circuit depressurization leads to loss of coolant, and an ex-vessel core melt stabilization system as well as hydrogen igniters can generate high pressure and temperature loads on the containment. To ensure that during a severe accident any available systems are used to their full beneficial extent while minimizing their potential negative impact, AREVA has implemented a severe accident management for German nuclear power plants. This concept makes use of extensive numerical simulations of the entire plant, quantifying the impact of system activations (operational systems, safety systems, as well as dedicated severe accident systems) on the accident progression for various scenarios. Based on the knowledge gained, a handbook has been developed, allowing the plant operators to understand the current state of the plant (supported by computational aids), to predict

  19. Consequence analysis of core damage states following severe accidents for the CANDU reactor design

    International Nuclear Information System (INIS)

    Wahba, N.N.; Kim, Y.T.; Lie, S.G.

    1997-01-01

    The analytical methodology used to evaluate severe accident sequences is described. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression and source term estimate are summarized. The postulated sever accidents analyzed, in general, mainly differ in the timing to reach and progress through each defined c ore damage state . This paper presents the methodology and results of the timing and steam discharge calculations as well as source term estimate out of containment for accident sequences classified as potentially leading to core disassembly following a small break loss-of-coolant accident (LOCA) scenario as a specific example. (author)

  20. Historical aspects of radiation accidents

    International Nuclear Information System (INIS)

    Mettler, F.A. Jr.; Ricks, R.C.

    1990-01-01

    Radiation accidents are extremely rare events; however, the last two years have witnessed the largest radiation accidents in both the eastern and western hemispheres. It is the purpose of this chapter to review how radiation accidents are categorized, examine the temporal changes in frequency and severity, give illustrative examples of several types of radiation accidents, and finally, to describe the various registries for radiation accidents

  1. Accidents in nuclear ships

    Energy Technology Data Exchange (ETDEWEB)

    Oelgaard, P L [Risoe National Lab., Roskilde (Denmark); [Technical Univ. of Denmark, Lyngby (Denmark)

    1996-12-01

    This report starts with a discussion of the types of nuclear vessels accidents, in particular accidents which involve the nuclear propulsion systems. Next available information on 61 reported nuclear ship events in considered. Of these 6 deals with U.S. ships, 54 with USSR ships and 1 with a French ship. The ships are in almost all cases nuclear submarines. Only events that involve the sinking of vessels, the nuclear propulsion plants, radiation exposures, fires/explosions, sea-water leaks into the submarines and sinking of vessels are considered. For each event a summary of available information is presented, and comments are added. In some cases the available information is not credible, and these events are neglected. This reduces the number of events to 5 U.S. events, 35 USSR/Russian events and 1 French event. A comparison is made between the reported Soviet accidents and information available on dumped and damaged Soviet naval reactors. It seems possible to obtain good correlation between the two types of events. An analysis is made of the accident and estimates are made of the accident probabilities which are found to be of the order of 10{sup -3} per ship reactor years. It if finally pointed out that the consequences of nuclear ship accidents are fairly local and does in no way not approach the magnitude of the Chernobyl accident. It is emphasized that some of the information on which this report is based, may not be correct. Consequently some of the results of the assessments made may not be correct. (au).

  2. Accidents in nuclear ships

    International Nuclear Information System (INIS)

    Oelgaard, P.L.

    1996-12-01

    This report starts with a discussion of the types of nuclear vessels accidents, in particular accidents which involve the nuclear propulsion systems. Next available information on 61 reported nuclear ship events in considered. Of these 6 deals with U.S. ships, 54 with USSR ships and 1 with a French ship. The ships are in almost all cases nuclear submarines. Only events that involve the sinking of vessels, the nuclear propulsion plants, radiation exposures, fires/explosions, sea-water leaks into the submarines and sinking of vessels are considered. For each event a summary of available information is presented, and comments are added. In some cases the available information is not credible, and these events are neglected. This reduces the number of events to 5 U.S. events, 35 USSR/Russian events and 1 French event. A comparison is made between the reported Soviet accidents and information available on dumped and damaged Soviet naval reactors. It seems possible to obtain good correlation between the two types of events. An analysis is made of the accident and estimates are made of the accident probabilities which are found to be of the order of 10 -3 per ship reactor years. It if finally pointed out that the consequences of nuclear ship accidents are fairly local and does in no way not approach the magnitude of the Chernobyl accident. It is emphasized that some of the information on which this report is based, may not be correct. Consequently some of the results of the assessments made may not be correct. (au)

  3. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    Snell, V.G.; Howieson, J.Q.; Alikhan, S.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

    1996-01-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10 -6 /year. 95 refs, 3 tabs

  4. CANDU safety under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Snell, V G; Howieson, J Q [Atomic Energy of Canada Ltd. (Canada); Alikhan, S [New Brunswick Electric Power Commission (Canada); Frescura, G M; King, F [Ontario Hydro (Canada); Rogers, J T [Carleton Univ., Ottawa, ON (Canada); Tamm, H [Atomic Energy of Canada Ltd. (Canada). Whiteshell Research Lab.

    1996-12-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10{sup -6}/year. 95 refs, 3 tabs.

  5. School accidents in Austria.

    Science.gov (United States)

    Schalamon, Johannes; Eberl, Robert; Ainoedhofer, Herwig; Singer, Georg; Spitzer, Peter; Mayr, Johannes; Schober, Peter H; Hoellwarth, Michael E

    2007-09-01

    The aim of this study was to obtain information about the mechanisms and types of injuries in school in Austria. Children between 0 and 18 years of age presenting with injuries at the trauma outpatient in the Department of Pediatric Surgery in Graz and six participating hospitals in Austria were evaluated over a 2-year prospective survey. A total of 28,983 pediatric trauma cases were registered. Personal data, site of the accident, circumstances and mechanisms of accident and the related diagnosis were evaluated. At the Department of Pediatric Surgery in Graz 21,582 questionnaires were completed, out of which 2,148 children had school accidents (10%). The remaining 7,401 questionnaires from peripheral hospitals included 890 school accidents (12%). The male/female ratio was 3:2. In general, sport injuries were a predominant cause of severe trauma (42% severe injuries), compared with other activities in and outside of the school building (26% severe injuries). Injuries during ball-sports contributed to 44% of severe injuries. The upper extremity was most frequently injured (34%), followed by lower extremity (32%), head and neck area (26%) and injuries to thorax and abdomen (8%). Half of all school related injuries occur in children between 10 and 13 years of age. There are typical gender related mechanisms of accident: Boys get frequently injured during soccer, violence, and collisions in and outside of the school building and during craft work. Girls have the highest risk of injuries at ball sports other than soccer.

  6. Statistical modelling of the frequency and severity of road accidents

    DEFF Research Database (Denmark)

    Janstrup, Kira Hyldekær

    -reporting. The problem of under-reporting is not unique for traffic accidents as severe under-reporting is a challenge in many other fields of incident reporting. In other incidents fields with intended or unintended harm, research has investigated the behavioural reasons for why people choose to report an incident......Under-reporting of traffic accidents is a well-discussed subject in traffic safety and it is well-known that the degree of under-reporting of traffic accidents is quite high in many countries. Nevertheless, very little literature has been made to investigate what causes the high degree of under...... on the service quality within the police none have looked at the service quality specific for the handling of traffic accidents.The objective of this Ph.D. thesis is to investigate the extent of under-reporting of traffic accidents in Denmark and trace the under-reporting systematically. As something new...

  7. Development of severe accident management guidance for Younggwang units 5 and 6

    International Nuclear Information System (INIS)

    Lee, K. W.; Beon, C. S.; Kim, M. K.; Hong, S. Y.; Park, K. S.

    2001-01-01

    Severe Accident Management Guidance (SAMG) has been developed for Younggwang Units 5 and 6. It is consisted of Severe Accident Control Room Guideline, Diagnostic Flow Chart, Severe Accident Guideline, Severe Challenge Guideline, TSC Long Term Monitoring, SAMG Termination. Severe Accident Control Room Guideline, which deals with severe accident after finishing Emergency Operation Procedure, consists of acitions before and after TSC actuation. Seven servere accident management strategies are developed. Diagnostic Flow Chart, Severe Accident Guideline, and Severe Challenge Guideline are developed for each strategy, which enables the users to the implementation of strategy easily and systematically. TSC Long Term Monitoring is also developed to monitor long term activities after a particular strategy. Total of 45 set points are developed for decision making during the implementation of the SAMG

  8. Interrogations to Learn from the Fukushima Accident

    International Nuclear Information System (INIS)

    Gisquet, E.; Jeffroy, F.

    2016-01-01

    On March 11, 2011, an earthquake in eastern Japan caused the reactors in operation at the Fukushima Daiichi nuclear power plant (NPP) to trip. The emergency generators started and then suddenly failed following the tsunami. The cooling water injection system no longer worked. Suddenly plunged into total darkness, the operators had to manage the accident. Starting from the official reports and testimonies on the Fukushima accident, IRSN has conducted a survey “Human and Organizational Factors Perspective on the Fukushima Nuclear Accident.” Four years after the accident, however, as more witness accounts become available, IRSN feels it useful to return to the human and organizational response to the accident inside the NPP itself. To what extent can the participants act and coordinate their actions when faced with such a dramatic situation? To what degree did their actions contribute to the disaster? Rather than looking at the causes of the accident, this study examines the unfolding of the crisis, particularly in the most urgent early stages, and draws lessons for safety culture from the decisions and actions of key actors. The main results would be presented in three key areas: 1. How to make sense of the situation? People had to make sense of what happened and create new indicators. Since instruments and controls, as well as many communication technologies, were knocked out by the tsunami, all the standard means of determining the status of the reactors were impossible. Although they were under normal circumstances almost completely dependent on these indicators, and although (or because) their lives were most directly at risk, the operators managed this uncertainty through various means that will be successively presented. 2. What are the challenges for the emergency structure? The Emergency Response Center (ERC) operations team was responsible for being in contact with the operators in the control rooms and providing them technical support as needed. The ERC

  9. Accident analysis in research reactors

    International Nuclear Information System (INIS)

    Adorni, M.; Bousbia-salah, A.; D'Auria, F.; Hamidouche, T.

    2007-01-01

    With the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. The challenge today is to revisit the safety features of the existing nuclear plants and particularly research reactors in order to verify that the safety requirements are still met and - when necessary - to introduce some amendments not only to meet the new requirements but also to introduce new equipment from recent development of new technologies. The purpose of the present paper is to provide an overview of the accident analysis technology applied to the research reactor, with emphasis given to the capabilities of computational tools. (author)

  10. Radiation accidents and dosimetry

    International Nuclear Information System (INIS)

    Sagstuen, E.; Theisen, H.; Henriksten, T.

    1982-12-01

    On September 2nd 1982 one of the employees of the gamma-irradiation facility at Institute for Energy Technology, Kjeller, Norway entered the irradiation cell with a 65.7 kCi *sp60*Co- source in unshielded position. The victim received an unknown radiation dose and died after 13 days. Using electron spin resonance spectroscopy, the radiation dose in this accident was subsequently determined based on the production of longlived free radicals in nitroglycerol tablets borne by the operator during the accident. He used nitroglycerol for heart problems and free radical are easily formed and trapped in sugar which is the main component of the tablets. Calibration experiments were carried out and the dose given to the tablets during the accident was determined to 37.2 +- 0.5 Gy. The general use of free radicals for dose determinations is discussed. (Auth.)

  11. Big nuclear accidents

    International Nuclear Information System (INIS)

    Marshall, W.

    1983-01-01

    Much of the debate on the safety of nuclear power focuses on the large number of fatalities that could, in theory, be caused by extremely unlikely but imaginable reactor accidents. This, along with the nuclear industry's inappropriate use of vocabulary during public debate, has given the general public a distorted impression of the safety of nuclear power. The way in which the probability and consequences of big nuclear accidents have been presented in the past is reviewed and recommendations for the future are made including the presentation of the long-term consequences of such accidents in terms of 'reduction in life expectancy', 'increased chance of fatal cancer' and the equivalent pattern of compulsory cigarette smoking. (author)

  12. Care of radiation accidents

    International Nuclear Information System (INIS)

    Renz, K.

    1983-01-01

    The small probability of a serious radiation accident happening dispenses neither the plants where radiation exposure occurs nor the employers' liability insurance associations from their obligation to make provision for such cases. On the other hand, the efforts involved in such preventive measures must be kept within reasonable limits. As a result of these considerations a concept for taking care of radiation accidents was developed that is based on already existing institutions. The most attention was demanded by questions of organization, logistics, communication and information. The syndrome appearing after acute whole-body irradiation is known. This syndrome in its different stages and the relative therapeutic measures form the basis for the organization of the care of radiation accidents. (orig./MG) [de

  13. Review of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Connelly, J.W.; Storr, G.J.

    1989-01-01

    Two types of severe reactor accidents - loss of coolant or coolant flow and transient overpower (TOP) accidents - are described and compared. Accidents in research reactors are discussed. The 1961 SL1 accident in the US is used as an illustration as it incorporates the three features usually combined in a severe accident - a design flaw or flaws in the system, a circumvention of safety circuits or procedures, and gross operator error. The SL1 reactor, the reactivity accident and the following fuel-coolant interaction and steam explosion are reviewed. 3 figs

  14. Enhanced accident-tolerant fuel (EATF)

    International Nuclear Information System (INIS)

    Strumpell, John

    2013-01-01

    The Fukushima accident provided a strong reminder that the exothermic reaction between zirconium and steam, and the attendant hydrogen generation, can significantly affect the course of a severe accident. Part of the response to the accident was increased interest in the extent to which the fuel itself can mitigate the consequences of a severe accident. Improved fuel alone is not sufficient to provide the desired increase in reactor safety, but it can provide an important contribution. With support from the US Department of Energy, AREVA has brought together a team that includes researchers (AREVA, Electric Power Research Institute, Savannah River National Laboratory, University of Florida, and University of Wisconsin), a fuel vendor (AREVA), and utilities (Duke Energy and Tennessee Valley Authority). The goal of the project is to develop new technologies that can be deployed in a lead assembly within ten years. The researchers have proposed a variety of approaches for improving the performance of the fuel, including new cladding and structural materials, fuel pellets with improved thermal characteristics, and coatings on the fuel rods. The expected performance of fuels that apply these technologies will be judged against the requirements of the vendor and utilities to determine those that are most promising for immediate development and those that may be suited for development in the future. The first review will consider the manufacturability of the proposed designs; the second will focus on performance. Materials that are suitable for immediate development will be considered for irradiation in a test reactor and subsequent use in lead assembly designs

  15. 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

  16. Managing the Fukushima Challenge

    Science.gov (United States)

    Suzuki, Atsuyuki

    2014-01-01

    The Fukushima Daiichi accident raises a fundamental question: Can science and technology prevent the inevitability of serious accidents, especially those with low probabilities and high consequences? This question reminds us of a longstanding challenge with the trans-sciences, originally addressed by Alvin Weinberg well before the Three Mile Island and Chernobyl accidents. This article, revisiting Weinberg's issue, aims at gaining insights from the accident with a special emphasis on the sociotechnical or human behavioral aspects lying behind the accident's causes. In particular, an innovative method for managing the challenge is explored referring to behavioral science approaches to a decision-making process on risk management; such as managing human behavioral risks with information asymmetry, seeking a rational consensus with communicative action, and pursuing procedural rationality through interactions with the outer environment. In short, this article describes the emerging need for Japan to transform its national safety management institutions so that these might be based on interactive communication with parties inside and outside Japan. PMID:24954604

  17. Accident at Harrisburg

    International Nuclear Information System (INIS)

    1979-05-01

    The course of events during the accident on 28 March 1979 at Three Mile Island-2 Reactor at Harrisburg, Pennsylvania, is described in detail. The effects (in the environment and within the safety containment) are described. The following points are then discussed: the possibility of a comparable accident occurring in the nuclear power stations in the German Federal Republic; the possibility of any point having been overlooked in the design of nuclear power stations in the Federal Republic; whether previous risk analyses are still valid; and how near the Three Mile Island reactor was to a core meltdown. Some conclusions are drawn. (U.K.)

  18. Mortal radiological accident

    International Nuclear Information System (INIS)

    Gimenez, J.C.

    1987-01-01

    After defining the concept of 'Radiological accident', statistical data from Radiation Emergency Assistance Center of ORNL (United States of America) are given about the deaths caused by acute irradiation between 1944 and April 24, 1986 -ie, the day before Chernobyl nuclear accident- as well as on the number of deaths caused by the latter. Next the different clinical stages of the Acute Irradiation Syndrome (AIS) as well as its possible treatment are described, and finally the different physical, clinical and biological characteristics linked to the AIS and to its diagnosis and prognosis are discussed. (M.E.L.) [es

  19. The TMI-2 accident

    International Nuclear Information System (INIS)

    Loureiro, L.A.

    1986-01-01

    A critical study about the technical and man-related facts in order to establish what is considered the worst commercial nuclear power accident until 1986. Radiological consequences and stress to the public are considered in contrast to antinuclear groups. This descriptive and technical study has the purpose to document written and oral opinions obtained abroad and then explain to the public in an easy language terminology. Preliminary study describing safety related systems fails and the accident itself with minute to minute description, conduct to the consequences and then, to learned lessons

  20. 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.

  1. Correlation of Steam Generator Mixing Parameters for Severe Accident Hot-Leg Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Liao, Yehong; Guentay, Salih [Paul Scherrer Institut, Villigen PSI, CH-5232 (Switzerland)

    2008-07-01

    Steam generator inlet plenum mixing phenomenon with hot-leg counter-current natural circulation during a PWR station blackout severe accident is one of the important processes governing which component will fail first as a result of thermal challenge from the circulating gas with high temperature and pressure. Since steam generator tube failure represents bypass release of fission product from the reactor to environment, study of inlet plenum mixing parameters is important to risk analysis. Probability distribution functions of individual mixing parameter should be obtained from experiments or calculated by analysis. In order to perform sensitivity studies of the synergetic effects of all mixing parameters on the severe accident-induced steam generator tube failure, the distribution and correlation of these mixing parameters must be known to remove undue conservatism in thermal-hydraulic calculations. This paper discusses physical laws governing three mixing parameters in a steady state and setups the correlation among these mixing parameters. The correlation is then applied to obtain the distribution of one of the mixing parameters that has not been given in the previous CFD analysis. Using the distributions and considering the inter-dependence of the three mixing parameters, three sensitivity cases enveloping the mixing parameter uncertainties are recommended for the plant analysis. (authors)

  2. ADAM: An Accident Diagnostic,Analysis and Management System - Applications to Severe Accident Simulation and Management

    International Nuclear Information System (INIS)

    Zavisca, M.J.; Khatib-Rahbar, M.; Esmaili, H.; Schulz, R.

    2002-01-01

    The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM's severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric approaches for elements including (but not limited to) thermal hydraulics; heat transfer; fuel heatup, meltdown, and relocation; fission product release and transport; combustible gas generation and combustion; and core-concrete interaction. The overall model is defined by a relatively coarse spatial nodalization of the reactor coolant and containment systems and is advanced explicitly in time. The result is to enable much faster than real time (i.e., 100 to 1000 times faster than real time on a personal computer) applications to on-line investigations and/or accident management training. Other features of the simulation module include provision for activation of water injection, including the Engineered Safety Features, as well as other mechanisms for the assessment of accident management and recovery strategies and the evaluation of PSA success criteria. The accident diagnostics module of ADAM uses on-line access to selected plant parameters (as measured by plant sensors) to compute the thermodynamic state of the plant, and to predict various margins to safety (e.g., times to pressure vessel saturation and steam generator dryout). Rule-based logic is employed to classify the measured data as belonging to one of a number of likely scenarios based on symptoms, and a number of 'alarms' are generated to signal the state of the reactor and containment. This paper will address the features and limitations of ADAM with particular focus on accident simulation and management. (authors)

  3. Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Albright, D.C.; Bari, R.A.

