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Sample records for accident consequence code

  1. Accident consequence assessment code development

    International Nuclear Information System (INIS)

    Homma, T.; Togawa, O.

    1991-01-01

    This paper describes the new computer code system, OSCAAR developed for off-site consequence assessment of a potential nuclear accident. OSCAAR consists of several modules which have modeling capabilities in atmospheric transport, foodchain transport, dosimetry, emergency response and radiological health effects. The major modules of the consequence assessment code are described, highlighting the validation and verification of the models. (author)

  2. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T.; Rollstin, J.A.; Chanin, D.I.

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs

  3. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    Rollstin, J.A.; Chanin, D.I.; Jow, H.N.

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projections, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management

  4. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T. (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA))

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.

  5. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    Chanin, D.I.; Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems

  6. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

    Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian (Sandia National Labs., Albuquerque, NM (USA))

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

  7. Health effects estimation code development for accident consequence analysis

    International Nuclear Information System (INIS)

    Togawa, O.; Homma, T.

    1992-01-01

    As part of a computer code system for nuclear reactor accident consequence analysis, two computer codes have been developed for estimating health effects expected to occur following an accident. Health effects models used in the codes are based on the models of NUREG/CR-4214 and are revised for the Japanese population on the basis of the data from the reassessment of the radiation dosimetry and information derived from epidemiological studies on atomic bomb survivors of Hiroshima and Nagasaki. The health effects models include early and continuing effects, late somatic effects and genetic effects. The values of some model parameters are revised for early mortality. The models are modified for predicting late somatic effects such as leukemia and various kinds of cancers. The models for genetic effects are the same as those of NUREG. In order to test the performance of one of these codes, it is applied to the U.S. and Japanese populations. This paper provides descriptions of health effects models used in the two codes and gives comparisons of the mortality risks from each type of cancer for the two populations. (author)

  8. An uncertainty analysis using the NRPB accident consequence code Marc

    International Nuclear Information System (INIS)

    Jones, J.A.; Crick, M.J.; Simmonds, J.R.

    1991-01-01

    This paper describes an uncertainty analysis of MARC calculations of the consequences of accidental releases of radioactive materials to atmosphere. A total of 98 parameters describing the transfer of material through the environment to man, the doses received, and the health effects resulting from these doses, was considered. The uncertainties in the numbers of early and late health effects, numbers of people affected by countermeasures, the amounts of food restricted and the economic costs of the accident were estimated. This paper concentrates on the results for early death and fatal cancer for a large hypothetical release from a PWR

  9. Evaluation of severe accident risks: Quantification of major input parameters: MAACS [MELCOR Accident Consequence Code System] input

    International Nuclear Information System (INIS)

    Sprung, J.L.; Jow, H-N; Rollstin, J.A.; Helton, J.C.

    1990-12-01

    Estimation of offsite accident consequences is the customary final step in a probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five US power reactors (NUREG-1150). Offsite accident consequences for NUREG-1150 source terms were estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value was recommended. This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric and biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report. 134 refs., 15 figs., 110 tabs

  10. Study on the code system for the off-site consequences assessment of severe nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

  11. Study on the code system for the off-site consequences assessment of severe nuclear accident

    International Nuclear Information System (INIS)

    Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk

    2016-01-01

    The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents

  12. Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

    1983-02-01

    The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

  13. Experience with COSYMA in an international intercomparison of probabilistic accident consequence assessment codes

    International Nuclear Information System (INIS)

    Hasemann, I.; Jones, J.A.; Steen, J. van der; Wonderen, E. van

    1996-01-01

    The Commission of the European Communities and the Nuclear Energy Agency of the OECD have organized an international exercise to compare the predictions of accident consequence assessment codes, and to identify those features of the models which lead to differences in the predicted results. Alongside this, a further exercise was undertaken in which the COSYMA code was used independently by several different organizations. Some of the findings of the COSYMA users' exercise are described that have general applications to accident consequence assessments. A number of areas are identified in which further work on accident consequence models may be justified. These areas, which are also of interest for codes other than COSYMA, are (a) the calculation and averaging of doses and risks to people sheltered in different types of buildings, particularly with respect to the evaluation of early health effects; (b) the modeling of long-duration releases and their description as a series of shorter releases; (c) meteorological sampling for results at a certain location, specifically for use with trajectory models of atmospheric dispersion; and (d) aspects of calculating probabilities of consequences at a point

  14. Methods and codes for assessing the off-site Consequences of nuclear accidents. Volume 2

    International Nuclear Information System (INIS)

    Kelly, G.N.; Luykx, F.

    1991-01-01

    The Commission of the European Communities, within the framework of its 1980-84 radiation protection research programme, initiated a two-year project in 1983 entitled methods for assessing the radiological impact of accidents (Maria). This project was continued in a substantially enlarged form within the 1985-89 research programme. The main objectives of the project were, firstly, to develop a new probabilistic accident consequence code that was modular, incorporated the best features of those codes already in use, could be readily modified to take account of new data and model developments and would be broadly applicable within the EC; secondly, to acquire a better understanding of the limitations of current models and to develop more rigorous approaches where necessary; and, thirdly, to quantify the uncertainties associated with the model predictions. This research led to the development of the accident consequence code Cosyma (COde System from MAria), which will be made generally available later in 1990. The numerous and diverse studies that have been undertaken in support of this development are summarized in this paper, together with indications of where further effort might be most profitably directed. Consideration is also given to related research directed towards the development of real-time decision support systems for use in off-site emergency management

  15. Quality assurance and verification of the MACCS [MELCOR Accident Consequence Code System] code, Version 1.5

    International Nuclear Information System (INIS)

    Dobbe, C.A.; Carlson, E.R.; Marshall, N.H.; Marwil, E.S.; Tolli, J.E.

    1990-02-01

    An independent quality assurance (QA) and verification of Version 1.5 of the MELCOR Accident Consequence Code System (MACCS) was performed. The QA and verification involved examination of the code and associated documentation for consistent and correct implementation of the models in an error-free FORTRAN computer code. The QA and verification was not intended to determine either the adequacy or appropriateness of the models that are used MACCS 1.5. The reviews uncovered errors which were fixed by the SNL MACCS code development staff prior to the release of MACCS 1.5. Some difficulties related to documentation improvement and code restructuring are also presented. The QA and verification process concluded that Version 1.5 of the MACCS code, within the scope and limitations process concluded that Version 1.5 of the MACCS code, within the scope and limitations of the models implemented in the code is essentially error free and ready for widespread use. 15 refs., 11 tabs

  16. A parametric study of MELCOR Accident Consequence Code System 2 (MACCS2) Input Values for the Predicted Health Effect

    International Nuclear Information System (INIS)

    Kim, So Ra; Min, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk

    2016-01-01

    The MELCOR Accident Consequence Code System 2, MACCS2, has been the most widely used through the world among the off-site consequence analysis codes. MACCS2 code is used to estimate the radionuclide concentrations, radiological doses, health effects, and economic consequences that could result from the hypothetical nuclear accidents. Most of the MACCS model parameter values are defined by the user and those input parameters can make a significant impact on the output. A limited parametric study was performed to identify the relative importance of the values of each input parameters in determining the predicted early and latent health effects in MACCS2. These results would not be applicable to every case of the nuclear accidents, because only the limited calculation was performed with Kori-specific data. The endpoints of the assessment were early- and latent cancer-risk in the exposed population, therefore it might produce the different results with the parametric studies for other endpoints, such as contamination level, absorbed dose, and economic cost. Accident consequence assessment is important for decision making to minimize the health effect from radiation exposure, accordingly the sufficient parametric studies are required for the various endpoints and input parameters in further research

  17. Proceedings of the Seminar on Methods and Codes for Assessing the off-site consequences of nuclear accidents. Volume 1

    International Nuclear Information System (INIS)

    Kelly, G.N.; Luykx, F.

    1991-01-01

    The Commission of the European Communities, within the framework of its 1980-84 radiation protection research programme, initiated a two-year project in 1983 entitled 'methods for assessing the radiological impact of accidents' (Maria). This project was continued in a substantially enlarged form within the 1985-89 research programme. The main objectives of the project were, firstly, to develop a new probabilistic accident consequence code that was modular, incorporated the best features of those codes already in use, could be readily modified to take account of new data and model developments and would be broadly applicable within the EC; secondly, to acquire a better understanding of the limitations of current models and to develop more rigorous approaches where necessary; and, thirdly, to quantify the uncertainties associated with the model predictions. This research led to the development of the accident consequence code Cosyma (COde System from MAria), which will be made generally available later in 1990. The numerous and diverse studies that have been undertaken in support of this development are summarized in this paper, together with indications of where further effort might be most profitably directed. Consideration is also given to related research directed towards the development of real-time decision support systems for use in off-site emergency management

  18. Parameterization of the driving time in the evacuation or fast relocation model of an accident consequence code

    International Nuclear Information System (INIS)

    Pfeffer, W.; Hofer, E.; Nowak, E.; Schnadt, H.

    1988-01-01

    The model of protective measures in the accident consequence code system UFOMOD of the German Risk Study, Phase B, requires the driving times of the population to be evacuated for the evaluation of the dose received during the evacuation. The parameter values are derived from evacuation simulations carried out with the code EVAS for 36 sectors from various sites. The simulations indicated that the driving time strongly depends on the population density, whereas other influences are less important. It was decided to use different driving times in the consequence code for each of four population density classes as well as for each of three or four fractions of the population in a sector. The variability between sectors of a class was estimated from the 36 sectors, in order to derive subjective probability distributions that are to model the uncertainty in the parameter value to be used for any of the fractions in a particular sector for which an EVAS simulation has not yet been performed. To this end also the impact of the uncertainties in the parameters and modelling assumptions of EVAS on the simulated times was quantified using expert judgement. The distributions permit the derivation of a set of driving times to be used as so-called ''best estimate'' or reference values in the accident consequence code. Additionally they are directly applicable in an uncertainty and sensitivity analysis

  19. 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.)

  20. A review of the Melcor Accident Consequence Code System (MACCS): Capabilities and applications

    International Nuclear Information System (INIS)

    Young, M.

    1995-01-01

    MACCS was developed at Sandia National Laboratories (SNL) under U.S. Nuclear Regulatory Commission (NRC) sponsorship to estimate the offsite consequences of potential severe accidents at nuclear power plants (NPPs). MACCS was publicly released in 1990. MACCS was developed to support the NRC's probabilistic safety assessment (PSA) efforts. PSA techniques can provide a measure of the risk of reactor operation. PSAs are generally divided into three levels. Level one efforts identify potential plant damage states that lead to core damage and the associated probabilities, level two models damage progression and containment strength for establishing fission-product release categories, and level three efforts evaluate potential off-site consequences of radiological releases and the probabilities associated with the consequences. MACCS was designed as a tool for level three PSA analysis. MACCS performs probabilistic health and economic consequence assessments of hypothetical accidental releases of radioactive material from NPPs. MACCS includes models for atmospheric dispersion and transport, wet and dry deposition, the probabilistic treatment of meteorology, environmental transfer, countermeasure strategies, dosimetry, health effects, and economic impacts. The computer systems MACCS is designed to run on are the 386/486 PC, VAX/VMS, E3M RISC S/6000, Sun SPARC, and Cray UNICOS. This paper provides an overview of MACCS, reviews some of the applications of MACCS, international collaborations which have involved MACCS, current developmental efforts, and future directions

  1. Comparison of MACCS users calculations for the international comparison exercise on probabilistic accident consequence assessment code, October 1989--June 1993

    International Nuclear Information System (INIS)

    Neymotin, L.

    1994-04-01

    Over the past several years, the OECD/NEA and CEC sponsored an international program intercomparing a group of six probabilistic consequence assessment (PCA) codes designed to simulate health and economic consequences of radioactive releases into atmosphere of radioactive materials following severe accidents at nuclear power plants (NPPs): ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this effort, two separate groups performed similar calculations using the MACCS and COSYMA codes. Results produced in the MACCS Users Group (Greece, Italy, Spain, and USA) calculations and their comparison are contained in the present report. Version 1.5.11.1 of the MACCS code was used for the calculations. Good agreement between the results produced in the four participating calculations has been reached, with the exception of the results related to the ingestion pathway dose predictions. The main reason for the scatter in those particular results is attributed to the lack of a straightforward implementation of the specifications for agricultural production and counter-measures criteria provided for the exercise. A significantly smaller scatter in predictions of other consequences was successfully explained by differences in meteorological files and weather sampling, grids, rain distance intervals, dispersion model options, and population distributions

  2. Evaluation of dose calculation models for inhabited areas applicable in nuclear accident consequence assessment codes

    International Nuclear Information System (INIS)

    Katalin Eged; Zoltan Kis; Natalia Semioschkina; Gabriele Voigt

    2004-01-01

    One of the objectives of the EC project EVANET-TERRA is to provide suitable inputs to the RODOS system. This study gives an overview on urban dose calculation models with special emphasis on the RECLAIM-EDEM2M and TEMAS-urban codes. The TEMAS-urban code is more complex compared to the RECLAIM-EDEM2M code although both models use similar and some times even same model parameters. The database and the way of its data collection as used in RECLAIM-EDEM2M is recommended as a preferred option because it contains many data from local and regional measurements. However in a decision situation the outputs of the TEMASurban model may better help stake holders by providing a ranking of the surfaces to be decontaminated. (author)

  3. Assessment of Radiological and Economic Consequences of a Hypothetical Accident for ETRR-2, Egypt Utilizing COSYMA Code

    International Nuclear Information System (INIS)

    Tawfik, F.S.; Abdel-Aal, M.M.

    2008-01-01

    A comprehensive probabilistic study of an accident consequence assessment (ACA) for loss of coolant accident (LOCA) has accomplished to the second research reactor ETRR-2, located at Inshas Nuclear Research Center, Cairo, Egypt. PC-COSYMA, developed with the support of European Commission, has adopted to assess the radiological and economic consequences of a proposed accident. The consequences of the accident evaluated in case of early and late effects. The effective doses and doses in different organs carried out with and without countermeasures. The force mentioned calculations were required the following studies: the core inventory due to the hypothetical accident, the physical parameters of the source term, the hourly basis meteorological parameters for one complete year, and the population distribution around the plant. The hourly stability conditions and height of atmospheric boundary layers (ABL) of the concerned site were calculated. The results showed that, the nuclides that have short half-lives (few days) give the highest air and ground concentrations after the accident than the others. The area around the reactor requires the early and late countermeasures action after the accident especially in the downwind sectors. Economically, the costs of emergency plan are effectively high in case of applying countermeasures but countermeasures reduce the risk effects

  4. Assessing economic consequences of radiation accidents

    International Nuclear Information System (INIS)

    Rowe, M.D.; Lee, J.C.; Grimshaw, C.A.; Kalb, P.D.

    1987-01-01

    A recent review of existing models and methods for assessing potential consequences of accidents in the high-level radioactive waste (HLW) disposal system identifies economic consequence assessment methods as a weak point. Existing methods have mostly been designed to assess economic consequences of reactor accidents, the possible scale of which can be several orders of magnitude greater than anything possible in the HLW disposal system. There is therefore some question about the applicability of these methods, their assumptions, and their level of detail to assessments of smaller accidents. The US Dept. of Energy funded this study to determine needs for code modifications or model development for assessing economic costs of accidents in the HLW disposal system. The objectives of the study were as follows: (1) review the literature on economic consequences of accidents to determine the availability of assessment methods and data and their applicability to the HLW disposal system before closure. (2) Determine needs for expansion, revision, or adaptation of methods and data for modeling economic consequences of accidents of the scale projected for the disposal system. (3) Gather data that might be useful for the needed revisions for modeling economic impacts on this scale

  5. Modeling atmospheric dispersion for reactor accident consequence evaluation

    International Nuclear Information System (INIS)

    Alpert, D.J.; Gudiksen, P.H.; Woodard, K.