    1976-04-01

    In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident.

  4. Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Albright, D.C.; Bari, R.A.

    1976-04-01

    In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident

  5. Traffic Accidents in Kosovo – A Major Concern for Kosovo

    Directory of Open Access Journals (Sweden)

    Merita Muharremi

    2017-03-01

    Full Text Available The number of traffic c accidents in Kosovo is increasing rapidly from year to year. As a result of traffi c accidents the number of injured and deaths has increased. During the period of January-December 2016 there were 18541 accidents, which resulted in 110 casualties (Kosovo Police Annual Reports. Campaigns undertaken by the Government of Kosovo, Ministry of Infrastructure and Transport and the Kosovo Police, as well as various video-spots intended to raise awareness of the dangers of traffic c accidents have not accomplished the expected results and did not reduce the number of accidents. Therefore, we can conclude that the number of accidents taking place in Kosovo is concerning. Despite the best efforts of relevant institutions, despite the increased engagement to pass legislation, which would be in line with the European standards, and despite substantial improvements and major investments on the infrastructure, and despite all the measures taken, reducing the number of road accidents remains a significant challenge. With this paper, I will try to draw attention to the actions, measures and activities that I consider the relevant institutions of Kosovo should focus on in order to prevent and to reduce the high number of traffic accidents.

  6. 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

  7. A methodology for the transfer of probabilities between accident severity categories

    International Nuclear Information System (INIS)

    Whitlow, J.D.; Neuhauser, K.S.

    1993-01-01

    This paper will describe a methodology which has been developed to allow accident probabilities associated with one severity category scheme to be transferred to another severity category scheme, permitting some comparisons of different studies at the category level. In this methodology, the severity category schemes to be compared are mapped onto a common set of axes. The axes represent critical accident environments (e.g., impact, thermal, crush, puncture) and indicate the range of accident parameters from zero (no accident) to the most sever credible forces. The choice of critical accident environments for the axes depends on the package being transported and the mode of transportation. The accident probabilities associated with one scheme are then transferred to the other scheme. This transfer of category probabilities is based on the relationships of the critical accident parameters to probability of occurrence. The methodology can be employed to transfer any quantity between category schemes if the appropriate supporting information is available. (J.P.N.)

  8. [Drugs and occupational accident].

    Science.gov (United States)

    Bratzke, H; Albers, C

    1996-02-01

    In a case of a fatal occupational accident (construction worker, fall from roof, urine test positive for cocaine and THC, e.g. cannabis) the question arised to what extent those drug-related occupational accidents occur. In the literature only few cases, mainly dealing with cannabis influence, have been reported, however, a higher number is suspected. Cocaine and other stimulating drugs (amphetamine) are more often used to increase physical fitness. By direct or indirect interference with vigilance these compounds may provoke accidents. Due to the lack of a legal basis proving of the influence of drugs at the working place is still very limited, although highly sensitive chemical-toxicological assay procedures are available to detect even the chronic abuse (in hair). In the general conditions of accident insurances a compensation is excluded when alcohol is involved, but drugs are not mentioned. It is indeed difficult to establish a concentration limit for drugs like that existing for alcohol (1.1%). In each case the assay of the drug involved and exact knowledge of its specific effects is in an essential prerequisite to prove the causal relationship.

  9. Note nuclear accidents combat

    International Nuclear Information System (INIS)

    1989-01-01

    In this document the starting points are described which underlie the new framework for the nuclear-accident combat in the Netherlands. All the elaboration of this is indicated in main lines. The juridical consequences of the proposed structure are enlightened and the sequel activities are indicated. (H.W.). 6 figs.; 8 tabs

  10. Measures against nuclear accidents

    International Nuclear Information System (INIS)

    1992-01-01

    A select committee appointed by the Norwegian Ministry of Social Affairs put forward proposals concerning measures for the improvement of radiation protection preparedness in Norway. On the basis on an assessment of the potential radiation accident threat, the report examines the process of response, and identifies the organizational and management factors that influence that process

  11. The Chernobyl reactor accident

    International Nuclear Information System (INIS)

    1986-01-01

    The documentation abstracted contains a complete survey of the broadcasts transmitted by the Russian wire service of the Deutsche Welle radio station between April 28 and Mai 15, 1986 on the occasion of the Chernobyl reactor accident. Access is given to extracts of the remarkable eastern and western echoes on the broadcasts of the Deutsche Welle. (HP) [de

  12. Chernobyl reactor accident: medical management

    International Nuclear Information System (INIS)

    Iyer, G.K.

    1996-01-01

    Chernobyl reactor accident on 26th April, 1986 is by far the worst radiation accident in the history of the nuclear industry. Nearly 500 plant personnel and rescue workers received doses varying from 1-16 Gy. Acute radiation syndrome (ARS) was seen only in the plant personnel. 499 individuals were screened for ARS symptoms like nausea, vomitting, diarrhoea and fever. Complete blood examination was done which showed initial granulocytosis followed by granulocytopenia and lymphocytopenia. Cytogenetic examinations were confirmatory in classifying the patients on the basis of the doses received. Two hundred and thirty seven cases of ARS were hospitalised in the first 24-36 hrs. No member of general public suffered from ARS. There were two immediate deaths and subsequently 28 died in hospital and one of the cases died due to myocardial infarction, making a total of 31 deaths. The majority of fatal cases had whole body doses of about 6 Gy, besides extensive skin burns. Two cases of radiation burns had thermal burns also. Treatment of ARS consisted of isolation, barrier nursing, replacement therapy with fluid electrolytes, platelets and RBC transfusions and antibiotic therapy for bacterial, fungal and viral infections. Bone marrow transplantations were given to 13 cases out of which 11 died due to various causes. Radiation burns due to beta, gamma radiations were seen in 56 cases and treated with dressings, surgical excision, skin grafting and amputation. Oropharangeal syndrome, producing extensive mucous in the oropharynx, was first seen in Chernobyl. The patients were treated with saline wash of the mouth. The patients who had radioactive contamination due to radioactive iodine were given stable iodine, following wash with soap, water and monitored. Fourteen survivors died subsequently due to other causes. Late health effects seen so far include excess of thyroid cancer in the children and psychological disorders due to stress. No excess leukemia has been reported so

  13. The challenge of backup power in the NPP after the accident at Fukushima, alternatives; El reto de la energia de resplado en las CC.NN. despues del accidente de Fukushima, alternativas

    Energy Technology Data Exchange (ETDEWEB)

    Contreras, C.; Gomez, A.; Gavilan, C.; Archilla, J.; Tagle, J. A.

    2014-07-01

    After the accident in Fukushima has been demonstrated the importance of having in the first hours with an independent power supply of the emergency diesel generators. In conclusion it can be said that usual (Pb batteries) storage technology is insufficient, since much more space in the rooms would be needed to achieve the required stored energy. And although just appear on the market solutions alternatives, the most advanced battery technologies still have some problems to be able to use in these applications with guarantees. A solution that arises is the combination of batteries and fuel cells. On the other hand is needed a closer collaboration with researchers and developers of batteries so that they take into account the needs of plants, so that manufactured batteries with more density energy, more power and they are more reliable. These features provide benefits in other fields besides nuclear plants. (Author)

  14. Neutron personal dosimetry in criticality accidents

    International Nuclear Information System (INIS)

    Fonseca, E.S. da; Mauricio, C.L.P.

    1996-01-01

    In the present work an innovating method is proposed to estimate the absorbed dose received by individuals irradiated with neutrons in an accident, even in the case that the victim is not using any kind of neutron dosemeter. The method combines direct measurements of 24 Na and 32 P activated in the human body. The calculation method was developed using data taken from previously published papers and experimental measurements. Other irradiations results in different neutron spectra prove the validity of the method here proposed. Using a whole body counter to measure 24 Na activity, it is possible to evaluate neutron absorbed doses in the order of 140 μ Gy of very soft (thermal) spectra. For fast neutron fields, the lower limit for neutron dose detection increases, but the present method continues to be very useful in accidents, with higher neutron doses. (author)

  15. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Behavior of fine droplet flow. JAERI's nuclear research promotion program, H10-027-7. Contract research

    International Nuclear Information System (INIS)

    Kataoka, Isao; Yoshida, Kenji; Matsuura, Keizo

    2002-03-01

    Analytical and experimental researches were carried out on the behavior of fine droplet flow in relation to the fundamental phenomena of thermohydraulics in severe accident. Simulation program of fine droplet behavior in turbulent gas flow was developed based on the eddy interaction model with improvement of Graham's stochastic model on eddy lifetime and eddy size. Furthermore, the developed program are capable of simulating the droplet behavior in annular dispersed flow based on the models of droplet entrainment from liquid film and turbulence modification of gas phase by liquid film. This program was confirmed by the various experimental data on droplet diffusion, deposition. Furthermore, this program was applied to the three dimensional droplet flow with the satisfactory agreement of experimental data. This means the developed program can be used as a simulation program for analysis of severe accident. Experimental research was carried out on the effect of liquid film on the turbulence field of gas flow in annular and annular dispersed flow. Averaged and turbulent velocity of gas phase were measured under various gas and liquid film flow rates. Turbulent velocity of gas phase in annular flow increased compared with single phase gas flow. This is due to turbulence generation by waves in liquid film. Corresponding to the turbulence modification by liquid film, distribution of averaged velocity of gas phase became flattened compared with single phase gas flow. (author)

  16. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Behavior of fine droplet flow. JAERI's nuclear research promotion program, H10-027-7. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Kataoka, Isao; Yoshida, Kenji [Osaka Univ., Graduate School of Engineering, Osaka (Japan); Matsuura, Keizo [Nuclear Fuel Industry, Co., Ltd., Tokyo (Japan)

    2002-03-01

    Analytical and experimental researches were carried out on the behavior of fine droplet flow in relation to the fundamental phenomena of thermohydraulics in severe accident. Simulation program of fine droplet behavior in turbulent gas flow was developed based on the eddy interaction model with improvement of Graham's stochastic model on eddy lifetime and eddy size. Furthermore, the developed program are capable of simulating the droplet behavior in annular dispersed flow based on the models of droplet entrainment from liquid film and turbulence modification of gas phase by liquid film. This program was confirmed by the various experimental data on droplet diffusion, deposition. Furthermore, this program was applied to the three dimensional droplet flow with the satisfactory agreement of experimental data. This means the developed program can be used as a simulation program for analysis of severe accident. Experimental research was carried out on the effect of liquid film on the turbulence field of gas flow in annular and annular dispersed flow. Averaged and turbulent velocity of gas phase were measured under various gas and liquid film flow rates. Turbulent velocity of gas phase in annular flow increased compared with single phase gas flow. This is due to turbulence generation by waves in liquid film. Corresponding to the turbulence modification by liquid film, distribution of averaged velocity of gas phase became flattened compared with single phase gas flow. (author)

  17. Computational features of the MELT-III neutronics, thermal-hydraulics computer code system

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Waltar, A.E.

    1976-01-01

    A multichannel, thermal-hydraulics, neutronic accident analysis program for simulating fast reactor behavior from a hypothetical accident inception to the start of core disassembly or to reactor shutdown is described

  18. Lessons learned from accidents investigations

    Energy Technology Data Exchange (ETDEWEB)

    Zuniga-Bello, P. [Consejo Nacional de Ciencia y Tecnologia (CONACYT), Mexico City (Mexico); Croft, J. [National Radiological Protection Board (United Kingdom); Glenn, J

    1997-12-31

    Accidents from three main practices: medical applications, industrial radiography and industrial irradiators are used to illustrate some common causes of accidents and the main lessons to be learned. A brief description of some of these accidents is given. Lessons learned from the described accidents are approached by subjects covering: safety culture, quality assurance, human factors, good engineering practice, defence in depth, security of sources, safety assessment and monitoring and verification compliance. (author)

  19. Overview of core disruptive accidents

    International Nuclear Information System (INIS)

    Marchaterre, J.F.

    1977-01-01

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

  20. Lessons learned from accident investigations

    International Nuclear Information System (INIS)

    Zuniga-Bello, P.; Croft, J.R.; Glenn, J.

    1998-01-01

    Accidents in three main practices - medical applications, industrial radiography and industrial irradiators - are used to illustrate some common causes of accidents and the main lessons to be learned from them. A brief description of some of these accidents is given. Lessons learned from the accidents described are approached bearing in mind: safety culture, quality assurance, human factors, good engineering practice, defence in depth, security of sources, safety assessment and monitoring and verification compliance. (author)

  1. Risk/benefits of fast breeder reactors societal risks and accident modelling

    International Nuclear Information System (INIS)

    Farmer, F.R.

    1979-01-01

    A special feature of the Fast Breeder Reactor is the possibility of fuel vaporisation, hence accidents may have more severe consequences than thermal reactors. This article discusses the process of accident modelling, the identification and assessment of risk, not yet incurred. 10 refs

  2. EPRI research on accident management

    International Nuclear Information System (INIS)

    Oehlberg, R.N.; Chao, J.

    1991-01-01

    The paper discusses Nuclear Regulatory Commission (NRC) efforts regarding severe reactor accident management and the Nuclear Management and Resources Council (NUMAEX), activities. (EPRI) Electric Power Research Institute accident management program consists of the two products just mentioned plus one related to severe accident plant status information and the MAAP 4.0 computer code. These are briefly discussed

  3. Accident management on french PWRS

    International Nuclear Information System (INIS)

    Queniart, D.

    1990-06-01

    After a brief recall of French safety rationale, the reactor operation and severe accident management is given. The research and development aimed at developing accident management procedures and emergency organization in France for the case of a NPP accident are also given

  4. The design of PSB-VVER experiments relevant to accident management

    International Nuclear Information System (INIS)

    Del Nevo, Alessandro; D'auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

    2008-01-01

    Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes, which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility, operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed. The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility. (author)

  5. The Design of PSB-VVER Experiments Relevant to Accident Management

    Science.gov (United States)

    Nevo, Alessandro Del; D'Auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

    Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes(1), which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility (2), operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed (3). The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility.

  6. Accident transient processes at NPPs with the WWER type reactors

    International Nuclear Information System (INIS)

    Bukrinskij, A.M.

    1982-01-01

    Thermal-physical and nuclear-physical transient processes at NPPs with the WWER type reactors during accidents with the main technological equipment failures and the accidents with loss of coolant in the primary and secondary coolant circuits are considered. Mathematical methods used for these processes modelling is described. Examples of concrete calculations for accidents with different failures are given. Comparative analysis of the results of dynamic tests at the Novo-Voronezh-3 reactor is presented. It is concluded that the modern NPP design is impossible without application of mathematical modelling methods. The mathematical modelling of transients is also necessary for proper and safe NPP operation. Mathematical modelling of accidents at NPPs is a comparatively new method of investigation. Its success and development are completely based on the progress in modern computer development. With their improvement the mathematical models will become more complicate and adequacy of real physical process representation by their means will increase

  7. Leakage potential through mechanical penetrations in a severe accident environment

    International Nuclear Information System (INIS)

    Koenig, L.N.

    1986-01-01

    This paper reviews the findings of an ongoing program, Integrity of Containment Penetrations Under Severe Accident Loads. The program is concerned with the leakage modes as well as the magnitude of leakage through mechanical penetrations in a containment building subject to a severe accident. Seal and gasket tests are used to evaluate the effect of radiation aging, thermal aging, seal geometry, and seal squeeze on seals and gaskets subjected to a hypothesized severe accident. The effects on leakage of the structural response of equipment hatches, personnel airlocks, and drywell heads subjected to severe accident pressures are studied by experiments and analyses. The data gathered during this program will be used to develop methodologies for predicting leakage

  8. Preliminary Calculation on a Spent Fuel Pool Accident using GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaehwan; Choi, Yu Jung; Hong, Tae Hyub; Kim, Hyeong-Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    The probability of an accident happening at the spent fuel pool was believed to be quite low until the 2011 Fukushima accident occurred. Notably, large amount of spent fuel are normally stored in the spent fuel pool for a long time compared to the amount of fuel in the reactor core and the total heat released from the spent fuel is high enough to boil the water of the spent fuel pool when the cooling system does not operate. In addition, the enrichment and the burnup of the fuel have both increased in the past decade and heat generation from the spent fuel thereby has also increased. The failure of the cooling system at the spent fuel pool (hereafter, a loss-of-cooling accident) is one of the principal hypothetical causes of an accident that could occur at the spent fuel pool. In this paper, the preliminary calculation of a loss-of-cooling accident was performed. In this paper, the preliminary calculation of a loss-of cooling accident was performed with GOTHIC. The calculation results show boiling away of water in the spent fuel pool due to the loss-of-cooling accident and similar thermal performance of the spent fuel pool with previous research results.

  9. Severe Accident Test Station Design Document

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Keiser, James R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

  10. Severe Accident Test Station Design Document

    International Nuclear Information System (INIS)

    Snead, Mary A.; Yan, Yong; Howell, Michael; Keiser, James R.; Terrani, Kurt A.

    2015-01-01

    The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

  11. Confinement barriers for loss of coolant accidents in the SEAFP reactor plant models

    International Nuclear Information System (INIS)

    Blomquist, R.; Ebert, E.; Gay, J.M.; Mazille, F.; Natalizio, A.; Rolandsson, S.; Ross, W.E.; Shen, K.; Sjoeberg, A.

    1995-01-01

    Loss of coolant accidents may mobilise radioactivity and pressurise confinement barriers thereby making a release to the environment possible. The paper defines the radioactivity confinements and presents principal results from the underlying thermal-hydraulic analyses. (orig.)

  12. Casebook on electric safety accidents

    International Nuclear Information System (INIS)

    1987-09-01

    This book gives concentration on electric safety accidents in domestic and abroad, which introduces general electrical safety with property of electricity, safe equipment and maintenance and protection of electric shock. It lists the cases of accident caused of electricity in domestic like accident in power substation, utilization equipment, load system and another accident by electricity like death in electric shock another by electricity like death in electric shock in new building construction, the cases caused of electricity in abroad like damage in electric shock by high voltage electric transformer, electric shock in summer and earth fault accident by fault cooling tower.

  13. Radiological accidents balance in medicine

    International Nuclear Information System (INIS)

    Nenot, J.C.

    1995-01-01

    This work deals with the radiological accidents in medicine. In medicine, the radiation accidents on medical personnel and patients can be the result of over dosage and bad focusing of radiotherapy sealed sources. Sometimes, the accidents, if they are unknown during a time enough for the source to be spread and to expose a lot of persons (in the case of source dismantling for instance) can take considerable dimensions. Others accidents can come from bad handling of linear accelerators and from radionuclide kinetics in some therapies. Some examples of accidents are given. (O.L.). 11 refs

  14. A comparison between the IAEA Safety Series 6 thermal environment and a proposed alternative thermal environment

    International Nuclear Information System (INIS)

    Wix, S.D.; Koski, J.A.

    1994-01-01

    The present regulations for packaging and transportation of radioactive materials, IAEA Safety Series No. 6; 1985, establish specific criteria for the thermal environment of a hypothetical accident. The regulation states: The scope of this paper is to examine the effects on modeling that result with the Fry proposed thermal boundary conditions. The examination is accomplished by comparing thermal model results using the current IAEA specified thermal environment and the Fry proposed thermal boundary conditions

  15. Improving performance with accident tolerant-fuels

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel S.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: rafael.orm@gmail.com, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Departamento de Engenharia Naval e Oceânica

    2017-07-01

    After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO{sub 2} - Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density - above that supported by UO{sub 2} - and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U{sub 3}Si{sub 2}, UN, and UC contain a higher uranium density and thermal conductivity than UO{sub 2}, providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U{sub 3}Si{sub 2}, UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO{sub 2}. These ATFs also have thermal conductivities approximately four times higher than that of UO{sub 2}. A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO{sub 2} - Zr as the fuel system of choice. (author)

  16. Three Mile Island accident

    International Nuclear Information System (INIS)

    Barre, B.; Olivier, E.; Roux, J.P.; Pelle, P.

    2010-01-01

    Deluded by equivocal instrumentation signals, operators at TMI-2 (Three Mile Island - unit 2) misunderstood what was going on in the reactor and for 2 hours were taking inadequate decisions that turned a reactor incident into a major nuclear event that led to the melting of about one third of the core. The TMI accident had worldwide impacts in the domain of nuclear safety. The main consequences in France were: 1) the introduction of the major accident approach and the reinforcement of crisis management; 2) the improvement of the reactor design, particularly that of the pressurizer valves; 3) the implementation of safety probabilistic studies; 4) a better taking into account of the feedback experience in reactor operations; and 5) a better taking into account of the humane factor in reactor safety. (A.C.)

  17. The ultimate nuclear accident

    International Nuclear Information System (INIS)

    Abdus Salam, A.

    1988-01-01

    The estimated energy equivalent of Chernobyl explosion was the 1/150 th of the explosive energy equivalent of atomic bomb dropped on Hiroshima; while the devastation that could be caused by the world's stock pile of nuclear weapons, could be equivalent to 160 millions of Chernobyl-like incidents. As known, the number of nuclear weapons is over 50,000 and 2000 nuclear weapons are sufficient to destroy the world. The Three Mile Island and Chernobyl accidents have been blamed on human factors but also the human element, particularly in the form of psychological stresses on those operating the nuclear weapons, could accidentally bring the world to a nuclear catastrophe. This opinion is encouraged by the London's Sunday Times magazine which gave a graphic description of life inside a nuclear submarine. So, to speak of nuclear reactor accidents and not of nuclear weapons is false security. (author)

  18. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Malinauskas, A.P.; Buchanan, J.R.; Lorenz, R.A.; Yamashita, T.

    1986-01-01

    On April 26, 1986, an explosion occurred at the newest of four operating nuclear reactors at the Chernobyl site in the USSR. The accident initiated an international technical exchange of almost unprecedented magnitude; this exchange was climaxed with a meeting at the International Atomic Energy Agency in Vienna during the week of August 25, 1986. The meeting was attended by more than 540 official representatives from 51 countries and 20 international organizations. Information gleaned from that technical exchange is presented in this report. A description of the Chernobyl reactor, which differs significantly from commercial US reactors, is presented, the accident scenario advanced by the Russian delegation is discussed, and observations that have been made concerning fission product release are described

  19. Nuclear ship accidents

    International Nuclear Information System (INIS)

    Oelgaard, P.L.

    1993-05-01

    In this report available information on 28 nuclear ship accident and incidents is considered. Of these 5 deals with U.S. ships and 23 with USSR ships. The ships are in almost all cases nuclear submarines. Only events that involve the nuclear propulsion plants, radiation exposures, fires/explosions and sea water leaks into the submarines are considered. Comments are made on each of the events, and at the end of the report an attempt is made to point out the weaknesses of the submarine designs which have resulted in the accidents. It is emphasized that much of the available information is of a rather dubious nature. consequently some of the assessments made may not be correct. (au)

  20. The Fukushima accident

    International Nuclear Information System (INIS)

    Maqua, M.; Stueck, R.

    2012-01-01

    On 11 March 2011, the Tohoku earthquake and the subsequent tsunami hit the Japanese east coast, causing more than 15,000 fatalities. To this date, 3,000 people are still missing. The Fukushima Dai-ichi NPP was the nuclear installation that was most affected by the tsunami. The earthquake cut off the NPP from the national grid. About 45 minutes later, the tsunami flooded units 1-4 and led to core meltdown events with large releases for units 1, 2 and 3. Unit 4 had been in refuelling outage at that time and lost the cooling of the spent fuel pool for several days. Considerable hydrogen explosions occurred in units 1, 3 and 4. Shortly after the accident, TEPCO started to mitigate the consequences of the accident by providing external cooling to the reactors and by removing the radioactive debris from the site. Great emphasis was laid on effective radiation protection measures for the clean-up workers. Thus, up to now there has been no fatality due to the radiation caused by the Fukushima accident. The main steps of the accident sequences are described, taking into account the latest findings of investigations performed by TEPCO or on behalf of the regulatory body. The presentation focuses on the description of the status of the Fukushima Dai-ichi nuclear power plant and the future steps for cleaning-up the site. In the presentation, the major phases of the roadmap that TEPCO has developed for the clean-up are highlighted. The risks associated with the current plant status and the clean-up phases are described. Abstract the content of the manuscript in a few lines.

  1. Severe accident management guidelines

    International Nuclear Information System (INIS)

    Uhle, Jennifer

    2014-01-01

    The events at Fukushima Daiichi have highlighted the importance of Severe Accident Management Guidelines (SAMGs). As the world has learned from the catastrophe and countries are considering changes to their nuclear regulatory programs, the content of SAMGs and their regulatory control are being evaluated. This presentation highlights several factors that are being addressed in the United States as rulemaking is underway pertaining to SAMGs. The question of how to be prepared for the unexpected is discussed with specific insights gleaned from Fukushima. (author)

  2. The Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Rassow, J.

    1986-01-01

    The documentation aims at giving a clearly arranged account of facts, interrelations and comparative evaluations of general interest. It deals with the course of events, atmospheric dispersion and fallout of the substances released and discusses the basic principles of the metering of radioactive radiation, the calculation of body doses and comparative evaluations with the radioactive exposure and risks involved by other sources. The author intends to contribute to an objective discussion about the Chernobyl reactor accident and nuclear energy as such. (DG) [de

  3. Radiation accident in Vietnam

    International Nuclear Information System (INIS)

    Wheatley, J.

    1994-01-01

    In November 1992 a Vietnamese research physicist was working with a microtron accelerator when he received a radiation overexposure that required the subsequent amputation of his right hand. A team from the International Atomic Energy Agency visited Hanoi in March 1993 to carry out an investigation. It was concluded that the accident occurred primarily because of a lack of safety systems, although the lack of both written procedures and training in basic radiation safety were also major contributors. (author)

  4. PREVENTION OF OCCUPATIONAL ACCIDENTS

    Directory of Open Access Journals (Sweden)

    Jovica Jovanovic

    2004-01-01

    Full Text Available Medical services, physicians and nurses play an essential role in the plant safety program through primary treatment of injured workers and by helping to identify workplace hazards. The physician and nurse should participate in the worksite investigations to identify specific hazard or stresses potentially causing the occupational accidents and injuries and in planning the subsequent hazard control program. Physicians and nurses must work closely and cooperatively with supervisors to ensure the prompt reporting and treatment of all work related health and safety problems. Occupational accidents, work related injuries and fatalities result from multiple causes, affect different segments of the working population, and occur in a myriad of occupations and industrial settings. Multiple factors and risks contribute to traumatic injuries, such as hazardous exposures, workplace and process design, work organization and environment, economics, and other social factors. With such a diversity of theories, it will not be difficult to understand that there does not exist one single theory that is considered right or correct and is universally accepted. These theories are nonetheless necessary, but not sufficient, for developing a frame of reference for understanding accident occurrences. Prevention strategies are also varied, and multiple strategies may be applicable to many settings, including engineering controls, protective equipment and technologies, management commitment to and investment in safety, regulatory controls, and education and training. Research needs are thus broad, and the development and application of interventions involve many disciplines and organizations.

  5. Radiological accident of Goiania

    International Nuclear Information System (INIS)

    Palacios, Elias; Gimenez, J.C.

    1988-01-01

    The accident of Goiania that took place in September, 1987, was the consequence of the dismantlement of the teletherapy equipment containing a Cs 137 source. The activity of the source was of about 5,2.10 13 Bq(1.400 Ci) and was made up by 1.10 -1 Kg of ClCs. This is one of the worst accidents, involving medical or industrial source, which happened up to the moment. The accident and the criteria adopted to face the emergency are described. The characteristics of the irradiation and superficial and internal contamination of the persons affected caused any dosimetric evaluation to be particularly difficult. The emergency control managed by capable persons was carried out with the expected difficulties in some areas within the Goiania city. About 90 % of the material could be recovered, in the removal works, in order to be treated as radioactive waste. Conclusions that should be taken into account in Argentina are reached. The authors of the article took part in the emergency by helping the brazilian Authorities within the Mutual Assistance Agreement between Argentina and Braxil. (M.E.L.) [es

  6. Thule accident 1968

    International Nuclear Information System (INIS)

    Melgaard, L.; Moeller Kristensen, H.

    1987-01-01

    On January 21, 1968 an American B-52 bomber crashed on the ice at Thule in Nothern Greenland. The bomber carried 4 nuclear weapons that were destroyed. The radioactive material of the bombs was spread over a large area of the ice. About 850 Danes stayed at the Thule base in 1968 for a shorter or longer period. Out of these 850 probably between 70 and 170 men took part in the clearing after the accident. Danish and American authorities establised that the radioactive contamination from the accident was too small to cause any health effects. For that reason the Danish authorities did not follow the Danish workers in order to show late effects, if any. In defiance of the authorities' very cocksure attitude towards possible late effets parts of the Danish press in the Summer 1986 started to be interested in the matter and to search for previous Thule-workers. Up till January 1987 aboput 600 workers have been contacted by the press, trade unions, and private persons. About 500 out of the 600 workers report on illness, cancers and deaths. This report tries to compile the accessible informations on the matter, to descibe the possible radioactive and chemical effects, to compare the Thule accident with a similar incident in Spain in 1966, and to propose a comprehensive health examination of all the workers. (LN)

  7. Severe accident research and management in Nordic Countries - A status report

    International Nuclear Information System (INIS)

    Frid, W.

    2002-01-01

    The report describes the status of severe accident research and accident management development in Finland, Sweden, Norway and Denmark. The emphasis is on severe accident phenomena and issues of special importance for the severe accident management strategies implemented in Sweden and in Finland. The main objective of the research has been to verify the protection provided by the accident mitigation measures and to reduce the uncertainties in risk dominant accident phenomena. Another objective has been to support validation and improvements of accident management strategies and procedures as well as to contribute to the development of level 2 PSA, computerised operator aids for accident management and certain aspects of emergency preparedness. Severe accident research addresses both the in-vessel and the ex-vessel accident progression phenomena and issues. Even though there are differences between Sweden and Finland as to the scope and content of the research programs, the focus of the research in both countries is on in-vessel coolability, integrity of the reactor vessel lower head and core melt behaviour in the containment, in particular the issues of core debris coolability and steam explosions. Notwithstanding that our understanding of these issues has significantly improved, and that experimental data base has been largely expanded, there are still important uncertainties which motivate continued research. Other important areas are thermal-hydraulic phenomena during reflooding of an overheated partially degraded core, fission product chemistry, in particular formation of organic iodine, and hydrogen transport and combustion phenomena. The development of severe accident management has embraced, among other things, improvements of accident mitigating procedures and strategies, further work at IFE Halden on Computerised Accident Management Support (CAMS) system, as well as plant modifications, including new instrumentation. Recent efforts in Sweden in this area

  8. Identification of the operating crew's information needs for accident management

    International Nuclear Information System (INIS)

    Nelson, W.R.; Hanson, D.J.; Ward, L.W.; Solberg, D.E.

    1988-01-01

    While it would be very difficult to predetermine all of the actions required to mitigate the consequences of every potential severe accident for a nuclear power plant, development of additional guidance and training could improve the likelihood that the operating crew would implement effective sever-accident management measures. The US Nuclear Regulatory Commission (NRC) is conducting an Accident Management Research Program that emphasizes the application of severe-accident research results to enhance the capability of the plant operating crew to effectively manage severe accidents. One element of this program includes identification of the information needed by the operating crew in severe-accident situations. This paper discusses a method developed for identifying these information needs and its application. The methodology has been applied to a generic reactor design representing a PWR with a large dry containment. The information needs were identified by systematically determining what information is needed to assess the health of the critical functions, identify the presence of challenges, select strategies, and assess the effectiveness of these strategies. This method allows the systematic identification of information needs for a broad range of severe-accident scenarios and can be validated by exercising the functional models for any specific event sequence

  9. Accidents and human factors

    International Nuclear Information System (INIS)

    Nishiwaki, Y.; Kawai, H.; Morishima, H.; Terano, T.; Sugeno, M.

    1984-01-01

    When the TMI accident occurred it was 4 a.m., an hour when the error potential of the operators would have been very high. The frequency of car and train accidents in Japan is also highest between 4 a.m. and 6 a.m. The error potential may be classified into five phases corresponding to the electroencephalogramic pattern (EEG). At phase 0, when the delta wave appears, a person is unconscious and in deep sleep; at phase I, when the theta wave appears, he is very tired, sleepy and subnormal; at phase II, when the alpha wave appears, he is normal, relaxed and passive; at phase III, when the beta wave appears, he is normal, clear-minded and active; at phase IV, when the strong beta or epileptic wave appears, he is hypernormal, excited and incapable of normal judgement. Should an accident occur at phase II, the brain condition may jump to phase IV. At this phase the error or accident potential is maximum. The response of the human brain to different types of noises and signals may vary somewhat for different individuals and for different groups of people. Therefore, the possibility that such differences in brain functions may influence the mental structure would be worthy of consideration in human factors and in the design of man-machine systems. Human reliability and performance would be affected by many factors: medical, physiological and psychological, etc. The uncertainty involved in human factors may not necessarily be probabilistic, but fuzzy. Therefore, it would be important to develop a theory by which both non-probabilistic uncertainties, or fuzziness, of human factors and the probabilistic properties of machines can be treated consistently. From the mathematical point of view, probabilistic measure is considered a special case of fuzzy measure. Therefore, fuzzy set theory seems to be an effective tool for analysing man-machine systems. To minimize human error and the possibility of accidents, new safety systems should not only back up man and make up for his

  10. Radiation accident/disaster

    International Nuclear Information System (INIS)

    Kida, Yoshiko; Hirohashi, Nobuyuki; Tanigawa, Koichi

    2013-01-01

    Described are the course of medical measures following Fukushima Daiichi Nuclear Power Plant (FNPP) Accident after the quake and tsunami (Mar. 11, 2011) and the future task for radiation accident/disaster. By the first hydrogen explosion in FNPP (Mar. 12), evacuation of residents within 20 km zone was instructed, and the primary base for measures of nuclear disaster (Off-site Center) 5 km afar from FNPP had to work as a front base because of damage of communicating ways, of saving of injured persons and of elevation of dose. On Mar. 13, the medical arrangement council consisting from stuff of Fukushima Medical University (FMU), National Institute of Radiological Sciences, Nuclear Safety Research Association and Prefectural officers was setup in residents' hall of Fukushima City, and worked for correspondence to persons injured or exposed, where communication about radiation and between related organizations was still poor. The Off-site Center's head section moved to Prefectural Office on Mar. 15 as headquarters. Early in the period, all residents evacuated from the 20 km zone, and in-hospital patients and nursed elderly were transported with vehicles, >50 persons of whom reportedly died mainly by their base diseases. The nation system of medicare for emergent exposure had consisted from the network of the primary to third facilities; there were 5 facilities in the Prefecture, 3 of which were localized at 4-9 km distance from FNPP and closed early after the Accident; and the secondary facility of FMU became responsible to all exposed persons. There was no death of workers of FNPP. Medical stuff also measured the ambient dose at various places near FNPP, having had risk of exposure. At the Accident, the important system of command, control and communication was found fragile and measures hereafter should be planned on assumption of the worst scenario of complete damage of the infrastructure and communication. It is desirable for Disaster Medical Assistance Team which

  11. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    International Nuclear Information System (INIS)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang

    2005-01-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  12. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    Energy Technology Data Exchange (ETDEWEB)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang [Nuclear Department, Suez-Tractebel Engineering, avenue Ariane 5, B-1200 Brussels (Belgium)

    2005-07-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  13. A Public Health Perspective of Road Traffic Accidents

    Science.gov (United States)

    Gopalakrishnan, S.

    2012-01-01

    Road traffic accidents (RTAs) have emerged as an important public health issue which needs to be tackled by a multi-disciplinary approach. The trend in RTA injuries and death is becoming alarming in countries like India. The number of fatal and disabling road accident happening is increasing day by day and is a real public health challenge for all the concerned agencies to prevent it. The approach to implement the rules and regulations available to prevent road accidents is often ineffective and half-hearted. Awareness creation, strict implementation of traffic rules, and scientific engineering measures are the need of the hour to prevent this public health catastrophe. This article is intended to create awareness among the health professionals about the various modalities available to prevent road accidents and also to inculcate a sense of responsibility toward spreading the message of road safety as a good citizen of our country. PMID:24479025

  14. A Public Health Perspective of Road Traffic Accidents

    Directory of Open Access Journals (Sweden)

    S Gopalakrishnan

    2012-01-01

    Full Text Available Road traffic accidents (RTAs have emerged as an important public health issue which needs to be tackled by a multi-disciplinary approach. The trend in RTA injuries and death is becoming alarming in countries like India. The number of fatal and disabling road accident happening is increasing day by day and is a real public health challenge for all the concerned agencies to prevent it. The approach to implement the rules and regulations available to prevent road accidents is often ineffective and half-hearted. Awareness creation, strict implementation of traffic rules, and scientific engineering measures are the need of the hour to prevent this public health catastrophe. This article is intended to create awareness among the health professionals about the various modalities available to prevent road accidents and also to inculcate a sense of responsibility toward spreading the message of road safety as a good citizen of our country.

  15. Accidents involving specialized aircraft in agriculture aerial spraying

    Directory of Open Access Journals (Sweden)

    Marcelo Boamorte Ravelli

    Full Text Available ABSTRACT: The great challenge for the practice of agricultural aviation has been to avoid accidents. Although, there are technological progress and high resources for safety, accidents continue to occur. The objective of this research was to analyze the influence and occurrence of factors in agricultural aviation accidents in Brazil recently. Based on research and technical - scientific papers written by researchers and aviation authorities, recommendations directed towards reducing the risks associated with this aircraft modality are assessed. The main factors responsible for accidents are normally operational errors and maneuvers that cause flight collisions, engine failures and altitude loss. Professional awareness and qualification converge towards the success of the agricultural pilot in the detection of inherent dangers or occasional in the various systems involved.

  16. Thermal reactor safety

    International Nuclear Information System (INIS)

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport

  17. Thermal reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

  18. 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

  19. Development of Severe Accident Containment Analysis Package

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang-Hwan; Kim, Dong-Min; Seo, Jea-Uk; Lee, Dea-Young; Park, Soon-Ho; Lee, Jae-Gwon; Lee, Jin-Yong; Lee, Byung-Chul [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    In safety viewpoint, the pressure and temperature of the containment is the important parameters, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In addition, there are possibilities of occurrence of other relevant phenomena following the reactor core melting such as DCH(direct containment heating) due to HPME(high pressure melt ejection), steam explosion due to fuel-coolant interaction in the reactor cavity and molten core concrete interaction at the late stage. It is important to predict the containment responses during a severe accident by a reasonable accuracy for establishing of effective mitigation strategies and preparation of the safety features required. In this paper, the overview of the SACAP development status is presented. SACAP is developed so as to be able to analyze, so called, Ex-Vessel severe accident phenomena including thermal-hydraulics, combustible gas burn, direct containment heating, steam explosion and molten core-concrete interaction. At the parallel time, SACAP and In-Vessel analysis module named CSPACE are processed for integration through MPI communication coupling. Development of the integrated severe accident analysis code system will be completed in following one year to make the code revision zero to be released.