    1982-01-01

    Atmospheric dispersion models are a central part of computer codes for the evaluation of potential reactor accident consequences. A variety of ways of treating to varying degrees the many physical processes that can have an impact on the predicted consequences exists. The currently available models are reviewed and their capabilities and limitations, as applied to reactor accident consequence analyses, are discussed

  6. Review of the chronic exposure pathways models in MACCS [MELCOR Accident Consequence Code System] and several other well-known probabilistic risk assessment models

    International Nuclear Information System (INIS)

    Tveten, U.

    1990-06-01

    The purpose of this report is to document the results of the work performed by the author in connection with the following task, performed for US Nuclear Regulatory Commission, (USNRC) Office of Nuclear Regulatory Research, Division of Systems Research: MACCS Chronic Exposure Pathway Models: Review the chronic exposure pathway models implemented in the MELCOR Accident Consequence Code System (MACCS) and compare those models to the chronic exposure pathway models implemented in similar codes developed in countries that are members of the OECD. The chronic exposures concerned are via: the terrestrial food pathways, the water pathways, the long-term groundshine pathway, and the inhalation of resuspended radionuclides pathway. The USNRC has indicated during discussions of the task that the major effort should be spent on the terrestrial food pathways. There is one chapter for each of the categories of chronic exposure pathways listed above

  7. Development of an accident consequence assessment code for evaluating site suitability of light- and heavy-water reactors based on the Korean Technical standards

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Won Tae; Jeong, Hae Sung; Jeong, Hyo Joon; Kil, A Reum; Kim, Eun Han; Han, Moon Hee [Nuclear Environment Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC's positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

  8. The consequences of Chernobyl accident

    Directory of Open Access Journals (Sweden)

    Ion Chioșilă

    2016-12-01

    Full Text Available These days marks 30 years since the Chernobyl nuclear accident, followed by massive radioactive contamination of the environment and human in Belarus, Ukraine and Russia, and resulted in many deaths among people who intervened to decrease the effects of the nuclear disaster. The 26 April 1986 nuclear accident contaminated all European countries, but at a much lower level, without highlighted consequences on human health. In special laboratories, the main radionuclides (I-131, Cs-137, Cs-134 and Sr-90 were also analyzed in Romania from environmental samples, food, even human subjects. These radionuclides caused the population to receive a low dose of about 1 mSv in 1986 that is half of the dose of the natural background radiation (2.4 mSv per year. As in all European countries (excluding Ukraine, Belarus and Russia this dose of about 1 mSv fell rapidly by 1990, reaching levels close to ones before the accident at the nuclear tests.

  9. Cost per severe accident as an index for severe accident consequence assessment and its applications

    International Nuclear Information System (INIS)

    Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

    2014-01-01

    The Fukushima Accident emphasizes the need to integrate the assessments of health effects, economic impacts, social impacts and environmental impacts, in order to perform a comprehensive consequence assessment of severe accidents in nuclear power plants. “Cost per severe accident” is introduced as an index for that purpose. The calculation methodology, including the consequence analysis using level 3 probabilistic risk assessment code OSCAAR and the calculation method of the cost per severe accident, is proposed. This methodology was applied to a virtual 1,100 MWe boiling water reactor. The breakdown of the cost per severe accident was provided. The radiation effect cost, the relocation cost and the decontamination cost were the three largest components. Sensitivity analyses were carried out, and parameters sensitive to cost per severe accident were specified. The cost per severe accident was compared with the amount of source terms, to demonstrate the performance of the cost per severe accident as an index to evaluate severe accident consequences. The ways to use the cost per severe accident for optimization of radiation protection countermeasures and for estimation of the effects of accident management strategies are discussed as its applications. - Highlights: • Cost per severe accident is used for severe accident consequence assessment. • Assessments of health, economic, social and environmental impacts are included. • Radiation effect, relocation and decontamination costs are important cost components. • Cost per severe accident can be used to optimize radiation protection measures. • Effects of accident management can be estimated using the cost per severe accident

  10. CINETHICA - Core accident analysis code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

    A computer program for nuclear accident analysis has been developed based on the point-kinetics approximation and one-dimensional heat transfer model for reactivity feedback calculation. Hansen's method/1/ were used for the kinetics equation solution and explicit Euler method were adopted for the thermohidraulic equations. The results were favorably compared to those from the GAPOTKIN Code/2/. (author) [pt

  11. Socioeconomic consequences of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Tawil, J.J.; Callaway, J.W.; Coles, B.L.; Cronin, F.J.; Currie, J.W.; Imhoff, K.L.; Lewis, P.M.; Nesse, R.J.; Strenge, D.L.

    1984-06-01

    This report identifies and characterizes the off-site socioeconomic consequences that would likely result from a severe radiological accident at a nuclear power plant. The types of impacts that are addressed include economic impacts, health impacts, social/psychological impacts and institutional impacts. These impacts are identified for each of several phases of a reactor accident - from the warning phase through the post-resettlement phase. The relative importance of the impact during each accident phase and the degree to which the impact can be predicted are indicated. The report also examines the methods that are currently used for assessing nuclear reactor accidents, including development of accident scenarios and the estimating of socioeconomic accident consequences with various models. Finally, a critical evaluation is made regarding the use of impact analyses in estimating the contribution of socioeconomic consequences to nuclear accident reactor accident risk. 116 references, 7 figures, 15 tables

  12. DOZIM - evaluation dose code for nuclear accident

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  13. Estimation of cost per severe accident for improvement of accident protection and consequence mitigation strategies

    International Nuclear Information System (INIS)

    Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

    2013-01-01

    To assess the complex situations regarding the severe accidents such as what observed in Fukushima Accident, not only radiation protection aspects but also relevant aspects: health, environmental, economic and societal aspects; must be all included into the consequence assessment. In this study, the authors introduce the “cost per severe accident” as an index to analyze the consequences of severe accidents comprehensively. The cost per severe accident consists of various costs and consequences converted into monetary values. For the purpose of improvement of the accident protection and consequence mitigation strategies, the costs needed to introduce the protective actions, and health and psychological consequences are included in the present study. The evaluations of these costs and consequences were made based on the systematic consequence analysis using level 2 and 3 probabilistic safety assessment (PSA) codes. The accident sequences used in this analysis were taken from the results of level 2 seismic PSA of a virtual 1,100 MWe BWR-5. The doses to the public and the number of people affected were calculated using the level 3 PSA code OSCAAR of Japan Atomic Energy Agency (JAEA). The calculations have been made for 248 meteorological sequences, and the outputs are given as expectation values for various meteorological conditions. Using these outputs, the cost per severe accident is calculated based on the open documents on the Fukushima Accident regarding the cost of protective actions and compensations for psychological harms. Finally, optimized accident protection and consequence mitigation strategies are recommended taking into account the various aspects comprehensively using the cost per severe accident. The authors must emphasize that the aim is not to estimate the accident cost itself but to extend the scope of “risk-informed decision making” for continuous safety improvements of nuclear energy. (author)

  14. The Chernobyl accidents: Causes and Consequences

    International Nuclear Information System (INIS)

    Chihab-Eddine, A.

    1988-01-01

    The objective of this communication is to discuss the causes and the consequences of the Chernobyl accident. To facilitate the understanding of the events that led to the accident, the author gave a simplified introduction to the important physics that goes on in a nuclear reactor and he presented a brief description and features of chernobyl reactor. The accident scenario and consequences have been presented. The common contribution factors that led to both Three Mile Island and Chernobyl accidents have been pointed out.(author)

  15. Character and consequence of nuclear criticality accident

    International Nuclear Information System (INIS)

    Liu Xinhua; Liu Hua; Wu Deqiang; Li Bing

    2001-01-01

    The author describes some concepts, the process and magnitude of energy release and the destruction of the nuclear criticality accident and also describes the radiation consequence of criticality accidents from three aspects: prompt radiation, contamination in working place and release of fission products to the environment. It shows that the effects of radioactivity release from criticality accidents in the nuclear fuel processing plants on the environment and the public is minor, the main danger is from the external exposure of prompt rays. The paper make as have a correct understanding of the nuclear criticality accident and it would be helpful to take appropriate emergency response to potential criticality accident

  16. Processing Expert Judgements in Accident Consequence Modelling (invited paper)

    International Nuclear Information System (INIS)

    Kraan, B.C.P.; Cooke, R.M.

    2000-01-01

    In performing uncertainty analysis a distribution on the code input parameters is required. The construction of the distribution on the code input parameters for the joint CEC/USNRC Accident Consequence Code Uncertainty Analysis using Expert Judgement is discussed. An example from the food chain module is used to illustrate the construction. Different mathematical techniques have been developed to transform the expert judgements into the required format. Finally, the effect of taking account of correlations in performing uncertainty analysis is investigated. (author)

  17. Evaluation of nuclear accidents consequences. Risk assessment methodologies, current status and applications

    International Nuclear Information System (INIS)

    Rodriguez, J.M.

    1996-01-01

    General description of the structure and process of the probabilistic methods of assessment the external consequences in the event of nuclear accidents is presented. attention is paid in the interface with Probabilistic Safety Analysis level 3 results (source term evaluation) Also are described key issues in accident consequence evaluation as: effects evaluated (early and late health effects and economic effects due to countermeasures), presentation of accident consequences results, computer codes. Briefly are presented some relevant areas for the applications of Accident Consequence Evaluation

  18. Consequences of the Chernobyl accident

    International Nuclear Information System (INIS)

    Gerzabek, M.H.

    1990-10-01

    A collection of three papers about the fallout in Austria from the 1986 Chernobyl reactor accident is given: 1. An overview of the research projects in Austria; 2. On the transfer into and uptake by crops and animal fodder; 3. On the reduction of cesium concentration in food. 18 tabs., 21 figs., 69 refs

  19. Assessing economic consequences of radiation accidents

    International Nuclear Information System (INIS)

    Rowe, M.D.; Lee, J.C.; Grimshaw, C.A.; Kalb, P.D.

    1987-01-01

    This project reviewed the literature on the economic consequences of accidents to determine the availability of assessment methods and data and their applicability to the high-level radioactive waste (HLW) disposal system before closure; determined needs for expansion, revision, or adaptation of methods and data for modeling economic consequences of accidents of the scale projected for the disposal system; and gathered data that might be useful for the needed revisions. 8 refs., 1 tab

  20. Chernobylsk accident (Causes and Consequences)- Part 2

    International Nuclear Information System (INIS)

    Esteves, D.

    1986-09-01

    The causes and consequences of the nuclear accident at Chernobylsk-4 reactor are shortly described. The informations were provided by Russian during the specialist meeting, carried out at seat of IAEA. The Russian nuclear panorama; the site, nuclear power plant characteristics and sequence of events; the immediate measurements after accident; monitoring/radioactive releases; environmental contamination and ecological consequences; measurements of emergency; recommendations to increase the nuclear safety; and recommendations of work groups, are presented. (M.C.K.) [pt

  1. Medical consequences of Chernobyl accident

    Directory of Open Access Journals (Sweden)

    Galstyan I.A.

    2015-12-01

    Full Text Available Aim: to study the long-term effects of acute radiation syndrome (ARS, developed at the victims of the Chernobyl accident. Material and Methods. 237 people were exposed during the accident, 134 of them were diagnosed with ARS. Dynamic observation implies a thorough annual examination in a hospital. Results. In the first 1.5-2 years after the ARS mean group indices of peripheral blood have returned to normal. However, many patients had transient expressed moderate cytopenias. Granulocytopenia, thrombocytopenia, lymphopenia and erythropenia were the most frequently observed things during the first 5 years after the accident. After 5 years their occurences lowered. In 11 patients the radiation cataract was detected. A threshold dose for its development is a dose of 3.2 Gy Long-term effects of local radiation lesions (LRL range from mild skin figure smoothing to a distinct fibrous scarring, contractures, persistently recurrent late radiation ulcers. During all years of observation we found 8 solid tumors, including 2 thyroid cancers. 5 hematologic diseases were found. During 29 years 26 ARS survivors died of various causes. Conclusion. The health of ones with long-term ARS effects is determined by the evolution of the LRL effects on skin, radiation cataracts, hema-tological diseases and the accession of of various somatic diseases, not caused by radiation.

  2. The nuclear accidents: Causes and consequences

    International Nuclear Information System (INIS)

    Rochd, M.

    1988-01-01

    The author discussed and compared the real causes of T.M.I. and Chernobyl accidents and cited their consequences. To better understand how these accidents occurred, a brief description of PWR type (reactor type of T.M.I.) and of RBMK type (reactor type of Chernobyl) has been presented. The author has also set out briefly the safety analysis objectives and the three barriers established to protect the public against the radiological consequences. To distinguish failures that cause severe accidents and to analyze them in details, it is necessary to classify the accidents. There are many ways to do it according to their initiator event, or to their frequency, or to their degree of gravity. The safety criteria adopted by nuclear industry have been explained. These criteria specify the limits of certain physical parameters that should not be exceeded in case of incidents or accidents. To compare the real causes of T.M.I. and Chernobyl accidents, the events that led to both have been presented. As observed the main common contributing factors in both cases are that the operators did not pay attention to warnings and signals that were available to them and that they were not trained to handle these accident sequences. The essential conclusions derived from these severe accidents are: -The improvement of operators competence contribute to reduce the accident risks; -The rapid and correct diagnosis of real conditions at each point of the accidents permits an appropriate behavior that would bring the plant to a stable state; -Competent technical teams have to intervene and to assist the operators in case of emergency; -Emergency plans and an international collaboration are necessary to limit the accident risks. 11 figs. (author)

  3. The Chernobyl nuclear accident and its consequences

    International Nuclear Information System (INIS)

    1986-01-01

    An AAEC Task Group was set up shortly after the accident at the Chernobyl Nuclear Power Plant to monitor and evaluate initial reports and to assess the implications for Australia. The Task Group issued a preliminary report on 9 May 1986. On 25-29 August 1986, the USSR released details of the accident and its consequences and further information has become available from the Nuclear Energy Agency of OECD and the World Health Organisation. The Task Group now presents a revised report summarising this information and commenting on the consequences from the Australian viewpoint

  4. Development of a Methodology for VHTR Accident Consequence Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

    2016-05-15

    The substitution of the VHTR for burning fossil fuels conserves these hydrocarbon resources for other uses and eliminates the emissions of greenhouse. In Korea, for these reasons, constructing the VHTR plan for hydrogen production is in progress. In this study, the consequence analysis for the off-site releases of radioactive materials during severe accidents has been performed using the level 3 PRA technology. The offsite consequence analysis for a VHTR using the MACCS code has been performed. Since the passive system such as the RCCS(Reactor Cavity Cooling System) are equipped, the frequency of occurrence of accidents has been evaluated to be very low. For further study, the assessment for characteristic of VHTR safety system and precise quantification of its accident scenarios is expected to conduct more certain consequence analysis. This methodology shown in this study might contribute to enhancing the safety of VHTR design by utilizing the results having far lower effect on the environment than the LWRs.