  20. Severe accident consequence mitigation by filtered containment venting at Canadian nuclear power plants

    International Nuclear Information System (INIS)

    Lebel, Luke S.; Morreale, Andrew C.; Korolevych, Volodymyr; Brown, Morgan J.; Gyepi-Garbrah, Sam

    2017-01-01

    Highlights: • Use of filtered containment venting during a severe accident assessed. • Severe accident simulations performed using MAAP-CANDU and ADDAM. • Flow capacity, initiation protocols, efficiency, mass and thermal loading evaluated. • Efficient, robust system drastically reduces accident consequences. - Abstract: Having the capability to use filtered containment venting during a severe nuclear accident can significantly reduce its overall consequences. This study employs the MAAP-CANDU severe accident analysis code and the ADDAM atmospheric dispersion code to study the progression of: an unmitigated station blackout accident at a generic pressurized heavy water reactor, the release of radioactive material into the environment, the subsequent dispersion of the fission products through the atmosphere and the subsequent consequences (evacuation radius). The goal is to evaluate the application of filtered venting as an accident mitigation technology. Several aspects of filtered containment venting system design, like flow capacity, initiation protocols, filter efficiency, mass loading, and thermal loading are considered. An efficient and robust filtered containment venting system can reduce the amount of radiological materials emitted during an accident by 25 times or more, and as a result considerably reduce the off-site consequences of an accident.

  1. Investigation on accident management measures for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Tusheva, P.; Schaefer, F.; Rohde, U.; Reinke, N.

    2009-01-01

    A consequence of a total loss of AC power supply (station blackout) leading to unavailability of major active safety systems which could not perform their safety functions is that the safety criteria ensuring a secure operation of the nuclear power plant would be violated and a consequent core heat-up with possible core degradation would occur. Currently, a study which examines the thermal-hydraulic behaviour of the plant during the early phase of the scenario is being performed. This paper focuses on the possibilities for delay or mitigation of the accident sequence to progress into a severe one by applying Accident Management Measures (AMM). The strategy 'Primary circuit depressurization' as a basic strategy, which is realized in the management of severe accidents is being investigated. By reducing the load over the vessel under severe accident conditions, prerequisites for maintaining the integrity of the primary circuit are being created. The time-margins for operators' intervention as key issues are being also assessed. The task is accomplished by applying the GRS thermal-hydraulic system code ATHLET. In addition, a comparative analysis of the accident progression for a station blackout event for both a reference German PWR and a reference VVER-1000, taking into account the plant specifics, is being performed. (authors)

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

    International Nuclear Information System (INIS)

    Girard, P.; Dubreuil, G.H.

    1996-01-01

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

  3. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Vapor film collapse behavior on high temperature particle surface. JAERI's nuclear research promotion program, H10-027-3. Contract research

    International Nuclear Information System (INIS)

    Abe, Yutaka

    2002-03-01

    The experimental researches were conducted to study vapor film collapse behavior on high temperature melted core material coarsely mixed in the coolant under the film boiling condition. The film collapse is very important incipient incident of the trigger process for the vapor explosion in sever accident of nuclear reactor. In the experiment, pressure pulse was applied to the vapor film on a high temperature particle surface simulating melted core material to observed microscopic vapor film collapse behavior with a high-speed video camera of 40,500 fps. The particle surface temperature and pressure around the particle were simultaneously measured. The transition of the vapor film thickness and two-dimensional vapor-liquid interface movement and the velocity were estimated with visual data analysis technique, PIV and digital data analysis technique. Furthermore, heat conduction analysis was performed to estimate the vapor-liquid interfacial temperature with the measured temperature and estimated vapor film thickness. As the results, it was clarified that the vapor-liquid interface changed white from transparent view for all the experimental conditions. It is also clarified that the vapor-liquid interfacial temperature decreased under the saturation temperature when the pressure pulse arrive at the particle. The experimental facts indicates the possibility that the vapor film collapse occurs due to the liquid phase homogeneous moving toward the particle drove by the pressure reduction caused by the phase change inside the vapor film. (author)

  4. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Heat transfer on the liquid-liquid interface between molten core pool and coolant. JAERI's nuclear research promotion program, H10-027-6. Contract research

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Saito, Yasushi

    2002-03-01

    Heat transfer experiments under steady and transient conditions were performed using molten Wood's metal and distilled water to study heat transfer on the liquid-liquid interface between molten fuel pool and coolant under severe accident conditions. In the steady state experiment, boiling curve was measured over the range from natural convection region to film boiling region. The boiling behavior was observed using a high-speed video camera. In the transient experiment, distilled water was poured onto the hot molten metal surface, and the boiling curve was obtained in the cooling process. Comparing the measured boiling curve with existing correlations and experimental data for solid-liquid and liquid-liquid systems, the following conclusions were drawn: (a) When the interface surge is negligible and oxide layer is formed on the interface, the boiling curve at the liquid-liquid surface could be approximately reproduced by the heat transfer correlations for nucleate boiling and film boiling regions and the critical heat flux correlation for a liquid-solid system. (b) When no oxide layer is formed on the interface, the boiling curve at the liquid-liquid surface moved towards higher wall superheat than that at the liquid-solid surface, as Novakovic et al. observed in their experiment using mercury. (c) Transient heat transfer coefficient for film boiling at the liquid-liquid surface was about 100% higher than that predicted by the heat transfer correlation for a solid-liquid system. (author)

  5. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Vapor film collapse behavior on high temperature particle surface. JAERI's nuclear research promotion program, H10-027-3. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Yutaka [Tsukuba Univ., Institute of Engineering Mechanics and Systems, Tsukuba, Ibaraki (Japan)

    2002-03-01

    The experimental researches were conducted to study vapor film collapse behavior on high temperature melted core material coarsely mixed in the coolant under the film boiling condition. The film collapse is very important incipient incident of the trigger process for the vapor explosion in sever accident of nuclear reactor. In the experiment, pressure pulse was applied to the vapor film on a high temperature particle surface simulating melted core material to observed microscopic vapor film collapse behavior with a high-speed video camera of 40,500 fps. The particle surface temperature and pressure around the particle were simultaneously measured. The transition of the vapor film thickness and two-dimensional vapor-liquid interface movement and the velocity were estimated with visual data analysis technique, PIV and digital data analysis technique. Furthermore, heat conduction analysis was performed to estimate the vapor-liquid interfacial temperature with the measured temperature and estimated vapor film thickness. As the results, it was clarified that the vapor-liquid interface changed white from transparent view for all the experimental conditions. It is also clarified that the vapor-liquid interfacial temperature decreased under the saturation temperature when the pressure pulse arrive at the particle. The experimental facts indicates the possibility that the vapor film collapse occurs due to the liquid phase homogeneous moving toward the particle drove by the pressure reduction caused by the phase change inside the vapor film. (author)

  6. Thermal test options

    International Nuclear Information System (INIS)

    Koski, J.A.; Keltner, N.R.; Sobolik, K.B.

    1993-02-01

    Shipping containers for radioactive materials must be qualified to meet a thermal accident environment specified in regulations, such at Title 10, Code of Federal Regulations, Part 71. Aimed primarily at the shipping container design, this report discusses the thermal testing options available for meeting the regulatory requirements, and states the advantages and disadvantages of each approach. The principal options considered are testing with radiant heat, furnaces, and open pool fires. The report also identifies some of the facilities available and current contacts. Finally, the report makes some recommendations on the appropriate use of these different testing methods

  7. Thermodynamic correlations for the accident analysis of HTR's

    International Nuclear Information System (INIS)

    Rehm, W.; Jahn, W.; Finken, R.

    1976-12-01

    The thermal properties of Helium and for the case of a depressurized primary circuit, various mixtures of primary cooling gas were taken into consideration. The temperature dependence of the correlations for the thermal properties of the graphite components in the core and for the structural materials in the primary circuit are extrapolated about normal operation conditions. Furthermore the correlations for the effective thermal conductivity, the heat transfer and pressure drop are described for pebble bed HTR's. In addition some important heat transfer data of the steam generator are included. With these correlations, for example accident sequences with failure of the afterheat removal systems are discussed for pebble bed HTR's. It is concluded that the transient temperature behaviour demonstrates the inherent safety features of the HTR in extreme accidents. (orig.) [de

  8. Occupational accidents aboard merchant ships

    DEFF Research Database (Denmark)

    Hansen, H.L.; Nielsen, D.; Frydenberg, Morten

    2002-01-01

    Objectives: To investigate the frequency, circumstances, and causes of occupational accidents aboard merchant ships in international trade, and to identify risk factors for the occurrence of occupational accidents as well as dangerous working situations where possible preventive measures may...... be initiated. Methods: The study is a historical follow up on occupational accidents among crew aboard Danish merchant ships in the period 1993–7. Data were extracted from the Danish Maritime Authority and insurance data. Exact data on time at risk were available. Results: A total of 1993 accidents were...... aboard. Relative risks for notified accidents and accidents causing permanent disability of 5% or more were calculated in a multivariate analysis including ship type, occupation, age, time on board, change of ship since last employment period, and nationality. Foreigners had a considerably lower recorded...

  9. Some Examples of Accident Analyses for RB Reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    2002-01-01

    The RB reactor is heavy water critical assembly operated in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, since April 1959. The first Safety Analysis Report of the RB critical assembly was prepared in 1961/62. But, the first accidental analysis was done in late 1958 in aim the examine power transient and total equivalent doses received by the staff during the reactivity accident occurred on October 15, 1958. Since 1960, the RB reactor is modified few times. Beside initial natural uranium metal fuel rods, new fuel (TVR-S types) from 2% enriched metal uranium and 80% enriched UO 2 were available since 1962 and 1976, respectively. Also, modifications in control and safety systems of the reactor were done occasionally. Special reactor cores were created using all three types of fuel elements, among them, the coupled fast-thermal ones. Nuclear Safety Committee of the Vinca Institute, an independent regulatory body approved for usage all these modifications of the RB reactor. For those decisions of the Committee, the Preliminary Safety Analysis Reports were prepared that, beside proposed technical modifications and new regulation rules had included analyses of various possible accidents. Special attention is given and new methodology was proposed for thoroughly analyses of design based accidents related to coupled fast-thermal cores, that include reactor central zones filled by fuel elements without moderator. In these accidents, during assumed flooding of the fast zone by moderator, a very high reactivity could be inserted in the system with very high reactivity rate. It was necessary to provide that the safety system of the reactor had fast response to that accident and had enough high (negative) reactivity to shut down the reactor timely. In this paper, a brief overview of some accidents, methodology and computation tools used for the accident analyses at RB reactor are given. (author)

  10. Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

    International Nuclear Information System (INIS)

    Park, S. Y.; Song, Y. M.

    2015-01-01

    The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a Level 2 probabilistic safety assessment or severe accident management strategy developments. The code has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building

  11. Validation of severe accident management guidance for the wolsong plants

    International Nuclear Information System (INIS)

    Park, S. Y.; Jin, Y. H.; Kim, S. D.; Song, Y. M.

    2006-01-01

    Full text: Full text: The severe accident management(SAM) guidance has been developed for the Wolsong nuclear power plants in Korea. The Wolsong plants are 700MWe CANDU-type reactors with heavy water as the primary coolant, natural uranium-fueled pressurized, horizontal tubes, surrounded by heavy water moderator inside a horizontal calandria vessel. The guidance includes six individual accident management strategies: (1) injection into primary heat transport system (2) injection into calandria vessel (3) injection into calandria vault (4) reduction of fission product release (5) control of reactor building condition (6) reduction of reactor building hydrogen. The paper provides the approaches to validate the SAM guidance. The validation includes the evaluation of:(l) effectiveness of accident management strategies, (2) performance of mitigation systems or components, (3) calculation aids, (4) strategy control diagram, and (5) interface with emergency operation procedure and with radiation emergency plan. Several severe accident sequences with high probability is selected from the plant specific level 2 probabilistic safety analysis results for the validation of SAM guidance. Afterward, thermal hydraulic and severe accident phenomenological analyses is performed using ISAAC(Integrated Severe Accident Analysis Code for CANDU Plant) computer program. Furthermore, the experiences obtained from a table-top-drill is also discussed

  12. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    1986-05-01

    Following the accident at Chernobyl nuclear reactor, WHO organized on 6 May 1986 in Copenhagen a one day consultation of experts with knowledge in the fields of meteorology, radiation protection, biological effects, reactor technology, emergency procedures, public health and psychology in order to analyse the development of events and their consequences and to provide guidance as to the needs for immediate public health action. The present report provides detailed information on the transportation and dispersion of the radioactive material in the atmosphere, especially volatile elements, during the release period 26 April - 5 May. Presented are the calculated directions and locations of the radioactive plume over Europe in the first 5 days after the accident, submitted by the Swedish Meteorological and Hydrological Institute. The calculations have been made for two heights, 1500m and 750m and the plume directions are grouped into five periods, covering five European areas. The consequences of the accident inside the USSR and the radiological consequences outside the USSR are presented including the exposure routes and the biological effects, paying particular attention to iodine-131 effects. Summarized are the first reported measured exposure rates above background, iodine-131 deposition and concentrations in milk and the remedial actions taken in various European countries. Concerning the cesium-137 problem, based on the UNSCEAR assessment of the consequences of the nuclear fallout, one concludes that the cesium contamination outside the USSR is not likely to cause any serious problems. Finally, the conclusions and the recommendations of the meeting, taking into account both the short-term and longer term considerations are presented

  13. Thermal Properties Measurement Report

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, Jon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Braase, Lori [Idaho National Lab. (INL), Idaho Falls, ID (United States); Papesch, Cynthia [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hurley, David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tonks, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gofryk, Krzysztof [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Fielding, Randy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Knight, Collin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Meyer, Mitch [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    The Thermal Properties Measurement Report summarizes the research, development, installation, and initial use of significant experimental thermal property characterization capabilities at the INL in FY 2015. These new capabilities were used to characterize a U3Si2 (candidate Accident Tolerant) fuel sample fabricated at the INL. The ability to perform measurements at various length scales is important and provides additional data that is not currently in the literature. However, the real value of the data will be in accomplishing a phenomenological understanding of the thermal conductivity in fuels and the ties to predictive modeling. Thus, the MARMOT advanced modeling and simulation capability was utilized to illustrate how the microstructural data can be modeled and compared with bulk characterization data. A scientific method was established for thermal property measurement capability on irradiated nuclear fuel samples, which will be installed in the Irradiated Material Characterization Laboratory (IMCL).

  14. GAPCON-THERMAL-3

    International Nuclear Information System (INIS)

    Mohr, C.L.; Lanning, D.D.; Panisko, F.E.

    1979-01-01

    The fuel performance code GAPCON-THERMAL-3 has been expanded to include recent transient material deformation constitutive relations and the FLECHT heat transfer correlation. The modifications make it possible to compute the thermal and mechanical response of nuclear fuel to postulated Loss of Coolant Accidents (LOCA). The numerical formulation has the capability of predicting both steady state and transient behavior of a fuel rod using a single analytical procedure. GAPCON-THERMAL-3 (G-T-3) uses a specialized finite element procedure for mechanics predictions and the method of weighted residuals and finite difference techniques to compute temperature and thermal behavior. Fuel behavior, gas release models, gas conductance models, and stored energy calculations are applicable to both steady state and transient conditions. The code has been used to perform scoping analysis for in-reactor LOCA simulation testing. (orig.)

  15. Traffic accidents: an econometric investigation

    OpenAIRE

    Tito Moreira; Adolfo Sachsida; Loureiro Paulo

    2004-01-01

    Based on a sample of drivers in Brasilia's streets, this article investigates whether distraction explains traffic accidents. A probit model is estimated to determine the predictive power of several variables on traffic accidents. The main conclusion drawn from this study is that the proxies used to measure distraction, such as the use of cell phones and cigarette smoking in a moving vehicle, are significant factors in determining traffic accidents.

  16. Medical aspects of radiation accidents

    International Nuclear Information System (INIS)

    Messerschmidt, O.

    1990-01-01

    Reactor accidents and nuclear bomb explosions are compared including the release of radioactivity in an accident, results of risk studies, emergency measures of nuclear power plants, and evacuation of the population. The medical aspects refer to the prophylaxies of the thyroid gland, contamination and decontamination of body surfaces, recommendations of the ICRP, radiation injury after total body exposure and medical problems after a reactor accident. (DG)

  17. Accident management approach in Armenia

    International Nuclear Information System (INIS)

    Ghazaryan, K.

    1999-01-01

    In this lecture the accident management approach in Armenian NPP (ANPP) Unit 2 is described. List of BDBAs had been developed by OKB Gydropress in 1994. 13 accident sequences were included in this list. The relevant analyses had been performed in VNIIAES and the 'Guidelines on operator actions for beyond design basis accident (BDBA) management at ANPP Unit 2' had been prepared. These instructions are discussed

  18. Proceedings of the 8. Brazilian Meeting on Reactor Physics and Thermal Hydraulics

    International Nuclear Information System (INIS)

    1991-01-01

    Some papers about pressurized light water reactors, fast reactors, accident analysis, transients, research reactors, nuclear data collection, thermal hydraulics, reactor monitoring, neutronics are presented. (E.G.)

  19. Nuclear accidents and protection

    International Nuclear Information System (INIS)

    Biocanin, R.; Amidzic, B.

    2005-01-01

    The numerous threats are our cruel reality. There is a great arsenal of nuclear weapons. Nuclear terrorism and nuclear accidents are always possible, especially during the transport and handling different nuclear material. Terrorist organisation also goes for coming into the possession of the nuclear means. Specific and important problem is human radioactive contamination in using nuclear energy for peaceful and military purpose. So, realisation of the universal and united system of NBCD gives us a possibility by using the modern communication equipment and very effective mobile units to react in a real time and successfully perform monitoring, alarming, protection and decontamination. (author) [sr

  20. The accident of Chernobyl

    International Nuclear Information System (INIS)

    1986-10-01

    RBMK reactors (reactor control, protection systems, containment) and the nuclear power plant of Chernobyl are first presented. The scenario of the accident is given with a detailed chronology. The actions and consequences on the site are reviewed. This report then give the results of the source term estimation (fision product release, core inventory, trajectories, meteorological data...), the radioactivity measurements obtained in France. Health consequences for the French population are evoked. The medical consequences for the population who have received a high level of doses are reviewed [fr

  1. Psychological response of accident

    International Nuclear Information System (INIS)

    Novikov, V.S.; Nikiforov, A.M.; Cheprasov, V.Yu.

    1996-01-01

    The psychological status of rescuers of consequences of Chernobyl[s accidents, having planned stationary examination and treatment of common somatic diseases, has been examined. THe age of men represented the study group was 35-54 years old. The results of medical-psychological examination showed the development in rescuers of common dysadaptation and stress state, characterized by depressive-hypochondriac state with high anxiety. The course of psychotherapeutic activities made possible to improve essentionally the psychological status of the patients. 12 refs., 3 figs., 1 tab

  2. Reactor accident in Chernobyl

    International Nuclear Information System (INIS)

    Sokalski, A.; Kowalski, A.

    1990-11-01

    The bibliography contains 1568 descriptions of papers devoted to Chernobylsk accident and recorded in ''INIS Atomindex'' to 30 June 1990. The descriptions were taken from ''INIS Atomindex'' and are presented in accordance with volumes of this journal (chronology of recording). Therefore all descriptions have numbers showing first the number of volume and then the number of record. The bibliography has at the end the detailed subject index consisting of 465 main headings and a lot of qualifiers. Some of them are descriptors taken from ''INIS Atomindex'' and some are key words taken from natural language. The index is in English as descriptions in the bibliography. (author)

  3. Accident prevention programme

    International Nuclear Information System (INIS)

    1978-01-01

    This study by the Steel Industry Safety and Health Commission was made within the context of the application by undertakings of the principles of accident and disease prevention previously adopted by the said Commission. It puts forward recommendations for the effective and gradual implementation of a programme of action on occupational health and safety in the various departments of an undertaking and in the undertaking as a whole. The methods proposed in this study are likely to be of interest to all undertakings in the metallurgical industry and other industrial sectors

  4. Serious accident in Peru

    International Nuclear Information System (INIS)

    Anon.

    1999-01-01

    A peruvian man, victim of an important accidental irradiation arrived on the Saturday twenty ninth of may 1999 to the centre of treatment of serious burns at the Percy military hospital (Clamart -France). The accident spent on the twentieth of February 1999, on the site of a hydroelectric power plant, in construction at 300 km at the East of Lima. The victim has picked up an industrial source of iridium devoted to gamma-graphy operations and put it in his back pocket; of trousers. The workman has serious radiation burns. (N.C.)

  5. 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

  6. Guidance on accidents involving radioactivity

    International Nuclear Information System (INIS)

    1989-01-01

    This annex contains advice to Health Authorities on their response to accidents involving radioactivity. The guidance is in six parts:-(1) planning the response required to nuclear accidents overseas, (2) planning the response required to UK nuclear accidents a) emergency plans for nuclear installations b) nuclear powered satellites, (3) the handling of casualties contaminated with radioactive substances, (4) background information for dealing with queries from the public in the event of an accident, (5) the national arrangements for incident involving radioactivity (NAIR), (6) administrative arrangements. (author)

  7. 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

  8. Importance of risk communication during and after a nuclear accident.

    Science.gov (United States)

    Perko, Tanja

    2011-07-01

    Past nuclear accidents highlight communication as one of the most important challenges in emergency management. In the early phase, communication increases awareness and understanding of protective actions and improves the population response. In the medium and long term, risk communication can facilitate the remediation process and the return to normal life. Mass media play a central role in risk communication. The recent nuclear accident in Japan, as expected, induced massive media coverage. Media were employed to communicate with the public during the contamination phase, and they will play the same important role in the clean-up and recovery phases. However, media also have to fulfill the economic aspects of publishing or broadcasting, with the "bad news is good news" slogan that is a well-known phenomenon in journalism. This article addresses the main communication challenges and suggests possible risk communication approaches to adopt in the case of a nuclear accident. Copyright © 2011 SETAC.

  9. Nuclear energy safety - new challenges

    Energy Technology Data Exchange (ETDEWEB)

    Rausch, Julio Cezar; Fonseca, Renato Alves da, E-mail: jrausch@cnen.gov.b, E-mail: rfonseca@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    Fukushima accident in March this year, the second most serious nuclear accident in the world, put in evidence a discussion that in recent years with the advent of the 'nuclear renaissance' has been relegated in the background: what factors influence the use safe nuclear energy? Organizational precursor, latent errors, reduction in specific areas of competence and maintenance of nuclear programs is a scenario where the guarantee of a sustainable development of nuclear energy becomes a major challenge for society. A deep discussion of factors that influenced the major accidents despite the nuclear industry use of the so-called 'lessons learned' is needed. Major accidents continue to happen if a radical change is not implemented in the focus of safety culture. (author)

  10. Nuclear energy safety - new challenges

    International Nuclear Information System (INIS)

    Rausch, Julio Cezar; Fonseca, Renato Alves da

    2011-01-01

    Fukushima accident in March this year, the second most serious nuclear accident in the world, put in evidence a discussion that in recent years with the advent of the 'nuclear renaissance' has been relegated in the background: what factors influence the use safe nuclear energy? Organizational precursor, latent errors, reduction in specific areas of competence and maintenance of nuclear programs is a scenario where the guarantee of a sustainable development of nuclear energy becomes a major challenge for society. A deep discussion of factors that influenced the major accidents despite the nuclear industry use of the so-called 'lessons learned' is needed. Major accidents continue to happen if a radical change is not implemented in the focus of safety culture. (author)

  11. Nuclear energy safety - new challenges

    Energy Technology Data Exchange (ETDEWEB)

    Rausch, Julio Cezar; Fonseca, Renato Alves da, E-mail: jrausch@cnen.gov.b, E-mail: rfonseca@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    Fukushima accident in March this year, the second most serious nuclear accident in the world, put in evidence a discussion that in recent years with the advent of the 'nuclear renaissance' has been relegated in the background: what factors influence the use safe nuclear energy? Organizational precursor, latent errors, reduction in specific areas of competence and maintenance of nuclear programs is a scenario where the guarantee of a sustainable development of nuclear energy becomes a major challenge for society. A deep discussion of factors that influenced the major accidents despite the nuclear industry use of the so-called 'lessons learned' is needed. Major accidents continue to happen if a radical change is not implemented in the focus of safety culture. (author)

  12. MELCOR Severe Accident Analysis on the SMART Reactor

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Jin, Young Ho; Kim, Young In; Kim, Keung Koo; Wang, Ziao; Revankar, Shripad

    2014-01-01

    A severe accident is analyzed for Korea SMR reactor, SMART. Core melt down sequences are analyzed for SMART reactor core using MELCOR version 1.8.5. MELCOR is developed by Sandia National Laboratory for US NRC for the simulation of severe accidents in nuclear power plants. Two cases are simulated here and compared between them; one is the case for core having 3 concentric rings and the other is the case for core having 5 concentric rings. One inch break LOCA scenario is simulated and compared between these two core models. Time sequences for the thermal hydraulic behaviors of RPV and thermal heatup behaviors of reactor core are explained in graphically. Thermal hydraulic behavior such as the change of pressure, level, mass, and temperature of RPV is explained. Thermal heatup behavior of reactor core such as oxidation of cladding, hydrogen generation, core slumping down to lower plenum, and finally creep rupture of PRV lower head is explained. Engineered safety features such as safety injection systems (SIS), and Passive residual heat removal systems (PHRS), etc. are assumed to be not working. One inch break of severe accident is simulated on Korean SMR (SMART) Integral PWR with MELCOR code version 1.8.5. Core melt progression and lower head failure time is very slow compared to other commercial reactors. Simulation on 3 and 5 radial rings core models gives very similar pattern in core cell failure timings. Other various accident scenarios (for example, SBO in Fukushima) will be tried further. Containment behaviors and source term behaviors in severe accident conditions will be analyzed in future

  13. 1976 Hanford americium accident

    International Nuclear Information System (INIS)

    Heid, K.R.; Breitenstein, B.D.; Palmer, H.E.; McMurray, B.J.; Wald, N.

    1979-01-01

    This report presents the 2.5-year medical course of a 64-year-old Hanford nuclear chemical operator who was involved in an accident in an americium recovery facility in August 1976. He was heavily externally contaminated with americium, sustained a substantial internal deposition of this isotope, and was burned with concentrated nitric acid and injured by flying debris about the face and neck. The medical care given the patient, including the decontamination efforts and clinical laboratory studies, are discussed. In-vivo measurements were used to estimate the dose rates and the accumulated doses to body organs. Urinary and fecal excreta were collected and analyzed for americium content. Interpretation of these data was complicated by the fact that the intake resulted both from inhalation and from solubilization of the americium embedded in facial tissues. A total of 1100 μCi was excreted in urine and feces during the first 2 years following the accident. The long-term use of diethylenetriaminepentate (DTPA), used principally as the zinc salt, is discussed including the method, route of administration, and effectiveness. To date, the patient has apparently experienced no complications attributable to this extensive course of therapy, even though he has been given approximately 560 grams of DTPA. 4 figures, 1 table

  14. Serious reactor accidents reconsidered

    International Nuclear Information System (INIS)

    1987-12-01

    The chance is determined for damage of the reactor core and that sequel events will cause excursion of radioactive materials into the environment. The gravity of such an accident is expressed by the source term. It appears that the chance for such an accident varies with the source term. In general it is valid that how larger the source term how smaller the chance is for it and vice versa. The chance for excursion is related to two complexes of events: serious damage (meltdown) of the reactor core, and the escape of the liberated radionuclides into the environment. The results are an order of magnitude consideration of the relation between the extent of the source term and the chance for it. From the spectrum of possible source terms three representative ones have been chosen: a large, a medium and a relative small source term. This choice is in accordance with international considerations. The hearth of this study is the estimation of the chance for occurrence of the three chosen source terms for new light-water reactors. refs.; figs.; tabs

  15. Assessment of accident energetics in LMFBR core-disruptive accidents

    International Nuclear Information System (INIS)

    Fauske, H.K.

    1977-01-01

    An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggests that such events are highly unlikely following a postulated core meltdown event

  16. Accident analysis. A review of the various accidents classifications

    International Nuclear Information System (INIS)

    Martin Martin, L.; Figueras, J.M.

    1982-01-01

    The objective of the accident analysis, in relation with the safety evaluation, environmental impact and emergency planning, should be to identify the total risk to the population and workers from potential accidents in the facility, analizing it over full spectrum of severity. (auth.)

  17. Accounting for the cost of occupational accidents

    DEFF Research Database (Denmark)

    Rikhardsson, Pall M.

    2004-01-01

    consequences for the company. This, however, presents some challenges due to the current set up of many management accounting systems. The paper explores these issues in the context of the Systematic Accident Cost Analysis (SACA) project, which was carried out during 2001 by The Aarhus School of Business...... and PricewaterhouseCoopers Denmark with financial support from The Danish National Working Environment Authority. It focused on developing and testing a method for the evaluation of the occupational costs and how this might be linked to management accounting and control systems....

  18. Study on structural failure of RPV with geometric discontinuity under severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Mao, J.F., E-mail: jianfeng-mao@163.com [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Zhu, J.W. [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Department of Mechanical and Electrical engineering, Huzhou Vocational & Technical College Huzhou, Zhejiang 313000 (China); Bao, S.Y., E-mail: bsy@zjut.edu.cn [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Luo, L.J. [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Gao, Z.L. [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China)

    2016-10-15

    Highlights: • The RPV failure is investigated in depth under severe accident. • The creep and plastic damage are the major contributor to RPV failure. • A elastic core is found at the midpoint of the highly-eroded region. • Weakest location has some ‘accommodating’ quality to prevent ductile tearing. • The internal pressure is critical for the determination of structural failure. - Abstract: A severe accident management strategy known as ‘in-vessel retention (IVR)’ is widely adopted in most of advanced nuclear power plants. The IVR mitigation is assumed to be able to arrest the degraded melting core and maintain the structural integrity of reactor pressure vessel (RPV) within a prescribed period of time. This traditional concept of IVR without consideration of internal pressure effect wasn’t challenged until the occurrence of Fukushima accident on 2011, which showed that the structural behavior had not been appropriately assessed, and a certain pressure (up to 8.0 MPa) still existed inside the RPV. Accordingly, the paper tries to address the related issue on whether lower head (LH) integrity can be maintained, when the LH is subjected to the thermal-mechanical loads created during such a severe accident. Because of the presence of the high temperature melt (∼1300 °C) on the inside of RPV, some local material is melted down to create a unique RPV with geometric discontinuity, while the outside of RPV submerged in cavity water will remain in nucleate boiling (at ∼150 °C). Therefore, the failure mechanisms of RPV can span a wide range of structural behaviors, such as melt-through, creep damage, plastic yielding as well as thermal expansion. Through meticulous investigation, it is found that the RPV failure is mainly caused by creep and plasticity, especially for the inside of highly-eroded region. The elastic core (or layer) is found to exist in the proximity of mid-section of the highly-eroded wall. However, the elastic core is squeezed into

  19. Development of instrumentation systems for severe accidents. 4. New accident tolerant in-containment pressure transducer for containment pressure monitoring system

    International Nuclear Information System (INIS)

    Oba, Masato; Teruya, Kuniyuki; Yoshitsugu, Makoto; Ikeuchi, Takeshi

    2015-01-01

    The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident and are developing the Severe Accident Instrumentations and Monitoring Systems that could measure the parameters in severe accident conditions. Mitsubishi Heavy Industries, LTD (MHI) developed a new accident tolerant containment pressure monitoring system and demonstrated that the monitoring system could endure extremely harsh environmental conditions that envelop severe accident environmental conditions inside a containment such as maximum operating temperature of up to 300degC and total integrated dose (TID) of 1 MGy gamma. The new containment pressure monitoring system comprises of a strain gage type pressure transducer and a mineral insulated (MI) cable with ceramic connectors, which are located in the containment, and a strain measuring amplifier located outside the containment. Less thermal and radiation degradation is achieved because of minimizing use of organic materials for in-containment equipment such as the transducer and connectors. Several tests were performed to demonstrate the performance and capability of the in-containment equipment under severe accident environmental conditions and the major steps in this testing were run in the following test sequences: (1) the baseline functional tests (e.g., repeatability, non-linearity, hysteresis, and so on) under normal conditions, (2) accident radiation testing, (3) seismic testing, and (4) steam/temperature test exposed to simulated severe accident environmental conditions. The test results demonstrate that the new pressure transducer can endure the simulated severe accident conditions. (author)

  20. The lesser known challenge of climate change: thermal variance and sex-reversal in vertebrates with temperature-dependent sex determination.

    Directory of Open Access Journals (Sweden)

    Jennifer L Neuwald

    Full Text Available Climate change is expected to disrupt biological systems. Particularly susceptible are species with temperature-dependent sex determination (TSD, as in many reptiles. While the potentially devastating effect of rising mean temperatures on sex ratios in TSD species is appreciated, the consequences of increased thermal variance predicted to accompany climate change remain obscure. Surprisingly, no study has tested if the effect of thermal variance around high-temperatures (which are particularly relevant given climate change predictions has the same or opposite effects as around lower temperatures. Here we show that sex ratios of the painted turtle (Chrysemys picta were reversed as fluctuations increased around low and high unisexual mean-temperatures. Unexpectedly, the developmental and sexual responses around female-producing temperatures were decoupled in a more complex manner than around male-producing values. Our novel observations are not fully explained by existing ecological models of development and sex determination, and provide strong evidence that thermal fluctuations are critical for shaping the biological outcomes of climate change.

  1. Method of assessing severe accident management strategies

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.; Okrent, D.; Jae, M.; Lim, H.; Milici, T.; Park, H.; Swider, J.; Xing, L.; Yu, D.

    1991-01-01

    Accident management can be defined as the innovative use of existing and or alternative resources, systems, and actions to prevent or mitigate a severe accident. A significant number of probabilistic safety assessments (PSAs) have been completed that yield the principal plant vulnerabilities. These vulnerabilities can be categorized as (1) dominant sequences with respect to core-melt frequency. (2) dominant sequences with respect to various risk measures. (3) dominant threats that challenge safety functions. (4) dominant threats with respect to failure of safety systems. For each sequence/threat and each combination of strategy, there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These considerations include uncertainties in key phenomena, operator behavior, system availability and behavior, and available information. This paper presents a methodology for assessing severe accident management strategies given the key uncertainties delineated at two workshops held at the University of California, Los Angeles. Based on decision trees and influence diagrams, the methodology is currently being applied to two case studies: cavity flooding in a pressurized water reactor (PWR) to prevent vessel penetration or failure, and drywell flooding in a boiling water reactor to prevent vessel and/or containment failure

  2. Drinking and Driving: Alcohol Association with Traffic Accidents.

    Science.gov (United States)

    Stacey, Barrie G.

    1985-01-01

    Presents an analysis of drink-driving research methods and findings with reference to traffic accidents. Challenges some conclusions about drinking and driving in Australia and New Zealand. Evaluates the growing acceptance of Scandinavian-type laws. Rejects the demand to "criminalize" drink-driving offenses. Presents the reduction of…

  3. Legal consequences of the Chernobyl accident in the FRG

    International Nuclear Information System (INIS)

    1988-01-01

    The author considers that the Chernobyl accident was a challenge to lawmakers and lawyers. This paper reviews the different aspects under which it has tested the legal system governing the peaceful use of nuclear energy in the Federal Republic of Germany. In particular, regulations protecting the public from the dangers of ionizing radiation proved to be inadequate and had to be amended (NEA) [fr

  4. The international cooperation using the example of the reactor accident in Chernobyl

    International Nuclear Information System (INIS)

    Molitor, Norbert

    2017-01-01

    The explosion of the reactor unit 4 of the NPP Chernobyl and the subsequent fire was up to now the most severe accident in the civil nuclear industry. The consequences of the accident far outside the Ukraine and the former Soviet Union demonstrated that nuclear safety is a trans-border challenge. The mitigation of the accident consequences and the recovery of safety for the public, the workers and the environment required outstanding efforts and the international cooperation was of significant importance. The contribution discusses experiences and practical aspects of the international cooperation and implications for future cooperation options for the long-term removal of accident consequences.

  5. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Breakup and cooling of molten material jets. JAERI's nuclear research promotion program, H10-027-2. Contract research

    International Nuclear Information System (INIS)

    Sugiyama, Ken-ichiro; Iguchi, Kentarou

    2002-03-01

    Core melt accidents could lead to the pouring of molten core materials into a body of water accumulating in the reactor lower head in the form of jets with a few centimeters up to a few tens of centimeters. If molten core jets penetrate the body of water without breakup. A poor coolability of the molten core bed would occur, which means the difficulty of maintaining the molten core bed in the reactor vessel. Hence, the breakup mechanism of molten core jets had to be well understood for the evaluation of the coolability of molten core bed. The objective of the present experimental study is to confirm that, even in molten material jets, the breakup of jet originating in the coolant entrained within a molten material jet due to 'the organized motion' between the coolant and the jet, which has been recognized in the field of fluid mechanics, is caused. The first series of experiment was conducted to observe this type of breakup by using molten tin jets up to 25 mm in diameter. Molten tin jet was expected to easily cause this kind of breakup of jet because of a low kinematic viscosity, which means a easy transformation of jet due to the organized motion for the coolant to entrain. The second series of experiment was conducted by using molten copper jet of 25 mm in diameter, of which kinematic viscosity is about same as that of molten UO 2 . The breakup of jet due to the entrainment of the coolant was observed up to high ambient Weber numbers, which cover the atomization regime. The mechanism of the breakup observed in the present study is able to reasonably explain the apparent difference between the breakup lengths of 150 kg-scale corium jets and the breakup lengths of about 8 kg-scale lead-bismuth alloy jets. The breakup by the mechanism reported here also assures a high coolability of molten jets because of an efficient entrainment of coolant within the jet. (author)

  6. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Breakup and cooling of molten material jets. JAERI's nuclear research promotion program, H10-027-2. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Sugiyama, Ken-ichiro; Iguchi, Kentarou [Hokkaido Univ., Graduate School of Engineering, Sapporo, Hokkaido (Japan)

    2002-03-01

    Core melt accidents could lead to the pouring of molten core materials into a body of water accumulating in the reactor lower head in the form of jets with a few centimeters up to a few tens of centimeters. If molten core jets penetrate the body of water without breakup. A poor coolability of the molten core bed would occur, which means the difficulty of maintaining the molten core bed in the reactor vessel. Hence, the breakup mechanism of molten core jets had to be well understood for the evaluation of the coolability of molten core bed. The objective of the present experimental study is to confirm that, even in molten material jets, the breakup of jet originating in the coolant entrained within a molten material jet due to 'the organized motion' between the coolant and the jet, which has been recognized in the field of fluid mechanics, is caused. The first series of experiment was conducted to observe this type of breakup by using molten tin jets up to 25 mm in diameter. Molten tin jet was expected to easily cause this kind of breakup of jet because of a low kinematic viscosity, which means a easy transformation of jet due to the organized motion for the coolant to entrain. The second series of experiment was conducted by using molten copper jet of 25 mm in diameter, of which kinematic viscosity is about same as that of molten UO{sub 2}. The breakup of jet due to the entrainment of the coolant was observed up to high ambient Weber numbers, which cover the atomization regime. The mechanism of the breakup observed in the present study is able to reasonably explain the apparent difference between the breakup lengths of 150 kg-scale corium jets and the breakup lengths of about 8 kg-scale lead-bismuth alloy jets. The breakup by the mechanism reported here also assures a high coolability of molten jets because of an efficient entrainment of coolant within the jet. (author)

  7. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

    Frid, W.