  5. Health consequences of nuclear accidents

    International Nuclear Information System (INIS)

    Masse, Roland

    2011-10-01

    The author first outlines that no exposure of mankind to environmental risks has been as exhaustively and continuously studied as that resulting from ionizing radiations. Apart from lethal effects, he describes non lethal cell lesions which are induced in tissues: mutations and modifications of gene expressions, either directly under the effect of radiation, or by water hydrolysis, or indirectly through a biochemical response to these initial events. Then, the author evokes the controversy about Chernobyl: according to scientists, there is no relationship between the health degradation (human morbidity and mortality) and fallouts whereas activist groups state that there is. The author then evokes that the WHO and the IAEA were accused to lie about the issue of victims and health consequences. He outlines that UNSCEAR reports are a reference for radio-biologists, and that the 2011 report confirmed the conclusions of the 2006 report. He comments some published data, notably those on the acute irradiation syndrome (ARS), on carcinogenic effects (essentially thyroid cancers for children, as there is no clear nor admitted relationship for other forms of cancer), on other pathologies. Finally, the author briefly discusses the issue of crisis management, the information about Fukushima, and the issue of Chernobyl fallouts in France

  6. Biological and medical consequences of nuclear accidents

    International Nuclear Information System (INIS)

    Latarjet, R.

    1988-01-01

    The study of the medical and biological consequences of the nuclear accidents is a vast program. The Chernobyl accident has caused some thirty deceases: Some of them were rapid and the others occurred after a certain time. The particularity of these deaths was that the irradiation has been associated to burns and traumatisms. The lesson learnt from the Chernobyl accident is to treat the burn and the traumatism before treating the irradiation. Contrary to what the research workers believe, the first wave of deaths has passed between 15 and 35 days and it has not been followed by any others. But the therapeutic lesson drawn from the accident confirm the research workers results; for example: the radioactive doses band that determines where the therapy could be efficacious or not. the medical cares dispensed to the irradiated people in the hospital of Moscow has confirmed that the biochemical equilibrium of proteinic elements of blood has to be maintained, and the transfusion of the purified elements are very important to restore a patient to health, and the sterilization of the medium (room, food, bedding,etc...) of the patient is indispensable. Therefore, it is necessary to establish an international cooperation for providing enough sterilized rooms and specialists in the irradiation treatment. The genetic consequences and cancers from the Chernobyl accident have been discussed. It is impossible to detect these consequences because of their negligible percentages. (author)

  7. Radiological consequences of the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Hill, P.; Hille, R.

    2003-01-01

    The reactor accident at unit 4 of the Chernobyl nuclear power plant in Ukraine has deeply affected the living conditions of millions of people. Especially the health consequences have been of public concern up to the present and also been the subject of sometimes absurd claims. The current knowledge on the radiological consequences of the accident is reviewed. Though an increased hazard for some risk groups with high radiation exposure, e.g., liquidators, still cannot be totally excluded for the future, the majority of the population shows no statistically significant indication of radiation-induced illnesses. The contribution of the Research Center Juelich to the assessment of the post-accidental situation and psychological relief of the population is reported. The population groups still requiring special attention include, in particular, children growing up in highly contaminated regions and the liquidators of the years 1986 and 1987 deployed immediately after the accident. (author)

  8. Consequences in Sweden of the Chernobyl accident

    International Nuclear Information System (INIS)

    Snihs, J.O.

    1986-01-01

    It summarizes the consequences in Sweden of the Chernobyl accident, describes the emergency response, the basis for decisions and countermeasures, the measurement strategies, the activity levels and doses and countermeasures and action levels used. Past and remaining problems are discussed and the major investigations and improvements are given. (author)

  9. Consequences in Guatemala of the Chernobyl accident

    International Nuclear Information System (INIS)

    Perez Sabino, J.F.; Ayala Jimenez, R.E.

    1997-01-01

    Because of the long distance between Guatemala and Chernobyl, the country did not undergo direct consequences of radioactive contamination in the short term. However, the accident repercussions were evident in the medium and long-term, mainly in two sectors, the economic-political and the environmental sectors

  10. Real and mythical consequences of Chernobyl accident

    International Nuclear Information System (INIS)

    Osmachkin, V.S.

    1999-01-01

    This presentation describes the public Unacceptance of Nuclear Power as a consequence of Chernobyl Accident, an accident which was a severest event in the history of the nuclear industry. It was a shock for everybody, who has been involved in nuclear power programs. But nobody could expect that it was also the end romantic page in the nuclear story. The scale of the detriment was a great, and it could be compared with other big technological man-made catastrophes. But immediately after an accident mass media and news agencies started to transmit an information with a great exaggerations of the consequences of the event. In a report on the Seminar T he lessons of the Chernobyl - 1' in 1996 examples of such incorrect information, were cited. Particularly, in the mass media it was declared that consequences of the accident could be compared with a results of the second world war, the number of victims were more than hundred thousand people, more than million of children have the serious health detriments. Such and other cases of the misconstruction have been called as myths. The real consequences of Chernobyl disaster have been summed on the International Conference 'One decade after Chernobyl' - 2, in April 1996. A very important result of the Chernobyl accident was a dissemination of stable unacceptance of the everything connected with 'the atom'. A mystic horror from invisible mortal radiation has been inspired in the masses. And from such public attitude the Nuclear Power Programs in many countries have changed dramatically. A new more pragmatic and more careful atomic era started with a slogan: 'Kernkraftwerk ? Nein, danke'. No doubt, a Chernobyl accident was a serious technical catastrophe in atomic industry. The scale of detriment is connected with a number of involved peoples, not with a number of real victims. In comparison with Bhopal case, earthquakes, crashes of the airplanes, floods, traffic accidents and other risky events of our life - the Chernobyl is

  11. Health consequences [of the Chernobyl accident

    International Nuclear Information System (INIS)

    Ramoutar, S.

    1996-01-01

    The World Health Organisation Conference on the Health Consequences of the Chernobyl and Other Radiological Accidents, held in Geneva last November, is reported. The lack of representation from the civil nuclear industry led often to one-sided debates instigated by the anti-nuclear lobbies present. Thyroid cancer in children as a result of the Chernobyl accident received particular attention. In Belarus, 400 cases have been noted, 220 in Ukraine and 60 in the Russian Federation. All have been treated with a high degree of success. The incidence of this cancer would be expected to follow the fallout path as the main exposure route was ingestion of contaminated foods and milk products. It was noted that the only way to confirm causality was if those children born since the accident failed to show the same increased incidence. Explanations were offered for the particular susceptibility of children to thyroid cancer following exposure to radiation. Another significant cause of concern was the health consequences to clean-up workers in radiological accidents. The main factor is psychological problems from the stress of knowing that they have received high radiation doses. A dramatic increase in psychological disorders has occurred in the Ukraine over the past ten years and this is attributed to stress generated by the Chernobyl accident, compounded by the inadequacy of the public advice offered at the time and the socio-economic uncertainties accompanying the breakup of the former USSR. (UK)

  12. The Chernobyl accident consequences; Consequences de l'accident de Tchernobyl

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.)

  13. Radioecological and dosimetric consequences of the Chernobyl accident in France

    International Nuclear Information System (INIS)

    Renaud, Ph.; Beaugelin, K.; Maubert, H.; Ledenvic, Ph.

    1997-11-01

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

  14. The Chernobyl accident and its consequences.

    Science.gov (United States)

    Saenko, V; Ivanov, V; Tsyb, A; Bogdanova, T; Tronko, M; Demidchik, Yu; Yamashita, S

    2011-05-01

    The accident at the Chernobyl nuclear power plant was the worst industrial accident of the last century that involved radiation. The unprecedented release of multiple different radioisotopes led to radioactive contamination of large areas surrounding the accident site. The exposure of the residents of these areas was varied and therefore the consequences for health and radioecology could not be reliably estimated quickly. Even though some studies have now been ongoing for 25 years and have provided a better understanding of the situation, these are yet neither complete nor comprehensive enough to determine the long-term risk. A true assessment can only be provided after following the observed population for their natural lifespan. Here we review the technical aspects of the accident and provide relevant information on radioactive releases that resulted in exposure of this large population to radiation. A number of different groups of people were exposed to radiation: workers involved in the initial clean-up response, and members of the general population who were either evacuated from the settlements in the Chernobyl nuclear power plant vicinity shortly after the accident, or continued to live in the affected territories of Belarus, Russia and Ukraine. Through domestic efforts and extensive international co-operation, essential information on radiation dose and health status for this population has been collected. This has permitted the identification of high-risk groups and the use of more specialised means of collecting information, diagnosis, treatment and follow-up. Because radiation-associated thyroid cancer is one of the major health consequences of the Chernobyl accident, a particular emphasis is placed on this malignancy. The initial epidemiological studies are reviewed, as are the most significant studies and/or aid programmes in the three affected countries. Copyright © 2011 The Royal College of Radiologists. Published by Elsevier Ltd. All rights

  15. Environmental consequences of releases from nuclear accidents

    International Nuclear Information System (INIS)

    Tveten, U.

    1990-01-01

    The primary purpose of this report is to present the results of a four-year Nordic cooperation program in the area of consequence assessment of nuclear accidents with large releases to the environment. This program was completed in 1989. Related information from other research programs has also been described, so that many chapters of the report reflect the current status in the respective areas, in addition to containing the results of the Nordic program. (author) 179 refs

  16. Chernobyl radiological data for accident consequence assessment

    International Nuclear Information System (INIS)

    Bottino, A.; Sacripanti, A.

    1989-01-01

    In this draft is presented the results of a first effort to summarize information related to the radionuclides behaviour in rural areas, in order to estimate pathway parameters to assess accident consequences. This topic encloses relevant aspects concerning contamination of rural environment, the most important being: 1) dry deposition velocities; 2) washout coefficient; 3) accumulation in lakes; 4) migration in soil; 5) winter conditions; 6) filtering effects of forests

  17. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.

    1988-07-01

    This report was commissioned by the Swedish State Power Board. The report contains a calculation of radiation doses in the surroundings caused by a theoretical core meltdown accident at Forsmark reactor No 3. The accident sequence chosen for the calculating was a release caused by total power failure. The calculations were made by means of the PLUCON4 code. Meteorological data for two years from the Forsmark meteorological tower were analysed to find representative weather situations. As typical weather, Pasquill D was chosen with a wind speed of 5 m/s, and as extreme weather, Pasquill F with a wind speed of 2 m/s. 23 tabs., 37 ills., 20 refs. (author)

  18. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.; Thykier-Nielsen, S.; Walmod-Larsen, O.

    1986-08-01

    This report was commissioned by the Swedish State Power Board, who wanted a method for calculation of radiation doses in the surroundings of nuclear power plants caused by severe accidents. The PLUCON4 code were used for the calculations. A TC-SV-accident at Ringhals 1 wer chosen as example. A transient without shutdown leads to core meltdown through the reactor vessel. The pressure peak at the moment of vessel failure opens a safety valve in the dry well. Meteorolgical data for two years from the Ringhals meteorological tower were analysed to find representative weather situations. As typical weather were chosen Pasquill D with wind speed 8 m/s, and as extreme weather were chosen Pasquill F with wind speed 4.8 m/s. (author)

  19. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.

    1988-01-01

    This report was commissioned by the Swedish State Power Board. The report contains a calculation of radiation doses in the surroundings caused by a theoretical core meltdown accident at Ringhals reactor No 3/4. The accident sequence chosen for the calcualtions was a release caused by total power failure. The calculations were made by means of the PLUCON4 code. A decontamination factor of 500 is used to account for the scrubber effect. Meteorological data for two years from the Ringhals meteorological tower were analysed to find representative weather situations. As typical weather, Pasquill D, was chosen with a wind speed of 10 m/s, and as extreme weather, Pasquill E, with a wind speed of 2 m/s. 19 refs. (author)

  20. Additional investigations on the consequences of accidents

    International Nuclear Information System (INIS)

    Ehrhardt, J.; Bayer, A.; Burkart, K.

    1982-01-01

    As a first step to improve the accident consequence model of the German Risk Study within the Phase B, additional investigations on special problems and questions were performed. In detail attention is given to the following topics: emergency protective actions in the vicinity of the site; latent cancer fatalities - allocated to the population living during the nuclear accident and to persons born afterwards, within and beyond a distance of 540 km from the site, caused by radiation doses below the dose limits of the German radiation protection regulations estimated assuming a nonlinear dose response function; risk assessments of nuclear power plants with lower capacities; loss of life expectancy after accidental radiation exposure. All results are presented separately for the 8 release categories of the German Risk Study. (orig.) [de

  1. Consequences of severe nuclear accidents in Europe

    Science.gov (United States)

    Seibert, Petra; Arnold, Delia; Mraz, Gabriele; Arnold, Nikolaus; Gufler, Klaus; Kromp-Kolb, Helga; Kromp, Wolfgang; Sutter, Philipp

    2013-04-01

    A first part of the presentation is devoted to the consequences of the severe accident in the 1986 Chernobyl NPP. It lead to a substantial radioactive contaminated of large parts of Europe and thus raised the awareness for off-site nuclear accident consequences. Spatial patterns of the (transient) contamination of the air and (persistent) contamination of the ground were studied by both measurements and model simulations. For a variety of reasons, ground contamination measurements have variability at a range of spatial scales. Results will be reviewed and discussed. Model simulations, including inverse modelling, have shown that the standard source term as defined in the ATMES study (1990) needs to be updated. Sensitive measurements of airborne activities still reveal the presence of low levels of airborne radiocaesium over the northern hemisphere which stems from resuspension. Over time scales of months and years, the distribution of radionuclides in the Earth system is constantly changing, for example relocated within plants, between plants and soil, in the soil, and into water bodies. Motivated by the permanent risk of transboundary impacts from potential major nuclear accidents, the multidisciplinary project flexRISK (see http://flexRISK.boku.ac.at) has been carried out from 2009 to 2012 in Austria to quantify such risks and hazards. An overview of methods and results of flexRISK is given as a second part of the presentation. For each of the 228 NPPs, severe accidents were identified together with relevant inventories, release fractions, and release frequencies. Then, Europe-wide dispersion and dose calculations were performed for 2788 cases, using the Lagrangian particle model FLEXPART. Maps of single-case results as well as various aggregated risk parameters were produced. It was found that substantial consequences (intervention measures) are possible for distances up to 500-1000 km, and occur more frequently for a distance range up to 100-300 km, which is in

  2. Radiological consequences of the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Hill, P.; Hille, R.

    2002-01-01

    Fifty years of peaceful utilization of nuclear power were interrupted by the reactor accident in unit 4 of the Chernobyl nuclear power station in Ukraine in 1986, a disruptive event whose consequences profoundly affected the way of life of millions of people, and which has moved the public to this day. Releases of radioactive materials contaminated large areas of Belarus, the Russian Federation, and Ukraine. Early damage in the form of radiation syndrome was suffered by a group of rescue workers and members of the reactor operating crew, in some cases with fatal consequences, while the population does not, until now, show a statistically significant increase in the rate of late damage due to ionizing radiation expect for thyroid diseases in children. In particular, no increases in the rates of solid tumors, leukaemia, genetic defects, and congenital defects were detected. For some risk groups exposed to high radiation doses (such as liquidators) the hazard may still be greater, but the large majority of the population need not live in fear of serious impacts on health. Nevertheless, the accident shows major negative social and psychological consequences reinforced by the breakdown of the Soviet Union. This may be one reason for the observed higher incidence of other diseases whose association with the effects of radiation as a cause has not so far been proven. The measurement campaign conducted by the federal government in 1991-1993 addressed these very concerns of the public in an effort to provide unbiased information about exposures detected, on the one hand, in order to alleviate the fears of the public and reduce stress and, on the other hand, to contribute to the scientific evaluation of the radiological situation in the regions most highly exposed. The groups of the population requiring special attention in the future include especially children growing up in highly contaminated regions, and the liquidators of 1986 and 1987 employed in the period immediately

  3. Accident consequence assessment and siting criteria development

    International Nuclear Information System (INIS)

    Kollas, J.G.