    1991-08-01

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

  8. Expert software for accident identification

    International Nuclear Information System (INIS)

    Dobnikar, M.; Nemec, T.; Muehleisen, A.

    2003-01-01

    Each type of an accident in a Nuclear Power Plant (NPP) causes immediately after the start of the accident variations of physical parameters that are typical for that type of the accident thus enabling its identification. Examples of these parameter are: decrease of reactor coolant system pressure, increase of radiation level in the containment, increase of pressure in the containment. An expert software enabling a fast preliminary identification of the type of the accident in Krsko NPP has been developed. As input data selected typical parameters from Emergency Response Data System (ERDS) of the Krsko NPP are used. Based on these parameters the expert software identifies the type of the accident and also provides the user with appropriate references (past analyses and other documentation of such an accident). The expert software is to be used as a support tool by an expert team that forms in case of an emergency at Slovenian Nuclear Safety Administration (SNSA) with the task to determine the cause of the accident, its most probable scenario and the source term. The expert software should provide initial identification of the event, while the final one is still to be made after appropriate assessment of the event by the expert group considering possibility of non-typical events, multiple causes, initial conditions, influences of operators' actions etc. The expert software can be also used as an educational/training tool and even as a simple database of available accident analyses. (author)

  9. Fukushima accident - reasons and impacts

    International Nuclear Information System (INIS)

    Slugen, V.

    2011-01-01

    The Fukushima accident influenced dramatically the current view on safety of nuclear facilities. Consideration about possible impacts of natural catastrophe in design of nuclear facilities seems to be much more important than before. European commission is focused on the stress-tests at nuclear power plants. His paper will go more in details having in mind reasons and impacts of Fukushima accident (Author)

  10. First Responders and Criticality Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Valerie L. Putman; Douglas M. Minnema

    2005-11-01

    Nuclear criticality accident descriptions typically include, but do not focus on, information useful to first responders. We studied these accidents, noting characteristics to help (1) first responders prepare for such an event and (2) emergency drill planners develop appropriate simulations for training. We also provide recommendations to help people prepare for such events in the future.

  11. Occupational accidents among mototaxi drivers.

    Science.gov (United States)

    Amorim, Camila Rego; de Araújo, Edna Maria; de Araújo, Tânia Maria; de Oliveira, Nelson Fernandes

    2012-03-01

    The use of motorcycles as a means of work has contributed to the increase in traffic accidents, in particular, mototaxi accidents. The aim of this study was to estimate and characterize the incidence of occupational accidents among the mototaxis registered in Feira de Santana, BA. This is a cross-sectional study with descriptive and census data. Of the 300 professionals registered at the Municipal Transportation Service, 267 professionals were interviewed through a structured questionnaire. Then, a descriptive analysis was conducted and the incidence of accidents was estimated based on the variables studied. Relative risks were calculated and statistical significance was determined using the chi-square test and Fisher's exact test, considering p accidents were observed in 10.5% of mototaxis. There were mainly minor injuries (48.7%), 27% of them requiring leaves of absence from work. There was an association between the days of work per week, fatigue in lower limbs and musculoskeletal complaints, and accidents. Knowledge of the working conditions and accidents involved in this activity can be of great importance for the adoption of traffic education policies, and to help prevent accidents by improving the working conditions and lives of these professionals.

  12. Compliance of Bombardier's Challenger 604 and CRJ200 to FAR25.856(a) : flame propagation of thermal/acoustic insulation materials and future trends in aircraft materials fire safety

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, W.R. [Bombardier, Montreal, PQ (Canada); Schofield, C.M.A. [Transport Canada, Ottawa, ON (Canada)

    2007-07-01

    This paper provided details of a testing program designed to ensure the compliance of Bombardier's Challenger 604 to new rules established to improve flammability standards for thermal and acoustic insulation materials. The rule applied to both pressurized and unpressurized sections of the fuselage, as well as to ducting, sound damping foams, and insulation bags. Test samples of all non-metallic insulation components were collected. Testing of the samples was conducted in a chamber with an ignition source as well as a controlled heat flux provided by a radiant panel. The new rules were discussed as well as challenges faced by industry which will have to show compliance for flame propagation requirements. Issues related to the in-service replacement of thermal and acoustic insulation materials were reviewed along with potential changes to flammability regulations. Materials used by Bombardier for compliant constructions were also listed. It was concluded that the safety of airplane occupants will be improved through compliance to new testing methods under harsher fire threat conditions, with more stringent pass and fail criteria. 15 refs., 5 figs.

  13. Emotional reactivity: Beware its involvement in traffic accidents.

    Science.gov (United States)

    M'bailara, Katia; Atzeni, Thierry; Contrand, Benjamin; Derguy, Cyrielle; Bouvard, Manuel-Pierre; Lagarde, Emmanuel; Galéra, Cédric

    2018-04-01

    Reducing risk attributable to traffic accidents is a public health challenge. Research into risk factors in the area is now moving towards identification of the psychological factors involved, particularly emotional states. The aim of this study was to evaluate the link between emotional reactivity and responsibility in road traffic accidents. We hypothesized that the more one's emotional reactivity is disturbed, the greater the likelihood of being responsible for a traffic accident. This case-control study was based on a sample of 955 drivers injured in a motor vehicle crash. Responsibility levels were determined with a standardized method adapted from the quantitative Robertson and Drummer crash responsibility instrument. Emotional reactivity was assessed with the MATHYS. Hierarchical cluster analysis discriminated four distinctive driver's emotional reactivity profiles: basic emotional reactivity (54%), mild emotional hyper-reactivity (29%), emotional hyper-reactivity (11%) and emotional hypo-reactivity (6%). Drivers who demonstrated emotional hypo-reactivity had a 2.3-fold greater risk of being responsible for a traffic accident than those with basic emotional reactivity. Drivers' responsibility in traffic accidents depends on their emotional status. The latter can change the ability of drivers, modifying their behavior and thus increasing their propensity to exhibit risk behavior and to cause traffic accidents. Copyright © 2017 Elsevier B.V. All rights reserved.

  14. Simulation of operator's actions during severe accident management

    International Nuclear Information System (INIS)

    Viktorov, A.

    2015-01-01

    Implementing accident management counter measures or actions to mitigate consequences of a severe accident is essential to reduce radiological risks to the public and environment. Station-specific severe accident management guidelines (SAMGs) have been developed and implemented at all Canadian nuclear power plants. Following the Fukushima Daiichi nuclear accident certain enhancements were introduced to the SAMG, namely consideration of multi-units accidents, events involving spent fuel pools, incorporation of capability offered by the portable emergency mitigating equipment, and so on. To evaluate the adequacy and usability of the SAMGs, CNSC staff initiated a number of activities including a desktop review of SAMG documentation, evaluation of SAMG implementation through exercises and interviews with station staff, and independent verification of SAMG action effectiveness. This paper focuses on the verification of SAMG actions through analytical simulations. The objectives of the work are two-folds: (a) to understand the effectiveness of SAMG-specified mitigation actions in addressing the safety challenges and (b) to check for potential negative effects of the action. Some sensitivity calculations were performed to help understanding of the impact from actions that rely on the partially effective equipment or limited material resources. The severe accident computer code MAAP4-CANDU is used as a tool in this verification. This paper will describe the methodology used in the verification of SAMG actions and some results obtained from simulations. (author)

  15. Accident frequency and unrealistic optimism: Children's assessment of risk.

    Science.gov (United States)

    Joshi, Mary Sissons; Maclean, Morag; Stevens, Claire

    2018-02-01

    Accidental injury is a major cause of mortality and morbidity among children, warranting research on their risk perceptions. Three hundred and seven children aged 10-11 years assessed the frequency, danger and personal risk likelihood of 8 accidents. Two social-cognitive biases were manifested. The frequency of rare accidents (e.g. drowning) was overestimated, and the frequency of common accidents (e.g. bike accidents) underestimated; and the majority of children showed unrealistic optimism tending to see themselves as less likely to suffer these accidents in comparison to their peers, offering superior skills or parental control of the environment as an explanation. In the case of pedestrian accidents, children recognised their seriousness, underestimated the frequency of this risk and regarded their own road crossing skill as protection. These findings highlight the challenging task facing safety educators who, when teaching conventional safety knowledge and routines, also need to alert children to the danger of over-confidence without disabling them though fear. Copyright © 2017 Elsevier Ltd. All rights reserved.

  16. Thermal tests of a transport / Storage cask in buried conditions

    International Nuclear Information System (INIS)

    Yamakawa, H.; Gomi, Y.; Saegusa, T.; Ito, C.

    1998-01-01

    Thermal tests for a hypothetical accident which simulated accidents caused by building collapse in case of an earthquake were conducted using a full-scale dry type transport and storage cask (total heat load: 23 kW). The objectives of these tests were to clarify the heat transfer features of the buried cask under such accidents and the time limit for maintaining the thermal integrity of the cask. Moreover, thermal analyses of the test cask under the buried conditions were carried out on basis of experimental results to establish methodology for the thermal analysis. The characteristics of the test cask are described as well as the test method used. The heat transfer features of the buried cask under such accidents and a time for maintaining the thermal integrity of the cask have been obtained. (O.M.)

  17. Barriers to learning from incidents and accidents

    NARCIS (Netherlands)

    Dechy, N.; Dien, Y.; Drupsteen, L.; Felicio, A.; Cunha, C.; Roed-Larsen, S.; Marsden, E.; Tulonen, T.; Stoop, J.; Strucic, M.; Vetere Arellano, A.L.; Vorm, J.K.J. van der; Benner, L.

    2015-01-01

    This document provides an overview of knowledge concerning barriers to learning from incidents and accidents. It focuses on learning from accident investigations, public inquiries and operational experience feedback, in industrial sectors that are exposed to major accident hazards. The document

  18. Probability of spent fuel transportation accidents

    International Nuclear Information System (INIS)

    McClure, J.D.

    1981-07-01

    The transported volume of spent fuel, incident/accident experience and accident environment probabilities were reviewed in order to provide an estimate of spent fuel accident probabilities. In particular, the accident review assessed the accident experience for large casks of the type that could transport spent (irradiated) nuclear fuel. This review determined that since 1971, the beginning of official US Department of Transportation record keeping for accidents/incidents, there has been one spent fuel transportation accident. This information, coupled with estimated annual shipping volumes for spent fuel, indicated an estimated annual probability of a spent fuel transport accident of 5 x 10 -7 spent fuel accidents per mile. This is consistent with ordinary truck accident rates. A comparison of accident environments and regulatory test environments suggests that the probability of truck accidents exceeding regulatory test for impact is approximately 10 -9 /mile

  19. 29 CFR 1960.29 - Accident investigation.

    Science.gov (United States)

    2010-07-01

    ... reflective of the seriousness of the accident. (b) In any case, each accident which results in a fatality or... evidence uncovered during accident investigations which would be of benefit in developing a new OSHA...

  20. Traffic Accidents on Slippery Roads

    DEFF Research Database (Denmark)

    Fonnesbech, J. K.; Bolet, Lars

    2014-01-01

    Police registrations from 65 accidents on slippery roads in normally Danish winters have been studied. The study showed: • 1 accident per 100 km when using brine spread with nozzles • 2 accidents per 100 km when using pre wetted salt • 3 accidents per 100 km when using kombi spreaders The results...... of accidents in normally Danish winter seasons are remarkable alike the amount of salt used in praxis in the winter 2011/2012. • 2.7 ton NaCl/km when using brine spread with nozzles • 5 ton NaCl/km when using pre wetted salt. • 5.7 ton NaCl/km when using kombi spreaders The explanation is that spreading...

  1. Accident sequence quantification with KIRAP

    International Nuclear Information System (INIS)

    Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong.

    1997-01-01

    The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP's cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs

  2. Accident sequence quantification with KIRAP

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong

    1997-01-01

    The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP`s cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs.

  3. Corporate Cost of Occupational Accidents

    DEFF Research Database (Denmark)

    Rikhardsson, Pall M.; Impgaard, M.

    2004-01-01

    method could be used in all of the companies without revisions. The evaluation of accident cost showed that 2/3 of the costs of occupational accidents are visible in the Danish corporate accounting systems reviewed while 1/3 is hidden from management view. The highest cost of occupational accidents......The systematic accident cost analysis (SACA) project was carried out during 2001 by The Aarhus School of Business and PricewaterhouseCoopers Denmark with financial support from The Danish National Working Environment Authority. Its focused on developing and testing a method for evaluating...... occupational costs of companies for use by occupational health and safety professionals. The method was tested in nine Danish companies within three different industry sectors and the costs of 27 selected occupational accidents in these companies were calculated. One of the main conclusions is that the SACA...

  4. Dose assessment in radiological accidents

    International Nuclear Information System (INIS)

    Donkor, S.

    2013-04-01

    The applications of ionizing radiation bring many benefits to humankind, ranging from power generation to uses in medicine, industry and agriculture. Facilities that use radiation source require special care in the design and operation of equipment to prevent radiation injury to workers or to the public. Despite considerable development of radiation safety, radiation accidents do happen. The purpose of this study is therefore to discuss how to assess doses to people who will be exposed to a range of internal and external radiation sources in the event of radiological accidents. This will go a long way to complement their medical assessment thereby helping to plan their treatment. Three radiological accidents were reviewed to learn about the causes of those accidents and the recommendations that were put in place to prevent recurrence of such accidents. Various types of dose assessment methods were discussed.(au)

  5. Severe accidents in nuclear reactors

    International Nuclear Information System (INIS)

    Ohai, Dumitru; Dumitrescu, Iulia; Tunaru, Mariana

    2004-01-01

    The likelihood of accidents leading to core meltdown in nuclear reactors is low. The consequences of such an event are but so severe that developing and implementing of adequate measures for preventing or diminishing the consequences of such events are of paramount importance. The analysis of major accidents requires sophisticated computation codes but necessary are also relevant experiments for checking the accuracy of the predictions and capability of these codes. In this paper an overview of the severe accidents worldwide with definitions, computation codes and relating experiments is presented. The experimental research activity of severe accidents was conducted in INR Pitesti since 2003, when the Institute jointed the SARNET Excellence Network. The INR activity within SARNET consists in studying scenarios of severe accidents by means of ASTEC and RELAP/SCDAP codes and conducting bench-scale experiments

  6. JCO criticality accident termination operation

    International Nuclear Information System (INIS)

    Kanamori, Masashi

    2001-12-01

    On September 30 at around 10:35 AM, criticality accident occurred at the JCO's conversion building in Tokai-mura. Since criticality accident had not been anticipated, neither devices for termination of criticality accident nor neutron detectors were available. Immediately after the information of the accident, our emergency staff (Japan Nuclear Cycle development institute staff) went to JCO site, to measure the intensity of neutrons and gammas. There were four main tasks, first one was to measure the radiation intensity, second one was to terminate the criticality accident, third one is to alert the residents surrounding the JCO site, fourth one is to evacuate the employees in the site. These tasks were successfully performed until October 1. This paper describes about how these operations were performed by the relevant staffs. (author)

  7. Development of source term PIRT of Fukushima Daiichi NPPs accident

    International Nuclear Information System (INIS)

    Suehiro, S.; Okamoto, K.

    2017-01-01

    The severe accident evaluation committee of AESJ (Atomic Energy Society of Japan) developed the thermal hydraulic PIRT (Phenomena Identification and Ranking Table) and the source term PIRT based on findings during the Fukushima Daiichi NPPs accident. These PIRTs aimed to explore the debris distribution and the current condition in the NPPs with high accuracy and to extract higher priority from the aspect of the sophistication of the analytical technology to predict the severe accident phenomena by the code. The source term PIRT was divided into 3 phases for the time domain and 9 categories for the spatial domain. The 68 phenomena were extracted and the importance from viewpoint of the source term was ranked through brainstorming and discussion. This paper describes the developed source term PIRT list and summarized the high ranked phenomena in each phase. (author)

  8. The Integrated Approach to the Accident Evaluation for Advanced LWRs

    International Nuclear Information System (INIS)

    Oriolo, F.; Paci, S.

    1998-01-01

    The present paper discusses some relevant phenomena occurring in advanced LWRs during postulated accident scenarios. In particular, the operation of ESF is the starting point for analysis of those phenomena that cause the mutual influence between PS and containment in these plants. As a consequence, it is highlighted as accident analyses which treat PS and containment phenomena completely separated may be not adequate when applied to innovate reactors. Exemplified thermal-hydraulic analysis are presented for AP600 and SBWR, using the FUMO integrated model, for highlight accident evolution taking into account these interactions. The architecture of this integrated code is presented highlighting the importance of an integrated approach to the safety analysis in innovative reactors. (author)

  9. Homocysteine and cerebrovascular accidents.

    Science.gov (United States)

    Datta, Saikat; Pal, Salil K; Mazumdar, Hirak; Bhandari, Biswanath; Bhattacherjee, Sharmistha; Pandit, Sudipta

    2009-06-01

    Hyperhomocysteinaemia is rapidly emerging as an important risk factor for coronary artery disease, possibly because of its propensity to accelerate atherosclerosis. Whether it is also a risk factor for cerebrovascular accidents (CVA) is a matter of debate till now, as there are conflicting results of the various prospective studies. The present study was performed to correlate the levels of plasma homocysteine levels with that of ischaemic and haemorrhagic CVA. Forty-two cases of CVA were randomly selected over a period of one year, and their risk factors were assessed. It was observed that serum homocysteine levels were significantly raised in those with intracerebral infarcts when compared to those with intracerebral haemorrhage, although homocysteine levels didn't prove to be prognostically significant.