    1988-01-01

    The methodology developed is based on assessing the average over a large spectrum of meteorological conditions whole body collective dose resulting from a severe reference accident. The assessment of this dose is performed by code CRAC.GAEC, the Greek A.E.C. version of code CRAC2. The collective dose, which is chosen as a measure of the social radiation risk, is compared to the dose corresponding to a level of social risk encountered historically in energy production as a whole. The outcome of the comparison can be the determination of one or more sectors of acceptable sites for a set of specific conditions considered, such as the reactor characteristics. The present approach was aimed to deal with the problems stemming from the demographic idiomorphy of Greece, where one third of the country's population is concentrated in Athens, with the rest of the country exhibiting small population densities. One of the applications of the methodology developed concerned the identification of acceptable sites near Athens. For these sites the risk from the reference severe accident of a standard reactor to the over three millions inhabitants of Athens is less tan the risk corresponding to the same population that is due to energy production

  4. The consequences of the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Knoechel, A.

    1988-01-01

    After the decay of the iodine isotopes the measuring campaigns, in addition to the measuring of soil pollution and pollution of products, concentrated on the way of the cesium isotopes through the food chain, especially in crops, milk, meat and mother's milk. A special programme was developed for the analysis of foreign basic substances for teas, essences and tinctures. In connection with the incorporation measurements in the university hospital Eppendorf the measurement campaigns provided the data material in order to calculate with the aid of the computer program ECOSYS of the GSF the effective dose equivalent which the inhabitants of Hamburg additionally take up due to the accident of Chernobyl. Consequences with regard to measuring methods and social consequences are mentioned. (DG) [de

  5. Probabilistic Accident Consequence Uncertainty - A Joint CEC/USNRC Study

    International Nuclear Information System (INIS)

    Gregory, Julie J.; Harper, Frederick T.

    1999-01-01

    The joint USNRC/CEC consequence uncertainty study was chartered after the development of two new probabilistic accident consequence codes, MACCS in the U.S. and COSYMA in Europe. Both the USNRC and CEC had a vested interest in expanding the knowledge base of the uncertainty associated with consequence modeling, and teamed up to co-sponsor a consequence uncertainty study. The information acquired from the study was expected to provide understanding of the strengths and weaknesses of current models as well as a basis for direction of future research. This paper looks at the elicitation process implemented in the joint study and discusses some of the uncertainty distributions provided by eight panels of experts from the U.S. and Europe that were convened to provide responses to the elicitation. The phenomenological areas addressed by the expert panels include atmospheric dispersion and deposition, deposited material and external doses, food chain, early health effects, late health effects and internal dosimetry

  6. Probabilistic Accident Consequence Uncertainty - A Joint CEC/USNRC Study

    Energy Technology Data Exchange (ETDEWEB)

    Gregory, Julie J.; Harper, Frederick T.

    1999-07-28

    The joint USNRC/CEC consequence uncertainty study was chartered after the development of two new probabilistic accident consequence codes, MACCS in the U.S. and COSYMA in Europe. Both the USNRC and CEC had a vested interest in expanding the knowledge base of the uncertainty associated with consequence modeling, and teamed up to co-sponsor a consequence uncertainty study. The information acquired from the study was expected to provide understanding of the strengths and weaknesses of current models as well as a basis for direction of future research. This paper looks at the elicitation process implemented in the joint study and discusses some of the uncertainty distributions provided by eight panels of experts from the U.S. and Europe that were convened to provide responses to the elicitation. The phenomenological areas addressed by the expert panels include atmospheric dispersion and deposition, deposited material and external doses, food chain, early health effects, late health effects and internal dosimetry.

  7. Cosyma a new programme package for accident consequence assessment

    International Nuclear Information System (INIS)

    Kelly, G.N.

    1991-01-01

    This report gives details of a new programme package for accident consequence assessment, prepared under the CEC's Maria programme (Methods for assessing the radiological impact of accidents) initiated in 1982 to review and build on the nuclear accident consequence assessment methods in use within the European Community

  8. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.; Thykier-Nielsn, S.

    1987-03-01

    This report was commissioned by the Swedish State Power Board. The report contains a calculation of radiation doses in the surroundings caused by a theoretical core meltdown accident at Forsmark reactor No 3. The assumption used for the calculations were a 0.06% release of iodine and cesium corresponding to a 0.1% release through the FILTRA plant at Barsebaeck. The calculations were made by means of the PLUCON4 code. Meteorological data for two years from the Forsmark meteorological tower were analysed to find representative weather situations. As typical weather pasquill D was chosen with wind speed 5 m/s, and as extreme weather, Pasquill F with wind speed 2 m/s. 23 tabs., 36 ills., 21 refs. (author)

  9. Method for consequence calculations for servere accidents

    International Nuclear Information System (INIS)

    Nielsen, F.

    1987-01-01

    With the exception of the part about collective doses, this report was commissioned by the Swedish State Power Board. The part about collective doses was commissioned by the Swedish National Institute of Radiation Protection. The report contains a calculation of radiation doses in the sursurroundings caused by a theoretical core meltdown accident at one of the Barsebaeck reactors with filtered venting through the FILTRA plant. The calculations were made by means of the PLUCON4 code. The assumption used for the calculations were givon by the Swedish National Institute of Radiation Protection as follows: Pasquill D with wind speed 3 m/s and a mixing layer at 300 m height. Elevation of the release: 100 m with no energy release. The release starts 12 hours after shut-down and its duration is one hour. The release contains 100% of the noble gasses and 0,1% of all other isotopes in a 1800 MW t -reactor. (author)

  10. Prevention of radiation accidents and their consequences

    International Nuclear Information System (INIS)

    Khiski, J.

    1976-01-01

    Clearing out reasons for nuclear accidents enables to take effective measures to minimize them. The number of accidents in 1957 - 1974 is given. The frequency of accidents at various working places, while operating with various radioisotopes is presented. The analysis of accidents and the confirmation of these estimates can lead to the generalization of data and to the formulation of preventive measures [ru

  11. Environmental consequences of releases from nuclear accidents

    International Nuclear Information System (INIS)

    Tveten, U.

    1990-03-01

    The report presents the results of a four-year Nordic cooperation project (AKTU-200). The results have impact upon many facets of accident consequence assessment, ranging from new computational tools to recommendations concerning food preparation methods to be utilized in a fallout situation. Some of the subprojects have approached areas where little or no research has been performed previously, like the project on winter conditions, the project on the physico/chemical form of radionuclides in the Chernobyl fallout, and the project on resuspension. The conclusion from the first of these projects is that the impact of an accident or fallout situation occuring during winter may be considerable smaller than in a similar situation during summer conditions. The most important conclusion from the second of these projects is that bioavailability of radiocesium in soil is significantly lower than that of radiocesium in plant material taken up via the roots. In the third project is was found that the resuspension factor is several orders of magnitude lower than the values traditionally cited, and that resuspension is a local phenomenon in a majority of weather conditions. The development of large-scale testing of mitigating actions to prevent uptake of radiocesium in animals in a fallout situation is also one of the projects where new ground has been sucessfully broken. 189 refs., 89 figs., 55 tabs

  12. Modeling the economic consequences of LWR accidents

    International Nuclear Information System (INIS)

    Burke, R.P.; Aldrich, D.C.; Rasmussen, N.C.

    1984-01-01

    Models to be used for analyses of economic risks from events which may occur during LWR plant operation are developed in this study. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The models can be used by both the nuclear power industry and regulatory agencies in cost-benefit analyses for decisionmaking purposes. The newly developed economic consequence models are applied in an example to estimate the economic risks from operation of the Surry Unit 2 plant. The analyses indicate that economic risks from US LWR operation, in contrast to public health risks, are dominated by relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for the Surry site. The implications of these conclusions for nuclear power plant operation and regulation are discussed

  13. A description of nuclear reactor accidents and their consequences

    International Nuclear Information System (INIS)

    Murray, A.

    1989-01-01

    Nuclear reactor accidents which have caused core damage, released a significant amount of radioactivity, or caused death or serious injury are described. The reactor accidents discussed in detail include Chernobyl, Three Mile Island, SL-1 and Windscale, although information on other less consequential accidents is also provided. The consequences of these accidents are examined in terms of the amounts of radioactivity released, the radiation doses received, and remedial actions and interventions taken following the accident. 10 refs., 1 fig., 2 tabs

  14. Chernobylsk accident (Causes and Consequences)-Part 1

    International Nuclear Information System (INIS)

    Esteves, D.

    1986-07-01

    Facts, project data, hypothesis, calculations, evaluations, monitoring, standard requirements and several considerations, related to causes, effects and consequences of Chernobylsk-4 accident. (M.C.K.) [pt

  15. 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.

  16. Transfrontier consequences to the population of Greece of large scale nuclear accidents: a preliminary assessment

    International Nuclear Information System (INIS)

    Kollas, J.G.; Catsaros, Nicolas.

    1985-06-01

    In this report the consequences to the population of Greece from hypothetical large scale nuclear accidents at the Kozlodui (Bulgaria) nuclear power station are estimated under some simplifying assumptions. Three different hypothetical accident scenarios - the most serious for pressurized water reactors - are examined. The analysis is performed by the current Greek version of code CRAC2 and includes health and economic consequences to the population of Greece. (author)

  17. causes and consequences of commercial motorcycle accidents

    African Journals Online (AJOL)

    PROF EKWUEME

    Accident associated with the use of motorcycle for commercial transportation in Makurdi metropolis was ... deaths, over speeding accounted for 27 percent of accidents and deaths respectively, .... 10. 7. 5. (a). (b). Possession of wing mirror and Crash helmet. Yes. No. 12 .... reduce the risk of serious head and brain injuries.

  18. Medical consequences of a nuclear plant accident

    International Nuclear Information System (INIS)

    Olsson, S.E.; Reizenstein, P.; Stenke, L.

    1987-01-01

    The report gives background information concerning radiation and the biological medical effects and damages caused by radiation. The report also discusses nuclear power plant accidents and efforts from the medical service in the case of a nuclear power plant accident. (L.F.)

  19. Accident consequence assessments with different atmospheric dispersion models

    International Nuclear Information System (INIS)

    Panitz, H.J.

    1989-11-01

    An essential aim of the improvements of the new program system UFOMOD for Accident Consequence Assessments (ACAs) was to substitute the straight-line Gaussian plume model conventionally used in ACA models by more realistic atmospheric dispersion models. To identify improved models which can be applied in ACA codes and to quantify the implications of different dispersion models on the results of an ACA, probabilistic comparative calculations with different atmospheric dispersion models have been performed. The study showed that there are trajectory models available which can be applied in ACAs and that they provide more realistic results of ACAs than straight-line Gaussian models. This led to a completely novel concept of atmospheric dispersion modelling in which two different distance ranges of validity are distinguished: the near range of some ten kilometres distance and the adjacent far range which are assigned to respective trajectory models. (orig.) [de

  20. Severe Accident Progression and Consequence Assessment Methodology Upgrades in ISAAC for Wolsong CANDU6

    International Nuclear Information System (INIS)

    Song, Y.M.; Kim, D.H.; Nijhawan, Sunil

    2015-01-01

    Amongst the applications of integrated severe accident analysis codes like ISAAC, the principal are to a) help develop an understanding of the severe accident progression and its consequences; b) support the design of mitigation measures by providing for them the state of the reactor following an accident; and c) to provide a training platform for accident management actions. After Fukushima accident there is an increased awareness of the need to implement effective and appropriate mitigation measures and empower the operators with training and understanding about severe accident progression and control opportunities. An updated code with reduced uncertainties can better serve these needs of the utility making decisions about mitigation measures and corrective actions. Optimal deployment of systems such as PARS and filtered containment venting require information on reactor transients for a number of critical parameters. Thus there is a greater consensus now for a demonstrated ability to perform accident progression and consequence assessment analyses with reduced uncertainties. Analyses must now provide source term transients that represent the best in available understanding and so meaningfully support mitigation measures. This requires removal of known simplifications and inclusion of all quantifiable and risk significant phenomena. Advances in understanding of CANDU6 severe accident progression reflected in the severe accident integrated code ROSHNI are being incorporated into ISAAC using CANDU specific component and system models developed and verified for Wolsong CANDU 6 reactors. A significant and comprehensive upgrade of core behavior models is being implemented in ISAAC to properly reflect the large variability amongst fuel channels in feeder geometry, fuel thermal powers and burnup. The paper summarizes the models that have been added and provides some results to illustrate code capabilities. ISAAC is being updated to meet the current requirements and

  1. Severe Accident Progression and Consequence Assessment Methodology Upgrades in ISAAC for Wolsong CANDU6

    Energy Technology Data Exchange (ETDEWEB)

    Song, Y.M.; Kim, D.H. [KAERI, Daejeon (Korea, Republic of); Nijhawan, Sunil [Prolet Inc. 98 Burbank Drive, Toronto (Canada)

    2015-05-15

    Amongst the applications of integrated severe accident analysis codes like ISAAC, the principal are to a) help develop an understanding of the severe accident progression and its consequences; b) support the design of mitigation measures by providing for them the state of the reactor following an accident; and c) to provide a training platform for accident management actions. After Fukushima accident there is an increased awareness of the need to implement effective and appropriate mitigation measures and empower the operators with training and understanding about severe accident progression and control opportunities. An updated code with reduced uncertainties can better serve these needs of the utility making decisions about mitigation measures and corrective actions. Optimal deployment of systems such as PARS and filtered containment venting require information on reactor transients for a number of critical parameters. Thus there is a greater consensus now for a demonstrated ability to perform accident progression and consequence assessment analyses with reduced uncertainties. Analyses must now provide source term transients that represent the best in available understanding and so meaningfully support mitigation measures. This requires removal of known simplifications and inclusion of all quantifiable and risk significant phenomena. Advances in understanding of CANDU6 severe accident progression reflected in the severe accident integrated code ROSHNI are being incorporated into ISAAC using CANDU specific component and system models developed and verified for Wolsong CANDU 6 reactors. A significant and comprehensive upgrade of core behavior models is being implemented in ISAAC to properly reflect the large variability amongst fuel channels in feeder geometry, fuel thermal powers and burnup. The paper summarizes the models that have been added and provides some results to illustrate code capabilities. ISAAC is being updated to meet the current requirements and

  2. Fire-accident analysis code (FIRAC) verification

    International Nuclear Information System (INIS)

    Nichols, B.D.; Gregory, W.S.; Fenton, D.L.; Smith, P.R.