  10. Ultra-fast photo-patterning of hydroxamic acid layers adsorbed on TiAlN: The challenge of modeling thermally induced desorption

    Energy Technology Data Exchange (ETDEWEB)

    Hemgesberg, Maximilian [Fachbereich Chemie, Technische Universitaet Kaiserslautern, Erwin-Schroedinger-Strasse 54, 67663 Kaiserslautern (Germany); Schuetz, Simon [Fachbereich Physik und Forschungszentrum OPTIMAS, Technische Universitaet Kaiserslautern, Erwin-Schroedinger-Strasse 45, 67663 Kaiserslautern (Germany); Heidelberger Druckmaschinen AG, Kurfuerstenanlage 54-60, 66120 Heidelberg (Germany); Mueller, Christine [Fachbereich Physik und Forschungszentrum OPTIMAS, Technische Universitaet Kaiserslautern, Erwin-Schroedinger-Strasse 45, 67663 Kaiserslautern (Germany); Schloerholz, Matthias; Latzel, Harald [Heidelberger Druckmaschinen AG, Kurfuerstenanlage 54-60, 66120 Heidelberg (Germany); Sun, Yu [Fachbereich Chemie, Technische Universitaet Kaiserslautern, Erwin-Schroedinger-Strasse 54, 67663 Kaiserslautern (Germany); Ziegler, Christiane [Fachbereich Physik und Forschungszentrum OPTIMAS, Technische Universitaet Kaiserslautern, Erwin-Schroedinger-Strasse 45, 67663 Kaiserslautern (Germany); Thiel, Werner R., E-mail: thiel@chemie.uni-kl.de [Fachbereich Chemie, Technische Universitaet Kaiserslautern, Erwin-Schroedinger-Strasse 54, 67663 Kaiserslautern (Germany)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Different hydroxamic acids are suitable for the hydrophobization of TiAlN surfaces used in photo-patterning applications. Black-Right-Pointing-Pointer The binding strength of the surfactant on the surface is dependent on the hydrogen and {pi}-bonding interactions within the organic layer. Black-Right-Pointing-Pointer Thermal desorption of selected HA species results in a 20% energy reduction compared to alkyl phosphates. - Abstract: Long-chain n-alkyl terminated hydroxamic acids (HA) are used for the modification of titanium aluminum nitride (TiAlN) surfaces. HA coatings improve the hydrophobicity of this wear resistant and industrially relevant ceramic. Therefore, HAs with different structural properties are evaluated with respect to their wear resistance and their thermal desorption properties. In order to find new coatings for rewritable offset printing plates, the changes in the surface polarity, composition, and morphology are analyzed by contact angle measurements, X-ray photoemission spectroscopy (XPS), and scanning force microscopy (SFM), respectively. The results are referenced to the strongly bonding molecule n-dodecyl phosphonate (PO11M), which has been used for surface hydrophobization before but proved difficult to remove due to the high laser outputs required for thermal desorption. It is found that for certain HAs, an equally good hydrophobization compared to PO11M can be achieved. Contact angles obtained for different hydroxamic acid coatings can be correlated to their modes of adsorption. Only for selected HA species, resistance to mechanical wear is sufficient for further investigations. Photo-patterning of these hydroxamic acid layers is achieved using a high energy IR laser beam at different energy inputs. Fitting of the obtained data and further evaluation using finite element analysis (FEM) calculations reveal significantly reduced energy consumption of about 20% for the removal of a specific hydroxamic

  11. Severe accident management. Prevention and Mitigation

    International Nuclear Information System (INIS)

    1992-01-01

    Effective planning for the management of severe accidents at nuclear power plants can produce both a reduction in the frequency of such accidents as well as the ability to mitigate their consequences if and when they should occur. This report provides an overview of accident management activities in OECD countries. It also presents the conclusions of a group of international experts regarding the development of accident management methods, the integration of accident management planning into reactor operations, and the benefits of accident management

  12. The Fukushima accident

    International Nuclear Information System (INIS)

    Loria Meneses, Luis Guillermo

    2011-01-01

    The accident happened on March 11, 2011 in the nuclear reactors at the Fukushima plant, Japan, is described. The reactors of the Fukushima plant have been power reactors. The electrical energy is produced by use of the heat released in the fission. Nuclear reactors were affected after of the power outage as a result of the earthquake and the tsunami, and this has kept in operation the refrigeration systems. The japanese reactors have been fission reactors and have used uranium 235 or plutonium 239 as fissionable material. The nuclear reactions of fission are explained. The control of the nuclear reactions at Fukushima was complicated by the decreased of the neutrons absorption and has produced more reactions, generating great amounts of heat. The steam contaminated with the products of fission is produced by to cool the reactor with water. The fissionable material released is dragged until the atmosphere. Radioactive contamination at sites near the reactor was covered in a zone of exclusion with a radius of 30 km. The effects of radioactive contamination in the zone of exclusion are mentioned. The radioactive material from Japan has traveled with the wind in direction toward the north pole. The radioactive cloud has continued until to reach the north Africa and south of Europe. The cloud has approximated to Costa Rica, but the activity of the material found has been less of 0,01 Bq/m3. The Centro de Investigacion en Ciencias Atomicas, Nucleares y Moleculares (Cicanum) has initiated the collection of soil samples, water and earth products to detect part of the radioactive material from the cloud. The Cicanum has had modern equipments to quantify the specific concentrations of radioactive isotope, alpha emitters, beta and gamma, in food, water and milk. The Cicanum has maintained the radiological surveillance of foods after the Chernobyl accident [es

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

    DEFF Research Database (Denmark)

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

    2010-01-01

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

  14. A model for the computation of the thermal processes in the reactor cavity during a severe accident in a LWR, at the presence of sump water, from the time of reactor pressure vessel failure to the start time of melt/concrete interaction

    International Nuclear Information System (INIS)

    Hirschmann, H.

    1990-04-01

    At present no experimental results are available which analyze that stage of a severe accident in a light water reactor, during which the reactor pressure vessel fails by melting, the core debris relocates into the water pool on the floor of the containment building (cavity) and again is heated up. Therefore an analytical model is described, with the help of which the process of material relocation, the heating of the material in the cavity interacting with the pool water, and the production rates of vapour and hydrogen can be estimated. The slumped mass accumulating in the cavity is taken to be the sum of infinitely small mass parts, assumed to slump at different times, which after slumping undergo individual thermal histories. The enthalpy of the slumped mass is the sum of the enthalpies of the single mass parts. The average temperature of the slumped mass is given by the enthalpy computed in this manner. The production rates of the gases are additive superpositions of all partial rates from the mass parts. The gas rates are computed using the balance of enthalpy and mass. (author) 5 refs

  15. Accident management information needs

    International Nuclear Information System (INIS)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R.

    1990-04-01

    The tables contained in this Appendix A describe the information needs for a pressurized water reactor (PWR) with a large, dry containment. To identify these information needs, the branch points in the safety objective trees were examined to decide what information is necessary to (a) determine the status of the safety functions in the plant, i.e., whether the safety functions are being adequately maintained within predetermined limits, (b) identify plant behavior (mechanisms) or precursors to this behavior which indicate that a challenge to plant safety is occurring or is imminent, and (c) select strategies that will prevent or mitigate this plant behavior and monitor the implementation and effectiveness of these strategies. The information needs for the challenges to the safety functions are not examined since the summation of the information needs for all mechanisms associated with a challenge comprise the information needs for the challenge itself

  16. Accident management information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-04-01

    The tables contained in this Appendix A describe the information needs for a pressurized water reactor (PWR) with a large, dry containment. To identify these information needs, the branch points in the safety objective trees were examined to decide what information is necessary to (a) determine the status of the safety functions in the plant, i.e., whether the safety functions are being adequately maintained within predetermined limits, (b) identify plant behavior (mechanisms) or precursors to this behavior which indicate that a challenge to plant safety is occurring or is imminent, and (c) select strategies that will prevent or mitigate this plant behavior and monitor the implementation and effectiveness of these strategies. The information needs for the challenges to the safety functions are not examined since the summation of the information needs for all mechanisms associated with a challenge comprise the information needs for the challenge itself.

  17. Review of accident analyses of RB experimental reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    2003-01-01

    The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VINCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62; yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin) consisting of 2% enriched uranium metal and 80% enriched U0 2 , dispersed in aluminum matrix, have been available since 1962 and 1976, respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements, as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINCA Institute, an independent regulator)' body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety' Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed) to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given. (author)

  18. Review of accident analyses of RB experimental reactor

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2003-01-01

    Full Text Available The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VTNCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62 yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin consisting of 2% enriched uranium metal and 80% enriched UO2 dispersed in aluminum matrix, have been available since 1962 and 1976 respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINĆA Institute, an independent regulatory body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given.

  19. Phenomenological and mechanistic modeling of melt-structure-water interactions in a light water reactor severe accident

    International Nuclear Information System (INIS)

    Bui, V.A.

    1998-01-01

    The objective of this work is to address the modeling of the thermal hydrodynamic phenomena and interactions occurring during the progression of reactor severe accidents. Integrated phenomenological models are developed to describe the accident scenarios, which consist of many processes, while mechanistic modeling, including direct numerical simulation, is carried out to describe separate effects and selected physical phenomena of particular importance

  20. A severe accident analysis for the system-integrated modular advanced reactor

    International Nuclear Information System (INIS)

    Jung, Gunhyo; Jae, Moosung

    2015-01-01

    The System-Integrated Modular Advanced Reactor (SMART) that has been recently designed in KOREA and has acquired standard design certification from the nuclear power regulatory body (NSSC) is an integral type reactor with 330MW thermal power. It is a small sized reactor in which the core, steam generator, pressurizer, and reactor coolant pump that are in existing pressurized light water reactors are designed to be within a pressure vessel without any separate pipe connection. In addition, this reactor has much different design characteristics from existing pressurized light water reactors such as the adoption of a passive residual heat removal system and a cavity flooding system. Therefore, the safety of the SMART against severe accidents should be checked through severe accident analysis reflecting the design characteristics of the SMART. For severe accident analysis, an analysis model has been developed reflecting the design information presented in the standard design safety analysis report. The severe accident analysis model has been developed using the MELCOR code that is widely used to evaluate pressurized LWR severe accidents. The steady state accident analysis model for the SMART has been simulated. According to the analysis results, the developed model reflecting the design of the SMART is found to be appropriate. Severe accident analysis has been performed for the representative accident scenarios that lead to core damage to check the appropriateness of the severe accident management plan for the SMART. The SMART has been shown to be safe enough to prevent severe accidents by utilizing severe accident management systems such as a containment spray system, a passive hydrogen recombiner, and a cavity flooding system. In addition, the SMART is judged to have been technically improved remarkably compared to existing PWRs. The SMART has been designed to have a larger reactor coolant inventory compared to its core's thermal power, a large surface area in

  1. Effect of In-Vessel Retention Strategies under Postulated SGTR Accidents of OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Wonjun; Lee, Yongjae; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of); Kim, Hwan-Yeol; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, MELCOR code was used to simulate the severe accident of the OPR1000. MELCOR code is computer code which enables to simulate the progression of the severe accident for light water reactors. It has been developed by Sandia National Laboratories for plant risk assessment and source term analysis since 1982. According to the probabilistic safety analysis (PSA) Level 1 of OPR1000, typical severe accident scenarios of high probability of a transition to severe accident for OPR1000 were identified as Small Break Loss of Coolant Accident (SBLOCA), Station Black out (SBO), Total Loss of Feed Water (TLOFW), and Steam Generator Tube Rupture. While the first three accidents are expected to result in the generation and transportation of the radioactive nuclides within the containment building as consequence of the core damage and subsequent reactor pressure vessel (RPV) failure, the latter accident scenario may be progressed with possible direct release of the radioactive nuclides to the environment by bypassing the containment building. Thus it is of significance to investigate the SGTR accident with a sophisticated severe accident code. This code can simulate the whole phenomena of a severe accident such as thermal-hydraulic response, core heat-up, oxidation and relocation, and fission product release and transport. Thus many researchers have used MELCOR in severe accident studies. In this study, in-vessel retention strategies were applied for postulated SGTR accidents. Mitigation effect and adverse effect of in-vessel strategies was studied in aspect of RPV failure, fission product release and containment thermal-hydraulic and hydrogen behavior. Base case of SGTR accident and three mitigation cases were simulated using MELCOR code 1.8.6. For each mitigation cases, mitigation effect and adverse effect were investigated. Conclusions can be summarized as follows: (1) RPV failure of SGTR base case occurred at 5.62 hours and fission product of RCS released to

  2. Insights from Severe Accident Analyses for Verification of VVER SAMG

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A. J.; Rao, R. S.; Gupta, A.; Obaidurrahaman, K., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

    2014-10-15

    The severe accident analyses of simultaneous rupture of all four steam lines (case-a), simultaneous occurrence of LOCA with SBO (case-b) and Station blackout (case-c) were performed with the computer code ASTEC V2r2 for a typical VVER-1000. The results obtained will be used for verification of sever accident provisions and Severe Accident Management Guidelines (SAMG). Auxiliary feed water and emergency core cooling systems are modelled as boundary conditions. The ICARE module is used to simulate the reactor core, which is divided into five radial regions by grouping similarly powered fuel assemblies together. Initially, CESAR module computes thermal hydraulics in primary and secondary circuits. As soon as core uncovery begins, the ICARE module is actuated based on certain parameters, and after this, ICARE module computes the thermal hydraulics in the core, bypass, downcomer and the lower plenum. CESAR handles the remaining components in the primary and secondary loops. CPA module is used to simulate the containment and to predict the thermal-hydraulic and hydrogen behaviour in the containment. The accident sequences were selected in such a way that they cover low/high pressure and slow/fast core damage progression events. Events simulated included slow progression events with high pressure and fast accident progression with low primary pressure. Analysis was also carried out for the case of SBO with the opening of the PORVs when core exit temperature exceeds certain value as part of SAMG. Time step sensitivity study was carried out for LOCA with SBO. In general the trends and magnitude of the parameters are as expected. The key results of the above analyses are presented in this paper. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

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

  4. Construction industry accidents in Spain.

    Science.gov (United States)

    Camino López, Miguel A; Ritzel, Dale O; Fontaneda, Ignacio; González Alcantara, Oscar J

    2008-01-01

    This paper analyzed industrial accidents that take place on construction sites and their severity. Eighteen variables were studied. We analyzed the influence of each of these with respect to the severity and fatality of the accident. This descriptive analysis was grounded in 1,630,452 accidents, representing the total number of accidents suffered by workers in the construction sector in Spain over the period 1990-2000. It was shown that age, type of contract, time of accident, length of service in the company, company size, day of the week, and the remainder of the variables under analysis influenced the seriousness of the accident. IMPACT ON INJURY PREVENTION: The results obtained show that different training was needed, depending on the severity of accidents, for different age, length of service in the company, organization of work, and time when workers work. The research provides an insight to the likely causes of construction injuries in Spain. As a result of the analysis, industries and governmental agencies in Spain can start to provide appropriate strategies and training to the construction workers.

  5. Contributing factors in construction accidents.

    Science.gov (United States)

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

    2005-07-01

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

  6. The early medical response to the Goiania accident

    International Nuclear Information System (INIS)

    Valverde, N.J.; Oliveira, A.R.

    2000-01-01

    The Goiania accident was the most severe radiological one that ever happened in the western hemisphere. The response to its human, social, environmental, economical and psychological burdens represented a huge challenge. Thanks to a multi-institutional intervention the consequences of the accident were greatly minimised. The medical response followed the same pattern and was based on a three-level system of progressive assistance. The early medical response encompassed medical and 'radiological' triage, admission to a specially prepared ward of a local hospital and treatment at a reference center in Rio de Janeiro. (author)

  7. International aspects of nuclear accidents

    International Nuclear Information System (INIS)

    Uematsu, K.

    1989-09-01

    The accident at Chernobyl revealed that there were shortcomings and gaps in the existing international mechanisms and brought home to governments the need for stronger measures to provide better protection against the risks of severe accidents. The main thrust of international co-operation with regard to nuclear safety issues is aimed at achieving a uniformly high level of safety in nuclear power plants through continuous exchanges of research findings and feedback from reactor operating experience. The second type of problem posed in the event of an accident resulting in radioactive contamination of several countries relates to the obligation to notify details of the circumstances and nature of the accident speedily so that the countries affected can take appropriate protective measures and, if necessary, organize mutual assistance. Giving the public accurate information is also an important aspect of managing an emergency situation arising from a severe accident. Finally, the confusion resulting from the unwarranted variety of protective measures implemented after the Chernobyl accident has highlighted the need for international harmonization of the principles and scientific criteria applicable to the protection of the public in the event of an accident and for a more consistent approach to emergency plans. The international conventions on third party liability in the nuclear energy sector (Paris/Brussels Conventions and the Vienna Convention) provide for compensation for damage caused by nuclear accidents in accordance with the rules and jurisdiction that they lay down. These provisions impose obligations on the operator responsible for an accident, and the State where the nuclear facility is located, towards the victims of damage caused in another country

  8. 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

  9. Thermal testing transport packages for radioactive materials: Reality vs regulation

    International Nuclear Information System (INIS)

    Hovingh, J.; Carlson, R.W.

    1994-03-01

    The principle objective of this paper is to provide information that will help describe the physical thermal tests performed to demonstrate compliance with the hypothetical accident conditions specified in 10 CFR 71.73. Physical testing should be applied to packages that cannot be modeled by analysis to adequately predict their response to hypothetical accident conditions. These tests should be used when chemical decomposition or material changes occur during an accident that would be difficult to analytically predict or model

  10. Accident termination by element dropout in the GCFR

    International Nuclear Information System (INIS)

    Torri, A.; Tomkins, J.L.

    1976-01-01

    Severe loss-of-flow accidents are being investigated for the GCFR in order to assess the risk from those low-probability accidents which lead to a loss of coolable core geometry. Accident mitigating phenomena unique to the GCFR have been identified for the loss of decay heat removal accident. Circumferential assembly duct melting is calculated to occur at the core mid-plane before the fuel within the assembly melts. The GCFR core assemblies are top-mounted and there is clearance between assemblies to accommodate swelling and thermal distortions without interference. No lateral core clamping system is employed and there are no structures in the plenum below the core. Thus it is possible for the lower portion of the individual assemblies, including most of the fuel, to drop to the cavity floor unless interference or bonding between assemblies develops during the accident. Due to the delay in duct corner melting the melt front at the duct mid-flat progresses over about one-half of the core height. The possibility of inter-element bonding by molten duct steel dislocated into the gap between assemblies has been recognized and a test program to verify the duct melting sequence and to investigate the duct dropout is being planned at the Los Alamos Scientific Laboratory

  11. Effect of Meteorological Parameters on Accident Rates in Petrochemical Industries

    International Nuclear Information System (INIS)

    Mansouri, N.; Farsi, E.

    2016-01-01

    Background and Objective: In this research the effectiveness of weather and climate parameters in incidence of accidents in the petrochemical industry was studied and management strategies to prevent these events have been presented. Method: Two of the petrochemical companies, one of them in Assaluyeh (named Zagros, located in warm climates) and the other one in Tabriz (in cold climates) were selected for pilot study. The required data were collected by questionnaire, interview and walking through under study fields. The analyses of data have been done by Excel, SPSS software and Correlation statistical test. Findings: Climate parameters don’t have a directly impact on the petrochemical occupational accidents and there is no significant relationship between them. Discussion and Conclusion: The role of climatic parameters in the incidence of accidents in the petrochemical industry is indirect. In fact, the thermal stress in the first stage caused unsafe conditions and then unsafe behavior, and finally cause human error and occupational accidents. In this study, appropriate solutions for instance engineering or managerial measures are also suggested in order to prevent accidents and injuries.

  12. EAC european accident code. A modular system of computer programs to simulate LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Wider, H.; Cametti, J.; Clusaz, A.; Devos, J.; VanGoethem, G.; Nguyen, H.; Sola, A.

    1985-01-01

    One aspect of fast reactor safety analysis consists of calculating the strongly coupled system of physical phenomena which contribute to the reactivity balance in hypothetical whole-core accidents: these phenomena are neutronics, fuel behaviour and heat transfer together with coolant thermohydraulics in single- and two-phase flow. Temperature variations in fuel, coolant and neighbouring structures induce, in fact, thermal reactivity feedbacks which are added up and put in the neutronics calculation to predict the neutron flux and the subsequent heat generation in the reactor. At this point a whole-core analysis code is necessary to examine for any hypothetical transient whether the various feedbacks result effectively in a negative balance, which is the basis condition to ensure stability and safety. The European Accident Code (EAC), developed at the Joint Research Centre of the CEC at Ispra (Italy), fulfills this objective. It is a modular informatics structure (quasi 2-D multichannel approach) aimed at collecting stand-alone computer codes of neutronics, fuel pin mechanics and hydrodynamics, developed both in national laboratories and in the JRC itself. EAC makes these modules interact with each other and produces results for these hypothetical accidents in terms of core damage and total energy release. 10 refs

  13. Accident Analysis and Highway Safety

    Directory of Open Access Journals (Sweden)

    Omar Noorliyana

    2017-01-01

    Full Text Available Since 2010, Federal Route FT050 (Jalan Batu Pahat-Kluang has undergone many changes, including the improvement of geometric features (i.e., construction of median, dedicated U-turns and additional lanes and upgrading the quality of the road surface. Unfortunately, even with these enhancements, accidents continue to occur along this route. This study covered both accident analysis and blackspot study. Accident point weightage was used to identify blackspot locations. The results reveal hazardous road locations and blackspot ranking along the route.