    1986-01-01

    The FIRAC computer code predicts fire-induced transients in nuclear fuel cycle facility ventilation systems. FIRAC calculates simultaneously the gas-dynamic, material transport, and heat transport transients that occur in any arbitrarily connected network system subjected to a fire. The network system may include ventilation components such as filters, dampers, ducts, and blowers. These components are connected to rooms and corridors to complete the network for moving air through the facility. An experimental ventilation system has been constructed to verify FIRAC and other accident analysis codes. The design emphasizes network system characteristics and includes multiple chambers, ducts, blowers, dampers, and filters. A larger industrial heater and a commercial dust feeder are used to inject thermal energy and aerosol mass. The facility is instrumented to measure volumetric flow rate, temperature, pressure, and aerosol concentration throughout the system. Aerosol release rates and mass accumulation on filters also are measured. We have performed a series of experiments in which a known rate of thermal energy is injected into the system. We then simulated this experiment with the FIRAC code. This paper compares and discusses the gas-dynamic and heat transport data obtained from the ventilation system experiments with those predicted by the FIRAC code. The numerically predicted data generally are within 10% of the experimental data

  3. Source term and radiological consequences of the Chernobyl accident

    International Nuclear Information System (INIS)

    Mourad, R.

    1987-09-01

    This report presents the results of a study of the source term and radiological consequences of the Chernobyl accident. The results two parts. The first part was performed during the first 2 months following the accident and dealt with the evaluation of the source term and an estimate of individual doses in the European countries outside the Soviet Union. The second part was performed after August 25-29, 1986 when the Soviets presented in a IAEA Conference in Vienna detailed information about the accident, including source term and radiological consequences in the Soviet Union. The second part of the study reconfirms the source term evaluated in the first part and in addition deals with the radiological consequences in the Soviet Union. Source term and individual doses are calculated from measured post-accident data, reported by the Soviet Union and European countries, microcomputer program PEAR (Public Exposure from Accident Releases). 22 refs

  4. The consequences of the Chernobyl nuclear accident in Greece

    International Nuclear Information System (INIS)

    1986-07-01

    In this report the radioactive fallout on Greece from the Chernobyl nuclear accident is described. The flow pattern to Greece of the radioactive materials released, the measurements performed on environmental samples and samples of the food chain, as well as some estimations of the population doses and of the expected consequences of the accident are presented. The analysis has shown that the radiological impact of the accident in Greece can be considered minor. (J.K.)

  5. Consequences of the Chernobyl accident in Lithuania

    International Nuclear Information System (INIS)

    Mastauskas, A.; Nedvecktaite, T.; Filistovic, V.

    1997-01-01

    After the Chernobyl accident of 26 April, 1986, population dose assessment favours the view that the radiation risk of population effected by the early fallout would be different from that in regions contaminated later. Taking into account the short half-time of the most important radioactive iodine isotopes, thyroid disorders would be expected mainly to follow the early fallout distribution. At the time of accident at Unite 4 of the Chernobyl NPP, surface winds were from the Southeast. The initial explosions and heat carried volatile radioactive materials to the 1,5 km height, from where they were transported over the Western part of Belarus, Southern and Western part of Lithuania toward Scandinavian countries. Thus the volatile radioiodine and some other radionuclides were detected in Lithuania on the very first days after the accident. The main task of the work - to conduct short Half-time radioiodine and long half-time radiocesium dose assessment of Lithuanian inhabitants a result of the early Chernobyl accident fallout

  6. The Chernobyl accident: Causes and consequences

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1987-01-01

    Two explosions, one immediately following the other, in Unit 4 of the Chernobyl nuclear power station in the Soviet Union signaled the worst disaster ever to befall the commercial nuclear power production industry. This accident, which occurred at 1:24 a.m. on April 26, 1986, resulted from an almost incredible series of operational errors associated, ironically, with an attempt to enhance the capability of the reactor to safely accommodate station blackout accidents (i.e., accidents arising from a loss of station electrical power). Disruption of the core, due to a prompt criticality excursion, resulted in the destruction of the core vault and reactor building and the sudden dispersal of about 3% of the fuel from the core region into the environment. Lesser but significant releases of radioactivity continued through May 6, 1986, before attempts to certain the radioactivity and cool the remnants of the core were successful. The amount and composition of material released in the course of the accident remain somewhat uncertain, and inconsistencies in the release estimates are evident. The Soviet estimates, in addition to the dispersal of about 3% of the fuel, include complete release of the noble gas core inventory, 20% of the fission product iodine inventory, 15% of the tellurium inventory, and 10 to 13% of the fission product cesium inventory. The iodine and cesium release estimates are not consistent with the noble gas values, and are as much as a factor of two less than some estimates made by experts outside the Soviet Union

  7. Probabilistic Assessment of Severe Accident Consequence in West Bangka

    Science.gov (United States)

    Sunarko; Su'ud, Zaki

    2017-07-01

    Probabilistic dose assessment for severe accident condition is performed for West Bangka area. Source-term from WASH-1400 reactor analysis is used as a conservative release scenario for 1000 MWe PWR. Seven groups of isotopes are used in the simulation based on core inventory and release fraction. Population distribution for Muntok district and the area within a 100 km radius is obtained from 2014 data. Meteorological data is provided through cyclic sampling from a database containing two-year site-specific hourly records in 2014-2015 periods. PC-COSYMA segmented plume dispersion code is used to investigate the assumed the consequence of the accident scenario. The result indicates that early or deterministic effect is important for areas close the release point while long-term or stochastic effect is related to population distribution and covers area of up to 100 km from the release point. The mean annual expected values for early mortality and late mortality for the population within 100 km radius from Muntok site are 2.38×10-4 yr -1 and 1.33×10-3 yr -1 respectively.

  8. Application of the severe accident code ATHLET-CD. Modelling and evaluation of accident management measures (Project WASA-BOSS)

    Energy Technology Data Exchange (ETDEWEB)

    Wilhelm, Polina; Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Schaefer, Frank [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety

    2016-07-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. Numerical analyses are used to investigate the accident progression and the complex physical phenomena during the core degradation phase, as well as to evaluate the effectiveness of possible countermeasures in the preventive and mitigative domain [1, 2]. The presented analyses have been performed with the computer code ATHLET-CD developed by GRS [3, 4].

  9. Consequences in Sweden of the Chernobyl accident

    International Nuclear Information System (INIS)

    Haegg, Conny.

    1990-01-01

    The radiation doses to man in Sweden due to the Chernobyl accident originate mainly from external irradiation from deposited radionuclides and internal irradiation from consumption of radioactively contaminated food stuffs. Inhalation and external irradiation from the passing cloud give only a minor contribution to the total dose. As an average for the Swedish population the individual radiation dose during the first year amounts to about 0.1 mSv, i.e. 10% of the natural background radiation. In the most contaminated areas, however, the individual dose may become 30 times higher than the average dose. The dose committed over 50 years has estimated to be about six times as high as the first year dose. The collective dose for the Swedish population has been estimated to about 1300 manSv the first year after the accident and the corresponding dose over 50 years to 5000 to 7000 manSv. This could lead to 100 to 200 extra fatal cancers. Furthermore, no damages on man that can be related to Chernobyl fallout, e.g. pre-natal effects, have so far been observed in Sweden. Shortly after the accident, several research projects were initiated in Sweden in order to follow the distribution of radionuclides in the aquatic and terrestrial environment. The results which in many cases are preliminary, shows that the recovery of the ecosystem will take several decades. (author)

  10. Information on economic and social consequences of the Chernobyl accident

    International Nuclear Information System (INIS)

    1990-07-01

    This ''Information on economic and social consequences of the Chernobyl accident'' was presented to the July 1990 session of the Economic and Social Council of the United Nations by the delegations of the Union of Soviet Socialist Republics, the Byelorussian Soviet Socialist Republic and the Ukrainian Soviet Socialist Republic. It presents the radiation situation, the medical aspects of the accident, the evacuation of the inhabitants from areas affected by radioactive contamination and their social welfare, the agro-industrial production and forestry in these areas, the decontamination operations, the scientific back-up for the work dealing with the consequences of the accident and the expenditure and losses resulting from the Chernobyl disaster

  11. Assessment of ICARE/CATHARE V1 Severe Accident Code

    International Nuclear Information System (INIS)

    Chatelard, Patrick; Fleurot, Joelle; Marchand, Olivier; Drai, Patrick

    2006-01-01

    The ICARE/CATHARE code system has been developed by the French 'Institut de Radioprotection et de Surete Nucleaire' (IRSN) in the last decade for the detailed evaluation of Severe Accident (SA) consequences in a primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermal-hydraulics French code CATHARE2. It has been extensively used to support the level 2 Probabilistic Safety Assessment (PSA-2) of the 900 MWe PWR. This paper presents the synthesis of the ICARE/CATHARE V1 assessment which was conducted in the frame of the 'International ICARE/CATHARE Users' Club', under the management of IRSN. The ICARE/CATHARE V1 validation matrix is composed of more than 60 experiments, distributed in few thermal-hydraulics non-regression tests (to handle the front end phase of a severe accident), numerous Separate-Effect Tests, about 30 Integral Tests covering both the early and the late degradation phases, as well as a 'circuit' experiment including hydraulics loops. Finally, the simulation of the TMI-2 accident was also added to assess the code against real conditions. This validation task was aimed at assessing the ICARE/CATHARE V1 capabilities (including the stand-alone ICARE2 V3mod1 version) and also at proposing recommendations for an optimal use of this version ('Users' Guidelines'). Thus, with a correct account for the recommended guidelines, it appeared that the last ICARE/CATHARE V1 version could be reasonably used to perform best-estimate reactor studies up to a large corium slumping into the lower head. (authors)

  12. Radiological attacks and accidents. Medical consequences

    International Nuclear Information System (INIS)

    Sakuta, Hidenari

    2007-01-01

    Probability of the occurrence of radiological attacks appears to be elevated after the terrorist attacks against the United States on September 11 in 2001. There are a lot of scenarios of radiological attack: simple radiological device, radiological disperse device (RDD or dirty bomb), attacks against nuclear reactor, improvised nuclear device, and nuclear weapons. Of these, RDD attack is the most probable scenario, because it can be easily made and can generate enormous psychological and economic damages. Radiological incidents are occurring to and fro in the world, including several cases of theft to nuclear facilities and unsuccessful terrorist attacks against them. Recently, a former Russian spy has allegedly been killed using polonium-210. In addition, serious radiological accidents have occurred in Chernobyl, Goiania, and Tokai-mura. Planning, preparation, education, and training exercise appear to be essential factors to cope with radiological attacks and accidents effectively without feeling much anxiety. Triage and psychological first aid are prerequisite to manage and provide effective medial care for mass casualties without inducing panic. (author)

  13. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-11-01

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

  16. Economic consequences assessment for scenarios and actual accidents do the same methods apply

    International Nuclear Information System (INIS)

    Brenot, J.

    1991-01-01

    Methods for estimating the economic consequences of major technological accidents, and their corresponding computer codes, are briefly presented with emphasis on the basic choices. When applied to hypothetic scenarios, those methods give results that are of interest for risk managers with a decision aiding perspective. Simultaneously the various costs, and the procedures for their estimation are reviewed for some actual accidents (Three Mile Island, Chernobyl,..). These costs are used in a perspective of litigation and compensation. The comparison of the methods used and cost estimates obtained for scenarios and actual accidents shows the points of convergence and discrepancies that are discussed

  17. Modeling of pipe break accident in a district heating system using RELAP5 computer code

    International Nuclear Information System (INIS)

    Kaliatka, A.; Valinčius, M.

    2012-01-01

    Reliability of a district heat supply system is a very important factor. However, accidents are inevitable and they occur due to various reasons, therefore it is necessary to have possibility to evaluate the consequences of possible accidents. This paper demonstrated the capabilities of developed district heating network model (for RELAP5 code) to analyze dynamic processes taking place in the network. A pipe break in a water supply line accident scenario in Kaunas city (Lithuania) heating network is presented in this paper. The results of this case study were used to demonstrate a possibility of the break location identification by pressure decrease propagation in the network. -- Highlights: ► Nuclear reactor accident analysis code RELAP5 was applied for accident analysis in a district heating network. ► Pipe break accident scenario in Kaunas city (Lithuania) district heating network has been analyzed. ► An innovative method of pipe break location identification by pressure-time data is proposed.

  18. Waste Evaporator Accident Simulation Using RELAP5 Computer Code

    International Nuclear Information System (INIS)

    POLIZZI, L.M.

    2004-01-01

    An evaporator is used on liquid waste from processing facilities to reduce the volume of the waste through heating the waste and allowing some of the water to be separated from the waste through boiling. This separation process allows for more efficient processing and storage of liquid waste. Commonly, the liquid waste consists of an aqueous solution of chemicals that over time could induce corrosion, and in turn weaken the tubes in the steam tube bundle of the waste evaporator that are used to heat the waste. This chemically induced corrosion could escalate into a possible tube leakage and/or the severance of a tube(s) in the tube bundle. In this paper, analyses of a waste evaporator system for the processing of liquid waste containing corrosive chemicals are presented to assess the system response to this accident scenario. This accident scenario is evaluated since its consequences can propagate to a release of hazardous material to the outside environment. It is therefore important to ensure that the evaporator system component structural integrity is not compromised, i.e. the design pressure and temperature of the system is not exceeded during the accident transient. The computer code used for the accident simulation is RELAP5-MOD31. The accident scenario analyzed includes a double-ended guillotine break of a tube in the tube bundle of the evaporator. A mitigated scenario is presented to evaluate the excursion of the peak pressure and temperature in the various components of the evaporator system to assess whether the protective actions and controls available are adequate to ensure that the structural integrity of the evaporator system is maintained and that no atmospheric release occurs

  19. Applicability of simplified methods to evaluate consequences of criticality accident using past accident data

    International Nuclear Information System (INIS)

    Nakajima, Ken

    2003-01-01

    Applicability of four simplified methods to evaluate the consequences of criticality accident was investigated. Fissions in the initial burst and total fissions were evaluated using the simplified methods and those results were compared with the past accident data. The simplified methods give the number of fissions in the initial burst as a function of solution volume; however the accident data did not show such tendency. This would be caused by the lack of accident data for the initial burst with high accuracy. For total fissions, simplified almost reproduced the upper envelope of the accidents. However several accidents, which were beyond the applicable conditions, resulted in the larger total fissions than the evaluations. In particular, the Tokai-mura accident in 1999 gave in the largest total specific fissions, because the activation of cooling system brought the relatively high power for a long time. (author)

  20. Consequence of potential accidents in heavy water plants

    International Nuclear Information System (INIS)

    Croitoru, C.; Lazar, R.E.; Preda, I.A.; Dumitrescu, M.

    1998-01-01

    Heavy water plants realize the primary isotopic concentrations of water using H 2 O-H 2 S chemical exchange and they are chemical plants. As these plants are handling and spreading large quantities of hydrogen sulphide (high toxic, corrosive, flammable and explosive as) maintained in the process at relative high temperatures and pressures, it is required an assessing of risks associated with the potential accidents. The H 2 S released in atmosphere as a result of an accident will have negative consequences to property, population and environment. This paper presents a model of consequences quantitative assessment and its outcome for the most dangerous accident in heavy water plants. Several states of the art risk based methods were modified and linked together to form a proper model for this analyse. Five basic steps to identify the risks involved in operating the plants are followed: hazard identification, accident sequence development, H 2 S emissions calculus, dispersion analyses and consequences determination. A brief description of each step and some information of analysis results are provided. The accident proportions, the atmospheric conditions and the population density in the respective area were accounted for consequences calculus. The specific results of the consequences analysis allow to develop the plant's operating safety requirements so that the risk remain at an acceptable level. (authors)

  1. The Chernobyl accident: The consequences in Europe

    International Nuclear Information System (INIS)

    Simmonds, J.R.