  14. Medical care of radiation accidents

    International Nuclear Information System (INIS)

    Nakao, Isamu

    1986-02-01

    This monograph, divided into six chapters, focuses on basic knowledge and medical strategies for radiation accidents. Chapters I to V deal with practice in emergency care for radiation exposure, covering 1) medical strategies for radiation accidents, 2) personnel dosimetry and monitoring, 3) nuclear facilities and their surrounding areas with the potential for creating radiation accidents, and emergency medical care for exposed persons, 4) emergency care procedures for radiation exposure and radioactive contamination, and 5) radiation hazards and their treatment. The last chapter provides some references. (Namekawa, K.)

  15. Judicial autopsy of radiation accidents

    International Nuclear Information System (INIS)

    Kannan, P.M.

    1990-01-01

    This paper discusses issues regarding the judicial autopsy of radiation accidents. In the litigation which follows a radiation accident, a claimant calls on the legal system to adjudicate a dispute. Scientific questions are thrust upon the court. The legal system (through attorneys for the parties) then invites scientists to assist the court in resolving such questions. The invitation, however, does not allow the scientist to bring along his full kit. Experimentation, such as repeating the accident with dosimeters to gather more accurate data, is generally not allowed. Also, the scientist must give up his practice of choosing which questions he will pursue

  16. A review of criticality accidents

    International Nuclear Information System (INIS)

    Stratton, W.R.; Smith, D.R.

    1989-03-01

    Criticality accidents and the characteristics of prompt power excursions are discussed. Forty-one accidental power transients are reviewed. In each case where available, enough detail is given to help visualize the physical situation, the cause or causes of the accident, the history and characteristics of the transient, the energy release, and the consequences, if any, to personnel and property. Excursions associated with large power reactors are not included in this study, except that some information on the major accident at the Chernobyl reactor in April 1986 is provided in the Appendix. 67 refs., 21 figs., 2 tabs

  17. Analysis of human error in occupational accidents in the power plant industries using combining innovative FTA and meta-heuristic algorithms

    OpenAIRE

    M. Omidvari; M. R. Gharmaroudi

    2015-01-01

    Introduction: Occupational accidents are of the main issues in industries. It is necessary to identify the main root causes of accidents for their control. Several models have been proposed for determining the accidents root causes. FTA is one of the most widely used models which could graphically establish the root causes of accidents. The non-linear function is one of the main challenges in FTA compliance and in order to obtain the exact number, the meta-heuristic algorithms can be used. ...

  18. Challenges for nuclear regulators

    International Nuclear Information System (INIS)

    Allen, M.R.

    1994-01-01

    The paper discusses issues that will present particular challenges to the nuclear industry, during the next decade, that should be addressed in order to achieve increased public acceptance of the industry. The issues of safety culture, peer review of safety, severe accidents, transport accidents and public communication are suggest as meriting special consideration. The need for nuclear regulatory agencies to establish credibility and for operating organisations to demonstrate accountability is emphasised, and the Australian experiences are reviewed. Increased collaboration by nuclear regulatory agencies in the Asian-pacific region is suggested, in order to encourage consistency of regulatory safety standards for the nuclear safety issues identified. Methods whereby this could be achieved within the framework of existing agreements and arrangements are suggested. 2 refs

  19. 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.

  20. Lifetime extension: what challenges for industrialists?

    International Nuclear Information System (INIS)

    Cudelou, C.

    2014-01-01

    Lifetime extension upgrading imposes to compel with ASN (Nuclear Safety Authority) requirements such as: -) to assess the conformity of the installation in a very broad sense by looking for and processing any discrepancy, -) to demonstrate that the facility can face an accident, -) to justify an efficient monitoring of the components that can not be replaced (reactor vessel, containment building), -) to anticipate the replacement of important components like steam generators, -) to study measures for diminishing the thermal and chemical stresses of components, and -) to maintain knowledge and technical competencies. All these requirements are a big challenge for the nuclear industry that will face in the few years to come important tasks. Between 2015 and 2020, it is expected that fifteen 900 MW units and twenty 1300 MW units will have their third decennial outage program. During this period there will also be the second decennial outage program for some 1450 MW units and the first fourth decennial outage program for a 900 MW unit. To succeed in all these challenges, nuclear industry will have to focus on innovation and on a better work organisation. (A.C.)

  1. Accident response in France

    International Nuclear Information System (INIS)

    Duco, J.; L'Homme, A.; Queniart, D.

    1988-07-01

    French PWR power plant design relies basically on a deterministic approach. A probabilistic approach was introduced in France in the early seventies to define safety provisions against external impacts. In 1977 an overall safety objective was issued by the safety authority in terms of an upper probability limit for having unacceptable consequences. Additional measures were taken (the ''H'' operating procedures) to complement the automatic systems normally provided by the initial design, so as to safisfy the safety objective. The TMI-2 accident enhanced the interest in confused situations in which possible multiple equipment failure and/or unappropriate previous actions of the operators impede the implementation of any of the existing event-oriented procedures. In such situations, the objective becomes to avoid core-melt by any means available: this is the goal of the Ul symptom-oriented procedure. Whenever a core-melt occurs, the radioactive releases into the environment must be compatible with the feasibility of the off-site emergency plans; that means that for some hypothetical, but still conceivable scenarios, provisions have to be made to delay and limit the consequences of the loss of the containment: the U2, U4 and U5 ultimate procedures have been elaborated for that purpose. For the case of an emergency, a nationwide organization has been set up to provide the plant operator with a redundant technical expertise, to help him save his plant or mitigate the radiological consequences of a core-melt

  2. RENEB accident simulation exercise.

    Science.gov (United States)

    Brzozowska, Beata; Ainsbury, Elizabeth; Baert, Annelot; Beaton-Green, Lindsay; Barrios, Leonardo; Barquinero, Joan Francesc; Bassinet, Celine; Beinke, Christina; Benedek, Anett; Beukes, Philip; Bortolin, Emanuela; Buraczewska, Iwona; Burbidge, Christopher; De Amicis, Andrea; De Angelis, Cinzia; Della Monaca, Sara; Depuydt, Julie; De Sanctis, Stefania; Dobos, Katalin; Domene, Mercedes Moreno; Domínguez, Inmaculada; Facco, Eva; Fattibene, Paola; Frenzel, Monika; Monteiro Gil, Octávia; Gonon, Géraldine; Gregoire, Eric; Gruel, Gaëtan; Hadjidekova, Valeria; Hatzi, Vasiliki I; Hristova, Rositsa; Jaworska, Alicja; Kis, Enikő; Kowalska, Maria; Kulka, Ulrike; Lista, Florigio; Lumniczky, Katalin; Martínez-López, Wilner; Meschini, Roberta; Moertl, Simone; Moquet, Jayne; Noditi, Mihaela; Oestreicher, Ursula; Orta Vázquez, Manuel Luis; Palma, Valentina; Pantelias, Gabriel; Montoro Pastor, Alegria; Patrono, Clarice; Piqueret-Stephan, Laure; Quattrini, Maria Cristina; Regalbuto, Elisa; Ricoul, Michelle; Roch-Lefevre, Sandrine; Roy, Laurence; Sabatier, Laure; Sarchiapone, Lucia; Sebastià, Natividad; Sommer, Sylwester; Sun, Mingzhu; Suto, Yumiko; Terzoudi, Georgia; Trompier, Francois; Vral, Anne; Wilkins, Ruth; Zafiropoulos, Demetre; Wieser, Albrecht; Woda, Clemens; Wojcik, Andrzej

    2017-01-01

    The RENEB accident exercise was carried out in order to train the RENEB participants in coordinating and managing potentially large data sets that would be generated in case of a major radiological event. Each participant was offered the possibility to activate the network by sending an alerting email about a simulated radiation emergency. The same participant had to collect, compile and report capacity, triage categorization and exposure scenario results obtained from all other participants. The exercise was performed over 27 weeks and involved the network consisting of 28 institutes: 21 RENEB members, four candidates and three non-RENEB partners. The duration of a single exercise never exceeded 10 days, while the response from the assisting laboratories never came later than within half a day. During each week of the exercise, around 4500 samples were reported by all service laboratories (SL) to be examined and 54 scenarios were coherently estimated by all laboratories (the standard deviation from the mean of all SL answers for a given scenario category and a set of data was not larger than 3 patient codes). Each participant received training in both the role of a reference laboratory (activating the network) and of a service laboratory (responding to an activation request). The procedures in the case of radiological event were successfully established and tested.

  3. Accident risk. Chapter 5

    International Nuclear Information System (INIS)

    1978-01-01

    Following a historical introduction in which WASH-740, WASH-1400, Swedish, Finnish, Scandinavian, Netherlands and West German analyses are briefly presented, the concept of risk itself is discussed, distinguishing between objective and subjective aspects, and between voluntary and involuntary risk. Risk analysis is briefly described and an attempt made to define acceptable risk. In treating the safety philosophy of nuclear power plants the engineered safety precautions are presented. The numerical results of the analysis made for Norwegian conditions are presented and discussed. Underground siting, which has been much discussed in Norway is also treated, and emergency planning briefly discussed. The probability and consequences of core meltdown in a light water reactor are then discussed, and the possible faults leading to this, both internal, human errors and external impacts are analysed. The failure mechanisms in the containment building which could lead to the release of activity are discussed, followed by the dispersion of the activity and the health and economic consequences. The accidents at Wuergassen and Brown's Ferry are briefly described as examples. A brief discussion of nuclear insurance and nuclear law in Norway form the concluding sections. (JIW)

  4. Analysing The Thermalhydraulic Parameters Of VVER-1000 Reactor For The Accident Of Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Luu Nam Hai; Truong Cong Thang

    2011-01-01

    To ensure the safety operation of nuclear power plant (NPP), a lot of postulated accident scenarios were considered and analysed. This research chose and analysed the accident of steam generator tube rupture (SGTR) under the actual plant conditions by using the simulation program PCTRAN. The SGTR accident is happen when the NPP is under operation with the steady state condition (power of 3000 MWth, primary pressure of 157 bar and secondary pressure of 63 bar). The accident is initiated by creating a break with equivalent diameter of 100 mm in the area of lower row heat exchanging tubes. The result of analysis is compared with the calculation of the Shariz University, Iran using the thermal hydraulics code RELAP5/mod3.2 and the report in the PSAR data of VVER-1000. This comparison shows that it is possible for using PCTRAN to analyse accidents of VVER-1000 reactor. (author)

  5. Radioactivity control after Fukushima accident

    International Nuclear Information System (INIS)

    Vukovic, D.; Mitrovic, R.; Vicentijevic, M.; Pantelic, G.

    2011-01-01

    Fukushima nuclear accident has influence on more attention when radioactivity of fish were controlled. Sea fish, freshwater fish, fish products and fish flour were analysed ( 95 samples). All products were safe for use with radiation-hygienic aspects. [sr

  6. The Goiania accident - environmental survey

    International Nuclear Information System (INIS)

    Godoy, J.M.; Moreira, M.C.F.; Fonseca, E.S. da

    1997-01-01

    The survey methods applied during the Goiania accident could be considered complementary one to the other, and were able to give a clear picture about the contamination in the city to guide the further decontamination works. (author)

  7. Severe accident recriticality analyses (SARA)

    DEFF Research Database (Denmark)

    Frid, W.; Højerup, C.F.; Lindholm, I.

    2001-01-01

    with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality-both super-prompt power bursts and quasi steady-state power......Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies......, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g(-1), was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s(-1). In most cases, however, the predicted energy deposition was smaller, below...

  8. Multidisciplinary accident investigation : volume 2

    Science.gov (United States)

    1976-05-01

    The Task II final report for 1974 of the Multidisciplinary Accident Investigation : Team of the Maryland Medical-Legal Foundation, Inc. is presented. This report describes some preliminary findings emanating from a series of comprehensive, multivaria...

  9. Multidisciplinary accident investigation : volume 1

    Science.gov (United States)

    1976-09-01

    The final report of the Multidisciplinary Accident Investigation Team of the Maryland Medical-Legal Foundation, Inc. is presented. The report describes the methodology, results, discussions, conclusions and recommendations pertaining to the investiga...

  10. [Early management of cerebrovascular accidents].

    Science.gov (United States)

    Libot, Jérômie; Guillon, Benoit

    2013-01-01

    A cerebrovascular accident requires urgent diagnosis and treatment.The management of a stroke must be early and adapted in order to improve the overall clinical outcome and lower the risk of mortality.

  11. 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...

  12. Preparedness against nuclear power accidents

    International Nuclear Information System (INIS)

    1985-01-01

    This booklet contains information about the organization against nuclear power accidents, which exist in the four Swedish counties with nuclear power plants. It is aimed at classes 7-9 of the Swedish schools. (L.E.)

  13. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  14. Porosity effects during a severe accident

    International Nuclear Information System (INIS)

    Cazares R, R. I.; Espinosa P, G.; Vazquez R, A.

    2015-09-01

    The aim of this work is to study the behaviour of porosity effects on the temporal evolution of the distributions of hydrogen concentration and temperature profiles in a fuel assembly where a stream of steam is flowing. The analysis considers the fuel element without mitigation effects. The mass transfer phenomenon considers that the hydrogen generated diffuses in the steam by convection and diffusion. Oxidation of the cladding, rods and other components in the core constructed in zirconium base alloy by steam is a critical issue in LWR accident producing severe core damage. The oxygen consumed by the zirconium is supplied by the up flow of steam from the water pool below the uncovered core, supplemented in the case of PWR by gas recirculation from the cooler outer regions of the core to hotter zones. Fuel rod cladding oxidation is then one of the key phenomena influencing the core behavior under high-temperature accident conditions. The chemical reaction of oxidation is highly exothermic, which determines the hydrogen rate generation and the cladding brittleness and degradation. The heat transfer process in the fuel assembly is considered with a reduced order model. The Boussinesq approximation was applied in the momentum equations for multicomponent flow analysis that considers natural convection due to buoyancy forces, which is related with thermal and hydrogen concentration effects. The numerical simulation was carried out in an averaging channel that represents a core reactor with the fuel rod with its gap and cladding and cooling steam of a BWR. (Author)

  15. Porosity effects during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Cazares R, R. I. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Posgrado en Energia y Medio Ambiente, San Rafael Atlixco 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Espinosa P, G.; Vazquez R, A., E-mail: ricardo-cazares@hotmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, San Rafael Atlixco 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2015-09-15

    The aim of this work is to study the behaviour of porosity effects on the temporal evolution of the distributions of hydrogen concentration and temperature profiles in a fuel assembly where a stream of steam is flowing. The analysis considers the fuel element without mitigation effects. The mass transfer phenomenon considers that the hydrogen generated diffuses in the steam by convection and diffusion. Oxidation of the cladding, rods and other components in the core constructed in zirconium base alloy by steam is a critical issue in LWR accident producing severe core damage. The oxygen consumed by the zirconium is supplied by the up flow of steam from the water pool below the uncovered core, supplemented in the case of PWR by gas recirculation from the cooler outer regions of the core to hotter zones. Fuel rod cladding oxidation is then one of the key phenomena influencing the core behavior under high-temperature accident conditions. The chemical reaction of oxidation is highly exothermic, which determines the hydrogen rate generation and the cladding brittleness and degradation. The heat transfer process in the fuel assembly is considered with a reduced order model. The Boussinesq approximation was applied in the momentum equations for multicomponent flow analysis that considers natural convection due to buoyancy forces, which is related with thermal and hydrogen concentration effects. The numerical simulation was carried out in an averaging channel that represents a core reactor with the fuel rod with its gap and cladding and cooling steam of a BWR. (Author)

  16. Alternative protections for loss of coolant accidents

    International Nuclear Information System (INIS)

    Estevez, E.A.

    1997-01-01

    One way to mitigate a small loss of coolant accident (LOCA) is by depressurizing the primary system, in order to turn the accident into a sequence where water is fed to a low pressure system. It can be achieved by two different ways: by incorporating a valve system (ADS - Automatic Depressurization System) to the design, which helps to diminish the pressure, obtaining a bigger LOCA, or by extracting heat from the system. Our analysis is centered in integrated reactors. The first characterization performed was on CAREM reactor. The idea was then to observe its behavior with LOCAs for different thermal power relations, water volume and rupture area. A simple depressurization model is presented, which enables us to find the parameter relationships which characterize this process, from which some particular cases will arise. ADS implementation is then analyzed, giving the criteria for the triggering time. A study on its reliability and the probability of a spurious opening is made, taking into account independent and dependent failures. An analysis on heat extraction as alternative for depressurizing is also made. Finally, the different reasons to choose between ADS or heat extraction as alternative are given, and the meaning of the parameters found are discussed. An alternative to classify LOCAs, instead of the traditional classification, by fracture size, is suggested. (author)

  17. Experimental Setup for Reflood Quench of Accident Tolerant Fuel Claddings

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan; Lee, Kwan Geun; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The concept of accident tolerant fuel (ATF) is a solution to suppress the hydrogen generation in loss of coolant accident (LOCA) situation without safety injection, which was the critical incident in the severe accident in the Fukushima. The changes in fuel and cladding materials may cause a significant difference in reactor performance in long term operation. Properties in terms of material science and engineering have been tested and showed promising results. However, numerous tests are still required to ensure the design performance and safety. Thermal hydraulic tests including boiling and quenching are partly confirmed, but not yet complete. We have been establishing the experimental setup to confirm the properties in the terms of thermal hydraulics. Design considerations and preliminary tests are introduced in this paper. An experimental setup to test thermal hydraulic characteristics of new ATF claddings are established and tested. The W heater set inside the cladding is working properly, exceeding 690 W/m linear power with thermocouples and insulating ceramic sheaths inside. The coolant injection control was also working in good conditions. The setup is about to complete and going to simulate quenching behavior of the ATF in the LOCA situation.

  18. Power Excursion Accident Analysis of Research Water Reactor

    International Nuclear Information System (INIS)

    Khaled, S.M.; Doaa, G.M.

    2009-01-01

    A three-dimensional neutronic code POWEX-K has been developed, and it has been coupled with the sub-channel thermal-hydraulic core analysis code SV based on the Single Mass Velocity Model. This forms the integrated neutronic/thermal hydraulics code system POWEX-K/SV for the accident analysis. The Training and Research Reactors at Budapest University of Technology and Economics (BME-Reactor) has been taken as a reference reactor. The cross-section generation procedure based on WIMS. The code uses an implicit difference approach for both the diffusion equations and thermal-hydraulics modules, with reactivity feedback effects due to coolant and fuel temperatures. The code system was applied to analyzing power excursion accidents initiated by ramp reactivity insertion of 1.2 $. The results show that the reactor is inherently safe in case of such accidents i.e. no core melt is expected even if the safety rods do not fall into the core

  19. Numerical module for debris behavior under severe accident conditions

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  20. Accident risk and factors regarding non-motorised road users

    DEFF Research Database (Denmark)

    Agerholm, Niels; Andersen, Camilla Sloth

    2015-01-01

    Almost half of all traffic fatalities worldwide are non-motorised road users (NMRUs). In Denmark, the number has increased with about 30%. NMRUs consist of about 63% of the injured in the Danish traffic. Much has been done to reduce the number of injured NMRUs with counterparts, while little effort...... is put into the reduction of the vast majority of the accidents, NMRU single accidents, which are about 90% of all injured NMRUs. There are no efficient tools available to reduce this number. A significantly better designed, maintained, and illuminated road network would most likely help. However......, that is expensive and not possible for most road authorities. Despite this, the challenges with NMRUs in single accidents need more attention, if road safety is to be improved. The situation in Denmark is more than likely the case in many other countries as well; although the documentation is scarce....