    1986-01-01

    The accidental release of radioactive material from the Chernobyl nuclear reactor in the USSR led to widespread contamination over Europe. The pattern of the contamination was determined by the weather conditions which occurred during the days when the release was continuing. Actual levels depended on a number of factors including the distance and direction from Chernobyl, rainfall during the passage of the radioactive cloud and local conditions such as topography. The highest levels of radioactivity have been found in parts of Scandanavia, which was affected by the early stages of the release, and in areas where it rained during the passage of the plume e.g. in parts of Italy, Greece and West Germany. Following the release of radionuclides to atmosphere people will be irradiated by a number of different routes. While the cloud is overhead people will be exposed to external irradiation from material in the cloud and internal irradiation following inhalation of the material. Radionuclides are removed from the cloud during transit and deposited on the ground. People are then exposed by other routes, notably external irradiation from the deposited material and the transfer of material through the terrestrial environment to foods consumed by people. These four exposure pathways are the most important in estimating the radiation doses received by the European population due to the Chernobyl accident. Environmental data are required to estimate the radiation doses. Such data are collected in all European countries by national authorities following the Chernobyl accident. In East Europe measurement data supplied by the national authorities were supplemented by information obtained by using the British embassies. The Embassies were supplied with instruments to measure external γ dose rates and they also collected food samples for analysis at NRPB. Various countermeasures were introduced in different countries to reduce exposure. These measures included restrictions on

  2. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    Brown, J. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)] [and others

    1997-06-01

    This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  3. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 1: Main report

    International Nuclear Information System (INIS)

    Brown, J.; Goossens, L.H.J.; Kraan, B.C.P.

    1997-06-01

    This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This document reports on an ongoing project to assess uncertainty in the MACCS and COSYMA calculations for the offsite consequences of radionuclide releases by hypothetical nuclear power plant accidents. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain variables that affect calculations of offsite consequences. The expert judgment elicitation procedure and its outcomes are described in these volumes. Other panels were formed to consider uncertainty in other aspects of the codes. Their results are described in companion reports. Volume 1 contains background information and a complete description of the joint consequence uncertainty study. Volume 2 contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures for both panels, (3) the rationales and results for the panels on soil and plant transfer and animal transfer, (4) short biographies of the experts, and (5) the aggregated results of their responses

  4. Radioecological and dosimetric consequences of Chernobyl accident in France

    International Nuclear Information System (INIS)

    Renaud, Ph.; Beaugelin, K.; Maubert, H.; Ledenvic, Ph.

    1997-01-01

    After ten years and the taking in account of numerous data, it can be affirmed that the dosimetric consequences of Chernobyl accident will have been limited in France. for the period 1986-2046, the individual middle efficient dose commitment, for the area the most reached by depositing is inferior to 1500 μSv, that represents about 1% of middle natural exposure in the same time. but mountains and forests can have more important surface activities than in plain. Everywhere else, it can be considered that the effects of Chernobyl accident are disappearing. the levels of cesium 137 are now often inferior to what they were before the accident. (N.C.)

  5. The Fukushima accident: radiological consequences and first lessons. Proceedings

    International Nuclear Information System (INIS)

    2012-02-01

    This document brings together the available presentations given at the conference organised by the French society of radiation protection about the Fukushima accident, its radiological consequences and the first lessons learnt. Sixteen presentations (slides) are compiled in this document and deal with: 1 - Accident progress and first actions (Thierry Charles, IRSN); 2 - Conditions and health monitoring of the Japanese intervention teams (Bernard Le Guen, EDF); 3 - The Intra Group action after the Fukushima accident (Michel Chevallier, Groupe Intra; Frederic Mariotte, CEA); 4 - Processing of effluents (Georges Pagis, Areva); 5 - Fukushima accident: impact on the terrestrial environment in Japan (Didier Champion, IRSN); 6 - Consequences of the Fukushima accident on the marine environment (Dominique Boust, IRSN); 7 - Territories decontamination perspectives (Pierre Chagvardieff, CEA); 8 - Actions undertaken by Japanese authorities (Florence Gallay, ASN); 9 - Japanese population monitoring and health stakes (Philippe Pirard, InVS); 10 - Citizen oversight actions implemented in Japan (David Boilley, ACRO); 11 - Implementation of ICRP's (International Commission on Radiological Protection) recommendations by Japanese authorities: first analysis (Jacques Lochard, CIPR); 12 - Control of Japan imported food stuff (David Brouque, DGAL); 13 - Questions asked by populations in France and in Germany (Florence-Nathalie Sentuc, GRS; Pascale Monti, IRSN); 14 - Labour law applicable to French workers working abroad (Thierry Lahaye, DGT); 15 - Protection of French workers working in Japan, Areva's experience (Patrick Devin, Areva); 16 - Fukushima accident experience feedback and post-accident nuclear doctrine (Jean-Luc Godet, ASN)

  6. Incidence Probability of Delayed Health Consequences of the Chernobyl Accident

    International Nuclear Information System (INIS)

    Abdel-Ghani, A.H.; El-Naggar, A.M.; El-Kadi, A.A.

    2000-01-01

    During the first international Conference on the long -term consequences of the Chernobyl disaster in 1995 at Kiev, and also during the 1996 International Conference at Vienna, Summing up the consequences of the Chernobyl accident, the data regarding the delayed health consequences were mainly related to thyroid cancer, hereditary disorders, general morbidity, mortality and psychological disturbances. Contrary to expectations, the incidences of Leukemia and Soft Tissue tumors were similar to the spontaneous incident. The expected delayed effects, however, among the accident survivors, the liquidators and populations resident in contaminated areas would show higher incidence probability to Leukemia. These population groups have been continuously exposed to low level radiation both externally and internally. Application of the new ICRP concept of radiation-induced Detriment, and the Nominal Probability Coefficient for Cancer and hereditary effects for both workers and populations are used as the rationale to calculate the incidence probability of occurrence of delayed health effects of the Chernobyl accidents

  7. Consequences of potential accidents in heavy water plants

    International Nuclear Information System (INIS)

    Croitoru, C.; Lazar, R.E.; Preda, I.A.; Dumitrescu, M.

    2002-01-01

    Heavy water plants achieve the primary isotopic concentration by H 2 O-H 2 S chemical exchange. In these plants are stored large quantities of hydrogen sulphide (high toxic, corrosive, flammable and explosive) maintained in process at relative high temperatures and pressures. It is required an assessment of risks associated with the potential accidents. The paper presents adopted model for quantitative consequences assessment in heavy water plants. Following five basic steps are used to identify the risks involved in plants operation: hazard identification, accident sequences development, H 2 S emissions calculus, dispersion analyses and consequences determination. A brief description of each step and some information from risk assessment for our heavy water pilot plant are provided. Accident magnitude, atmospheric conditions and population density in studied area were accounted for consequences calculus. (author)

  8. First international workshop on severe accidents and their consequences. [Chernobyl Accident

    Energy Technology Data Exchange (ETDEWEB)

    1989-07-01

    An international workshop on past severe nuclear accidents and their consequences was held in Dagomys region of Sochi, USSR on October 30--November 3, 1989. The plan of this meeting was approved by the USSR Academy of Sciences and by the USSR State Committee of the Utilization of Atomic Energy. The meeting was held under the umbrella of the ANS-SNS agreement of cooperation. Topics covered include analysis of the Chernobyl accident, safety measures for RBMK type reactors and consequences of the Chernobyl accident including analysis of the ecological, genetic and psycho-social factors. Separate reports are processed separately for the data bases. (CBS)

  9. Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

    International Nuclear Information System (INIS)

    2013-12-01

    Requests for severe accident investigations and assurance of mitigation measures have increased for operating nuclear power plants and the design of advanced nuclear power plants. Severe accident analysis investigations necessitate the analysis of the very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. The IAEA organizes coordinated research projects (CRPs) to facilitate technology development through international collaboration among Member States. The CRP on Benchmarking Severe Accident Computer Codes for HWR Applications was planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). This publication summarizes the results from the CRP participants. The CRP promoted international collaboration among Member States to improve the phenomenological understanding of severe core damage accidents and the capability to analyse them. The CRP scope included the identification and selection of a severe accident sequence, selection of appropriate geometrical and boundary conditions, conduct of benchmark analyses, comparison of the results of all code outputs, evaluation of the capabilities of computer codes to predict important severe accident phenomena, and the proposal of necessary code improvements and/or new experiments to reduce uncertainties. Seven institutes from five countries with HWRs participated in this CRP

  10. Dispersion parameters: impact on calculated reactor accident consequences

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D.C.

    1979-01-01

    Much attention has been given in recent years to the modeling of the atmospheric dispersion of pollutants released from a point source. Numerous recommendations have been made concerning the choice of appropriate dispersion parameters. A series of calculations has been performed to determine the impact of these recommendations on the calculated consequences of large reactor accidents. Results are presented and compared in this paper.

  11. Hanford Waste Tank Bump Accident and Consequence Analysis

    International Nuclear Information System (INIS)

    BRATZEL, D.R.

    2000-01-01

    This report provides a new evaluation of the Hanford tank bump accident analysis and consequences for incorporation into the Authorization Basis. The analysis scope is for the safe storage of waste in its current configuration in single-shell and double-shell tanks

  12. The accident of Chernobylsk-4 reactor and its consequences

    International Nuclear Information System (INIS)

    1986-01-01

    This report deals with the particulars of the accident as communicated by the Soviet delegation at an IAEA meeting by the and of August 1986. It was stated that the consequences emanated from the inherent instability of the design of the reactor, the deviation from the safety rules by the operators and the lack of a sight reactor containment. (G.B.)

  13. Radiological Consequence Analyses Following a Hypothetical Severe Accident in Japan

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  14. A simplified model for calculating early offsite consequences from nuclear reactor accidents

    International Nuclear Information System (INIS)

    Madni, I.K.; Cazzoli, E.G.; Khatib-Rahbar, M.

    1988-07-01

    A personal computer-based model, SMART, has been developed that uses an integral approach for calculating early offsite consequences from nuclear reactor accidents. The solution procedure uses simplified meteorology and involves direct analytic integration of air concentration equations over time and position. This is different from the discretization approach currently used in the CRAC2 and MACCS codes. The SMART code is fast-running, thereby providing a valuable tool for sensitivity and uncertainty studies. The code was benchmarked against both MACCS version 1.4 and CRAC2. Results of benchmarking and detailed sensitivity/uncertainty analyses using SMART are presented. 34 refs., 21 figs., 24 tabs

  15. Study on severe accidents and countermeasures for WWER-1000 reactors using the integral code ASTEC

    International Nuclear Information System (INIS)

    Tusheva, P.; Schaefer, F.; Altstadt, E.; Kliem, S.; Reinke, N.

    2011-01-01

    The research field focussing on the investigations and the analyses of severe accidents is an important part of the nuclear safety. To maintain the safety barriers as long as possible and to retain the radioactivity within the airtight premises or the containment, to avoid or mitigate the consequences of such events and to assess the risk, thorough studies are needed. On the one side, it is the aim of the severe accident research to understand the complex phenomena during the in- and ex-vessel phase, involving reactor-physics, thermal-hydraulics, physicochemical and mechanical processes. On the other side the investigations strive for effective severe accident management measures. This paper is focused on the possibilities for accident management measures in case of severe accidents. The reactor pressure vessel is the last barrier to keep the molten materials inside the reactor, and thus to prevent higher loads to the containment. To assess the behaviour of a nuclear power plant during transient or accident conditions, computer codes are widely used, which have to be validated against experiments or benchmarked against other codes. The analyses performed with the integral code ASTEC cover two accident sequences which could lead to a severe accident: a small break loss of coolant accident and a station blackout. The results have shown that in case of unavailability of major active safety systems the reactor pressure vessel would ultimately fail. The discussed issues concern the main phenomena during the early and late in-vessel phase of the accident, the time to core heat-up, the hydrogen production, the mass of corium in the reactor pressure vessel lower plenum and the failure of the reactor pressure vessel. Additionally, possible operator's actions and countermeasures in the preventive or mitigative domain are addressed. The presented investigations contribute to the validation of the European integral severe accidents code ASTEC for WWER-1000 type of reactors

  16. The Fukushima accident and its consequences. Facts, explanations and comments

    International Nuclear Information System (INIS)

    2012-01-01

    This document proposes an overview of the present situation in the different reactors of the Fukushima power station and discusses its control by the operator. It also describes what went on, the causes of the accident, and what occurred on the accident day (earthquake, tsunami, flooding). It discusses whether some mistakes regarding the design and the protection of reactors could explain the accident. It presents the various measures which have been immediately implemented to protect the populations and to confine the accident. It proposes an assessment of damages for the ground and marine environment in terms of contamination. It addresses the consequences of the released radioactivity on population health and on personnel intervening within the site. It discusses the restoration perspectives for contaminated areas and the possible return of evacuated population. Then, it describes the different phases for the station dismantling. It evokes the issue of fallouts beyond Japan and in Europe, outlines some lessons learned from the accident and new safety measures to be implemented in France. It discusses how nuclear risk management is organised in France and its efficiency. It addresses the consequences for the development of nuclear energy in the world

  17. Probabilistic accident consequence uncertainty analysis -- Early health effects uncertainty assessment. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [Univ. of New Mexico, Albuquerque, NM (United States); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA early health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on early health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  18. Probabilistic accident consequence uncertainty analysis -- Late health effects uncertain assessment. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Little, M.P.; Muirhead, C.R. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA late health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the expert panel on late health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  19. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for deposited material and external doses. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Boardman, J. [AEA Technology (United Kingdom); Jones, J.A. [National Radiological Protection Board (United Kingdom); Harper, F.T.; Young, M.L. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1997-12-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA deposited material and external dose models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on deposited material and external doses, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  20. Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for internal dosimetry. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Harrison, J.D. [National Radiological Protection Board (United Kingdom); Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States)

    1998-04-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library of uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA internal dosimetry models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on internal dosimetry, (4) short biographies of the experts, and (5) the aggregated results of their responses.

  1. Cost-effectiveness analysis of countermeasures using accident consequence assessment models

    International Nuclear Information System (INIS)

    Alonso, A.; Gallego, E.

    1987-01-01

    In the event of a large release of radionuclides from a nuclear power plant, protective actions for the population potentially affected must be implemented. Cost-effectiveness analysis will be useful to define the countermeasures and the criteria needed to implement them. This paper shows the application of Accident Consequence Assessment (ACA) models to cost-effectiveness analysis of emergency and long-term countermeasures, making use of the different relationships between dose, contamination levels, affected areas and population distribution, included in such a model. The procedure is illustrated with the new Melcor Accident Consequence Code System (MACCS 1.3), developed at Sandia National Laboratories (USA), for a fixed accident scenario. Different alternative actions are evaluated with regard to their radiological and economical impact, searching for an 'optimum' strategy. (author)

  2. Benchmarking severe accident computer codes for heavy water reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J.H. [International Atomic Energy Agency, Vienna (Austria)

    2010-07-01

    Consideration of severe accidents at a nuclear power plant (NPP) is an essential component of the defence in depth approach used in nuclear safety. Severe accident analysis involves very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. International cooperative research programmes are established by the IAEA in areas that are of common interest to a number of Member States. These co-operative efforts are carried out through coordinated research projects (CRPs), typically 3 to 6 years in duration, and often involving experimental activities. Such CRPs allow a sharing of efforts on an international basis, foster team-building and benefit from the experience and expertise of researchers from all participating institutes. The IAEA is organizing a CRP on benchmarking severe accident computer codes for heavy water reactor (HWR) applications. The CRP scope includes defining the severe accident sequence and conducting benchmark analyses for HWRs, evaluating the capabilities of existing computer codes to predict important severe accident phenomena, and suggesting necessary code improvements and/or new experiments to reduce uncertainties. The CRP has been planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for HWRs. (author)

  3. Accident consequence calculations for project W-058 safety analysis

    International Nuclear Information System (INIS)

    Van Keuren, J.C.

    1997-01-01

    This document describes the calculations performed to determine the accident consequences for the W-058 safety analysis. Project W-058 is the replacement cross site transfer system (RCSTS), which is designed to transort liquid waste between the 200 W and 200 E areas. Calculations for RCSTS safety analyses used the same methods as the calculations for the Tank Waste Remediation System (TWRS) Basis for Interim Operation (BIO) and its supporting calculation notes. Revised analyses were performed for the spray and pool leak accidents since the RCSTS flows and pressures differ from those assumed in the TWRS BIO. Revision 1 of the document incorporates review comments

  4. Consequences of the Chernobyl accident in France. Thematic sheets

    International Nuclear Information System (INIS)

    2006-01-01

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

  5. Severe accident modeling and offsite dose consequence evaluations for nuclear power plant emergency planning

    Energy Technology Data Exchange (ETDEWEB)

    Chen, S.H.; Feng, T.S.; Huang, K.C. [National Tsing-Hua Univ., Hsinchu, Taiwan (China); Wang, J.R. [Inst. of Nuclear Energy Research, Longtan, Taiwan (China); Cheng, Y.H. [Industrial Tech. Res. Inst., Hsinchu, Taiwan (China); Shih, C., E-mail: ckshih@ess.nthu.edu.tw [National Tsing-Hua Univ., Hsinchu, Taiwan (China)

    2011-07-01

    We have investigated the roles of Firewater Addition System and Passive Flooder in ABWR severe accidents, such as LOCA and SBO. The results are apparent that Firewater System is vital in the highly unlikely situation where all AC are lost. Also in this paper, we present EPZDose, an effective and faster-than-real time code for offsite dose consequences predictions and evaluations. Illustrations with the release from our severe accident scenario show friendly and informative user's interface for supporting decision makings in nuclear emergency situations. (author)

  6. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, main report

    International Nuclear Information System (INIS)

    Harper, F.T.; Young, M.L.; Miller, L.A.; Hora, S.C.; Lui, C.H.; Goossens, L.H.J.; Cooke, R.M.; Paesler-Sauer, J.; Helton, J.C.

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The ultimate objective of the joint effort was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. Experts developed their distributions independently. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. To validate the distributions generated for the dispersion code input variables, samples from the distributions and propagated through the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the first of a three-volume document describing the project

  7. 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.

  8. 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

  9. Estimated consequences from severe spent nuclear fuel transportation accidents

    International Nuclear Information System (INIS)

    Arnish, J.J.; Monette, F.; LePoire, D.; Biwer, B.M.

    1996-01-01

    The RISKIND software package is used to estimate radiological consequences of severe accident scenarios involving the transportation of spent nuclear fuel. Radiological risks are estimated for both a collective population and a maximally exposed individual based on representative truck and rail cask designs described in the U.S. Nuclear Regulatory Commission (NRC) modal study. The estimate of collective population risk considers all possible environmental pathways, including acute and long-term exposures, and is presented in terms of the 50-y committed effective dose equivalent. Radiological risks to a maximally exposed individual from acute exposure are estimated and presented in terms of the first year and 50-y committed effective dose equivalent. Consequences are estimated for accidents occurring in rural and urban population areas. The modeled pathways include inhalation during initial passing of the radioactive cloud, external exposure from a reduction of the cask shielding, long-term external exposure. from ground deposition, and ingestion from contaminated food (rural only). The major pathways and contributing radionuclides are identified, and the effects of possible mitigative actions are discussed. The cask accident responses and the radionuclide release fractions are modeled as described in the NRC modal study. Estimates of severe accident probabilities are presented for both truck and rail modes of transport. The assumptions made in this study tend to be conservative; however, a set of multiplicative factors are identified that can be applied to estimate more realistic conditions

  10. Health consequences of the Chernobyl accident: thyroid diseases

    International Nuclear Information System (INIS)

    Nagataki, Shigenobu; Ashizawa, Kiyoto

    1997-01-01

    An International Conference entitled 'One decade after Chernobyl: Summing up the consequences of the accident' was held at the Vienna from 8 to 12 April 1996. The aim of conference was to seek a common and conclusive understanding of the nature and magnitude of the consequences of the Chernobyl accident. It was concluded that a highly significant increase in the incidence of thyroid cancer among those persons in the affected areas who were children in 1986 is the only clear evidence to data of a public health impact of radiation exposure as a result of the Chernobyl accident and both temporal and geographical distributions clearly indicate a relationship of the increase in incidence to radiation exposure due to the Chernobyl accident. To clarify the relationship between thyroid cancer and radioactive fallout more clearly, a long term prospective study (case-control/cohort) should be conducted in the highly risk groups and the analysis of accurate estimation of exposure dose to external and/or internal radiation is needed. (author)

  11. Nuclear installations abroad the accident risks and their potential consequences

    International Nuclear Information System (INIS)

    Turvey, F.J.

    1996-01-01

    This paper endeavors to assess the threat to Ireland from severe accidents at civil nuclear installations. Among the various types of nuclear installations worldwide, reactors and reprocessing plants are considered to be the most threatening and so the paper focuses on these. The threat is assumed to be a function of the risk of severe accidents at the above types of installations and the probability of unfavourable weather conditions carrying the radioactive releases to Ireland. Although nuclear installations designed in eastern Europe and Asia are less safe than others, the greatest threat to Ireland arises from nearby installations in the UK. The difficulty of measuring the probabilities and consequences of severe nuclear accidents at nuclear installations in general is explained. In the case of the UK installations, this difficulty is overcome to some degree by using values of 'tolerable' risk adopted by the national nuclear regulator to define the radiotoxic releases from nuclear accidents. These are used as input to atmospheric dispersion models in which unfavourable weather conditions for Ireland are assumed and radiation doses are calculated to members of the Irish public. No countermeasures, such as sheltering, are assumed. In the worst cast scenario no deaths would be expected in Ireland in the immediate aftermath of the accident however, an increase in cancers over a period of 25 years or so would be expected assuming present-day models for the effect of low level radiation are valid

  12. Nuclear installations abroad the accident risks and their potential consequences

    Energy Technology Data Exchange (ETDEWEB)

    Turvey, F J [Radiological Protection Inst. of Ireland (Ireland)

    1996-10-01

    This paper endeavors to assess the threat to Ireland from severe accidents at civil nuclear installations. Among the various types of nuclear installations worldwide, reactors and reprocessing plants are considered to be the most threatening and so the paper focuses on these. The threat is assumed to be a function of the risk of severe accidents at the above types of installations and the probability of unfavourable weather conditions carrying the radioactive releases to Ireland. Although nuclear installations designed in eastern Europe and Asia are less safe than others, the greatest threat to Ireland arises from nearby installations in the UK. The difficulty of measuring the probabilities and consequences of severe nuclear accidents at nuclear installations in general is explained. In the case of the UK installations, this difficulty is overcome to some degree by using values of `tolerable` risk adopted by the national nuclear regulator to define the radiotoxic releases from nuclear accidents. These are used as input to atmospheric dispersion models in which unfavourable weather conditions for Ireland are assumed and radiation doses are calculated to members of the Irish public. No countermeasures, such as sheltering, are assumed. In the worst cast scenario no deaths would be expected in Ireland in the immediate aftermath of the accident however, an increase in cancers over a period of 25 years or so would be expected assuming present-day models for the effect of low level radiation are valid.

  13. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Cahalan, J.E.

    2009-01-01

    Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to

  14. INF Code related matters. Joint IAEA/IMO literature survey on potential consequences of severe maritime accidents involving the transport of radioactive material. 2 volumes. Vol. I - Report and publication titles. Vol. II - Relevant abstracts

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-10

    This literature survey was undertaken jointly by the International Maritime Organization (IMO) and the International Atomic Energy Agency (IAEA) as a step in addressing the subject of environmental impact of accidents involving materials subject to the IMO's Code for the Safe Carriage of Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes in Flasks on Board Ships, also known as the INF Code. The results of the survey are provided in two volumes: the first one containing the description of the search and search results with the list of generated publication titles, and the second volume containing the abstracts of those publications deemed relevant for the purposes of the literature survey. Literature published between 1980 and mid-1999 was reviewed by two independent consultants who generated publication titles by performing searches of appropriate databases, and selected the abstracts of relevant publications for inclusion in this survey. The IAEA operates INIS, the world's leading computerised bibliographical information system on the peaceful uses of nuclear energy. The acronym INIS stands for International Nuclear Information System. INIS Members are responsible for determining the relevant nuclear literature produced within their borders or organizational confines, and then preparing the associated input in accordance with INIS rules. INIS records are included in other major databases such as the Energy, Science and Technology database of the DIALOG service. Because it is the INIS Members, rather than the IAEA Secretariat, who are responsible for its contents, it was considered appropriate that INIS be the primary source of information for this literature review. Selected unpublished reports were also reviewed, e.g. Draft Proceedings of the Special Consultative Meeting of Entities involved in the maritime transport of materials covered by the INF Code (SCM 5), March 1996. Many of the formal papers at SCM 5 were included in the literature

  15. INF Code related matters. Joint IAEA/IMO literature survey on potential consequences of severe maritime accidents involving the transport of radioactive material. 2 volumes. Vol. I - Report and publication titles. Vol. II - Relevant abstracts

    International Nuclear Information System (INIS)

    2000-01-01

    This literature survey was undertaken jointly by the International Maritime Organization (IMO) and the International Atomic Energy Agency (IAEA) as a step in addressing the subject of environmental impact of accidents involving materials subject to the IMO's Code for the Safe Carriage of Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes in Flasks on Board Ships, also known as the INF Code. The results of the survey are provided in two volumes: the first one containing the description of the search and search results with the list of generated publication titles, and the second volume containing the abstracts of those publications deemed relevant for the purposes of the literature survey. Literature published between 1980 and mid-1999 was reviewed by two independent consultants who generated publication titles by performing searches of appropriate databases, and selected the abstracts of relevant publications for inclusion in this survey. The IAEA operates INIS, the world's leading computerised bibliographical information system on the peaceful uses of nuclear energy. The acronym INIS stands for International Nuclear Information System. INIS Members are responsible for determining the relevant nuclear literature produced within their borders or organizational confines, and then preparing the associated input in accordance with INIS rules. INIS records are included in other major databases such as the Energy, Science and Technology database of the DIALOG service. Because it is the INIS Members, rather than the IAEA Secretariat, who are responsible for its contents, it was considered appropriate that INIS be the primary source of information for this literature review. Selected unpublished reports were also reviewed, e.g. Draft Proceedings of the Special Consultative Meeting of Entities involved in the maritime transport of materials covered by the INF Code (SCM 5), March 1996. Many of the formal papers at SCM 5 were included in the literature

  16. Modeling of criticality accidents and their environmental consequences

    International Nuclear Information System (INIS)

    Thomas, W.; Gmal, B.

    1987-01-01

    In the Federal Republic of Germany, potential radiological consequences of accidental nuclear criticality have to be evaluated in the licensing procedure for fuel cycle facilities. A prerequisite to this evaluation is to establish conceivable accident scenarios. First, possibilities for a criticality exceeding the generally applied double contingency principle of safety are identified by screening the equipment and operation of the facility. Identification of undetected accumulations of fissile material or incorrect transfer of fissile solution to unfavorable geometry normally are most important. Second, relevant and credible scenarios causing the most severe consequences are derived from these possibilities. For the identified relevant scenarios, time-dependent fission rates and reasonable numbers for peak power and total fissions must be determined. Experience from real accidents and experiments (KEWB, SPERT, CRAC, SILENE) has been evaluated using empirical formulas. To model the time-dependent behavior of criticality excursions in fissile solutions, a computer program FELIX has been developed

  17. Assessment of off-site consequences of nuclear accidents (MARIA)

    International Nuclear Information System (INIS)

    Haywood, S.M.

    1985-01-01

    A brief report is given of a workshop held in Luxembourg in 1985 on methods for assessing the off-site radiological consequences of nuclear accidents (MARIA). The sessions included topics such as atmospheric dispersion; foodchain transfer; urban contamination; demographic and land use data; dosimetry, health effects, economic and countermeasures models; uncertainty analysis; and application of probabilistic risk assessment results as input to decision aids. (U.K.)

  18. A radiological accident consequence assessment system for Hong Kong

    International Nuclear Information System (INIS)

    Wong, M.C.; Lam, H.K.

    1993-01-01

    An account is given of the Hong Kong Radiological Accident Consequence Assessment System which would be used to assess the potential consequences of an emergency situation involving atmospheric release of radioactive material. The system has the capability to acquire real-time meteorological information from the Observatory's network of automatic stations, synoptic stations in the nearby region as well as forecast data from numerical prediction models. The system makes use of these data to simulate the transport and dispersion of the released radioactive material. The effectiveness of protective action on the local population is also modeled. The system serves as a powerful aid in the protective action recommendation processes

  19. Application of RASCAL code for multiunit accident in domestic nuclear sites

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Hyun; Jeong, Seung Young [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-10-15

    All of domestic nuclear power plant sites are multiunit site (at least 5 - 6 reactors are operating), so this capability has to be quickly secured for nuclear licensee and institutes responsible for nuclear emergency response. In this study, source term and offsite dose from multiunit event were assessed using a computer code, RASCAL. An emergency exercise scenario was chosen to verify applicability of the codes to domestic nuclear site accident. Employing tools and new features of the code, such as merging more than two individual source terms and source term estimate for long term progression accident, main parameters and information in the scenario, release estimates and dose projections were performed. Radiological releases and offsite doses from multiunit accident were calculated using RASCAL.. A scenario, in which three reactors were damaged coincidently by a great natural disaster, was considered. Surrogate plants were chosen for the code calculation. Source terms of each damaged unit were calculated individually first, and then total source term and integrated offsite dose assessment data was acquired using a source term merge function in the code. Also comparison between LTSBO and LOCA source term estimate options was performed. Differences in offsite doses were caused by release characteristics. From LTSBO option, iodines were released much higher than LOCA. Also LTSBO source term release was delayed and the duration was longer than LOCA. This option would be useful to accidents which progress with much longer time frame than LOCA. RASCAL can be useful tool for radiological consequence assessment in domestic nuclear site accidents.

  20. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, appendices A and B

    International Nuclear Information System (INIS)

    Harper, F.T.; Young, M.L.; Miller, L.A.; Hora, S.C.; Lui, C.H.; Goossens, L.H.J.; Cooke, R.M.; Paesler-Sauer, J.; Helton, J.C.

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the second of a three-volume document describing the project and contains two appendices describing the rationales for the dispersion and deposition data along with short biographies of the 16 experts who participated in the project

  1. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment, appendices A and B

    Energy Technology Data Exchange (ETDEWEB)

    Harper, F.T.; Young, M.L.; Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States); Hora, S.C. [Univ. of Hawaii, Hilo, HI (United States); Lui, C.H. [Nuclear Regulatory Commission, Washington, DC (United States); Goossens, L.H.J.; Cooke, R.M. [Delft Univ. of Technology (Netherlands); Paesler-Sauer, J. [Research Center, Karlsruhe (Germany); Helton, J.C. [and others

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the second of a three-volume document describing the project and contains two appendices describing the rationales for the dispersion and deposition data along with short biographies of the 16 experts who participated in the project.

  2. Nuclear Accidents: Consequences for Human, Society and Energy Sector

    Directory of Open Access Journals (Sweden)

    L. A. Bolshov

    2016-01-01

    Full Text Available The article examines radiation and hygienic regulations with regard to the elimination of consequences of the Chernobyl NPP accident in the context of relationships with other aspects, primarily socio-economic and political factors. This experience is reasonable to take into account when defining criteria in other regulatory fields, for example, in radioactive waste classification and remediation of areas. The article presents an analysis of joint features and peculiarities of nuclear accidents in the industry and energy sectors. It is noted that the scale of global consequences of the Chernobyl NPP accident is defined by the large-scale release of radioactivity into the environment, as well as an affiliation of the nuclear installation with the energy sector. Large-scale radiation accidents affect the most diverse spheres of human activities, what, in its turn, evokes the reverse reaction from the society and its institutions, including involvement of political means of settlement. If the latter is seeing for criteria that are scientifically justified and feasible, then the preconditions for minimizing socio-economic impacts are created. In other cases, political decisions, such as nuclear units’ shutdown and phasing out of nuclear energy, appear to be an economic price which society, as a whole and a single industry sector, pay to compensate the negative public response. The article describes fundamental changes in approaches to ensure nuclear and radiation safety that occurred after the Chernobyl NPP accident. Multiple and negative consequences of the Chernobyl accident for human and society are balanced to some extent by a higher level of operational safety, emergency preparedness, and life-cycle safety. The article indicates that harmonization and ensuring consistency of regulations that involve different aspects of nuclear and radiation safety are important to implement practical solutions to the nuclear legacy problems. The

  3. Development of the criticality accident analysis code, AGNES

    International Nuclear Information System (INIS)

    Nakajima, Ken

    1989-01-01

    In the design works for the facilities which handle nuclear fuel, the evaluation of criticality accidents cannot be avoided even if their possibility is as small as negligible. In particular in the system using solution fuel like uranyl nitrate, solution has the property easily becoming dangerous form, and all the past criticality accidents occurred in the case of solution, therefore, the evaluation of criticality accidents becomes the most important item of safety analysis. When a criticality accident occurred in a solution fuel system, due to the generation and movement of radiolysis gas voids, the oscillation of power output and pressure pulses are observed. In order to evaluate the effect of criticality accidents, these output oscillation and pressure pulses must be calculated accurately. For this purpose, the development of the dynamic characteristic code AGNES (Accidentally Generated Nuclear Excursion Simulation code) was carried out. The AGNES is the reactor dynamic characteristic code having two independent void models. Modified energy model and pressure model, and as the benchmark calculation of the AGNES code, the results of the experimental analysis on the CRAC experiment are reported. (K.I.)

  4. A study into the consequences of a nuclear accident

    International Nuclear Information System (INIS)

    Arnott, D.G.

    1987-07-01

    The nuclear industry in Britain would like to believe, and would like the general public to believe, that major accidents such as that at Chernobyl in 1986, could no happen in Britain, because the design and operating procedure have been made as safe as possible. However, because the designers and operators are human, they can make mistakes. Some of these are mentioned; errors of design, errors of maintenance or inspection and errors of judgement. In spite of protestations to the contrary, a major accident could occur at Sizewell-B reactor. Given that this a real possibility, plans should be drawn up to prepare for the situation. The study considers the possible consequences of a nuclear accident under the headings, human error, how nuclear fission works, radioactivity, the truth about Chernobyl, what patterns of reactor accident are possible, what can be done (this includes meteorological information, the issuing of potassium iodate tables, radiation monitoring and evacuation). Practical issues which should concern the local authorities, especially Wrekin Council, are discussed and a recommendation made for an environmental protection officer to be appointed to keep the matter under continuing review. (U.K.)

  5. Tank Bump Accident Potential and Consequences During Waste Retrieval

    International Nuclear Information System (INIS)

    BRATZEL, D.R.

    2000-01-01

    This report provides an evaluation of Hanford tank bump accident potential and consequences during waste retrieval operations. The purpose of this report is to consider the best available new information to support recommendations for safety controls. A new tank bump accident analysis for safe storage (Epstein et al. 2000) is extended for this purpose. A tank bump is a postulated event in which gases, consisting mostly of water vapor, are suddenly emitted from the waste and cause tank headspace pressurization. Tank bump scenarios, physical models, and frequency and consequence methods are fully described in Epstein et al. (2000). The analysis scope is waste retrieval from double-shell tanks (DSTs) including operation of equipment such as mixer pumps and air lift circulators. The analysis considers physical mechanisms for tank bump to formulate criteria for bump potential during retrieval, application of the criteria to the DSTs, evaluation of bump frequency, and consequence analysis of a bump. The result of the consequence analysis is the mass of waste released from tanks; radiological dose is calculated using standard methods (Cowley et al. 2000)

  6. Uncertainty and sensitivity analysis in nuclear accident consequence assessment

    International Nuclear Information System (INIS)

    Karlberg, Olof.

    1989-01-01

    This report contains the results of a four year project in research contracts with the Nordic Cooperation in Nuclear Safety and the National Institute for Radiation Protection. An uncertainty/sensitivity analysis methodology consisting of Latin Hypercube sampling and regression analysis was applied to an accident consequence model. A number of input parameters were selected and the uncertainties related to these parameter were estimated within a Nordic group of experts. Individual doses, collective dose, health effects and their related uncertainties were then calculated for three release scenarios and for a representative sample of meteorological situations. From two of the scenarios the acute phase after an accident were simulated and from one the long time consequences. The most significant parameters were identified. The outer limits of the calculated uncertainty distributions are large and will grow to several order of magnitudes for the low probability consequences. The uncertainty in the expectation values are typical a factor 2-5 (1 Sigma). The variation in the model responses due to the variation of the weather parameters is fairly equal to the parameter uncertainty induced variation. The most important parameters showed out to be different for each pathway of exposure, which could be expected. However, the overall most important parameters are the wet deposition coefficient and the shielding factors. A general discussion of the usefulness of uncertainty analysis in consequence analysis is also given. (au)

  7. Fukushima accident: the consequences in Japan, France and in Japan

    International Nuclear Information System (INIS)

    Foucher, N.; Sorin, F.

    2011-01-01

    This document begins with a description of the Fukushima accident, the second article reviews the main consequences in Japan of the accident: setting of a forbidden zone around the plant, restriction of the exports of food products, or the shutdown of the Hamaoka plant. The third article is the reporting of an interview of L. Oursel, deputy general director of the Areva group, this interview deals mainly with the safety standard of the EPR and with the issue of passive safety systems. The last part of the document is dedicated to the consequences in France (null sanitary impact, cooperation between Areva, EdF, CEA and the Japanese plant operator Tepco...) and in the rest of the world: the organization of resistance tests in the nuclear power plants operating in the European Union, the decision about the agreement of EPR and AP1000 reactor has been delayed in United-Kingdom, acceleration of the German program for abandoning nuclear energy, Italy suspends its nuclear program, China orders a general overhaul of the safety standard of its nuclear power plants, Poland and Romania reaffirm their trust in nuclear energy, France wishes a 'mechanism' allowing a quick international intervention in case of major nuclear accident, Russia proposes measures to improve nuclear safety. (A.C.)

  8. Severe accident analysis code Sampson for impact project

    International Nuclear Information System (INIS)

    Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh

    2001-01-01

    Four years of the IMPACT project Phase 1 (1994-1997) had been completed with financial sponsorship from the Japanese government's Ministry of Economy, Trade and Industry. At the end of the phase, demonstration simulations by combinations of up to 11 analysis modules developed for severe accident analysis in the SAMPSON Code were performed and physical models in the code were verified. The SAMPSON prototype was validated by TMI-2 and Phebus-FP test analyses. Many of empirical correlation and conventional models have been replaced by mechanistic models during Phase 2 (1998-2000). New models for Accident Management evaluation have been also developed. (author)

  9. The Role of Countermeasures in Mitigating the Radiological Consequences of Nuclear Power Plant Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Tawfik, F. S.; Abdel-Aal, M.M., E-mail: basant572000@yahoo.com [Siting & Environmental Department, Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

    2014-10-15

    During the Fukushima accident the mitigation actions played an important role to decrease the consequences of the accident. The countermeasures are the actions that should be taken after the occurrence of a nuclear accident to protect the public against the associated risk. The actions may be represented by sheltering, evacuation, distribution of stable iodine tablets and/or relocation. This study represents a comprehensive probabilistic study to investigate the role of the adoption of the countermeasures in case of a hypothetical accident of type LOCA for a nuclear power plant of PWR (1000 Mw) type. This work was achieved through running of the PC COSYMA{sup (1)} code. The effective doses in different organs, short and long term health effects, and the associated risks were calculated with and without countermeasures. In addition, the overall costs of the accident and the costs of countermeasures are estimated which represent our first trials to know how much the postulated accident costs. The source term of a hypothetical accident is determined by knowing the activity of the core inventory. The meteorological conditions around the site in addition to the population distribution were utilized as input parameters. The stability conditions and the height of atmospheric boundary layers ABL of the concerned site were determined by developing a computer program utilizing Pasquill-Gifford atmospheric stability conditions. The results showed that, the area around the site requires early and late countermeasures actions after the accident especially in the downwind sectors. For late countermeasures, the duration of relocation ranged from about two to 10 years. The adoption of the countermeasures increases the costs of emergency planning by 40% but reduces the risk associated with the accident. (author)

  10. Possible consequences of severe accidents at the Lubiatowo site, Poland

    Science.gov (United States)

    Seibert, Petra; Philipp, Anne; Hofman, Radek; Gufler, Klaus; Sholly, Steven

    2014-05-01

    The construction of a nuclear power plant is under consideration in Poland. One of the sites under discussion is near Lubiatowo, located on the cost of the Baltic Sea northwest of Gdansk. An assessment of possible environmental consequences is carried out for 88 real meteorological cases with the Lagrangian particle dispersion model FLEXPART. Based on literature research, three reactor designs (ABWR, EPR, AP 1000) were identified as being under discussion in Poland. For each of the designs, a set of accident scenarios was evaluated and two source terms per reactor design were selected for analysis. One of the selected source terms was a relatively large release while the second one was a severe accident with an intact containment. Considered endpoints of the calculations are ground contamination with Cs-137 and time-integrated concentrations of I-131 in air as well as committed doses. They are evaluated on a grid of ca. 3 km mesh size covering eastern Central Europe.

  11. Applications of probabilistic accident consequence evaluation in Cuba

    International Nuclear Information System (INIS)

    Rodriguez, J.M.

    1996-01-01

    Are presented the approaches and results of the application of Accident Consequence Evaluation methodologies in on emergency in the Juragua Nuclear Power Plant site and a population evaluation of a planned NPP site in the east of the country Findings on population sector weighing and assessment of effectiveness of primary countermeasures in the event of sever accidents (SST1 and PWR4 source terms) in Juragua NPP site are discussed Results on comparative risk-based evaluation of the population predicted evolution (in 3 temporal horizons: base year, 2005 year and 2050 year) for the planned site are described. Evaluation also included sector risk weighing, risk importance of small towns in the nearby of the effects on risk of population freezing and relocation of these villages

  12. Reports of the Chernobyl accident consequences in Brazilian newspapers

    International Nuclear Information System (INIS)

    Vicente, Roberto; Oliveira, Rosana Lagua de

    2009-01-01

    The public perception of the risks associated with nuclear power plants was profoundly influenced by the accidents at Three Mile Island and Chernobyl Power Plants which also served to exacerbate in the last decades the growing mistrust on the 'nuclear industry'. Part of the mistrust had its origin in the arrogance of nuclear spokesmen and in the secretiveness of nuclear programs. However, press agencies have an important role in shaping and upsizing the public awareness against nuclear energy. In this paper we present the results of a survey in reports of some Brazilian popular newspapers on Chernobyl consequences, as measured by the total death toll of the accident, to show the up and down dance of large numbers without any serious judgment. (author)

  13. Environmental radiological consequences of a loss of coolant accident

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1981-01-01

    The elaboration of a calculation model to determine safety areas, named Exclusion Zone and Low Population Zone for nuclear power plants, is dealt with. These areas are determined from a radioactive doses calculation for the population living around the NPP after occurence of a postulated ' Maximum Credible Accident' (MCA). The MCA is defined as an accident with complete loss of primary coolant and consequent fusion of a substantial portion of the reactor core. In the calculations carried out, data from NPP Angra I were used and the assumptions made were conservative, to be compatible with licensing requirements. Under the most pessimistic assumption (no filters) the values of 410m and 1000m were obtained for the Exclusion Zone and Low Population Zone radii, respectivily. (Author) [pt

  14. The accident consequence model of the German safety study

    International Nuclear Information System (INIS)

    Huebschmann, W.

    1977-01-01

    The accident consequence model essentially describes a) the diffusion in the atmosphere and deposition on the soil of radioactive material released from the reactor into the atmosphere; b) the irradiation exposure and health consequences of persons affected. It is used to calculate c) the number of persons suffering from acute or late damage, taking into account possible counteractions such as relocation or evacuation, and d) the total risk to the population from the various types of accident. The model, the underlying parameters and assumptions are described. The bone marrow dose distribution is shown for the case of late overpressure containment failure, which is discussed in the paper of Heuser/Kotthoff, combined with four typical weather conditions. The probability distribution functions for acute mortality, late incidence of cancer and genetic damage are evaluated, assuming a characteristic population distribution. The aim of these calculations is first the presentation of some results of the consequence model as an example, in second the identification of problems, which need possibly in a second phase of study to be evaluated in more detail. (orig.) [de

  15. Computer codes developed in FRG to analyse hypothetical meltdown accidents

    International Nuclear Information System (INIS)

    Hassmann, K.; Hosemann, J.P.; Koerber, H.; Reineke, H.

    1978-01-01

    It is the purpose of this paper to give the status of all significant computer codes developed in the core melt-down project which is incorporated in the light water reactor safety research program of the Federal Ministry of Research and Technology. For standard pressurized water reactors, results of some computer codes will be presented, describing the course and the duration of the hypothetical core meltdown accident. (author)

  16. Primary disability of the Chernobyl Accident consequences liquidators

    International Nuclear Information System (INIS)

    Zubritskij, M.K.; Plakhotya, L.P.; Kalinina, T.V.; Zhilinskaya, E.I.

    1994-01-01

    The structure of courses of the primary invalidism of the Chernobyl accident consequences liquidators is studies. The main reasons of the loss of a capacity for work are blood circulation diseases (41.9%), neoplasms (19.9%), diseases of the nervous system and sense organs (9.7%), mental disorders (5.9%) and endocrine diseases (5.5%). The invalids distribution in the different regions and in different age groups according to the disease forms is analysed. The average durations of the diseases resulting in the primary invalidism are about 2.8 years. In average the illnesses began in the 3.1 years. 6 refs

  17. Preliminary results of consequence assessment of a hypothetical severe accident using Thai meteorological data

    Science.gov (United States)

    Silva, K.; Lawawirojwong, S.; Promping, J.

    2017-06-01

    Consequence assessment of a hypothetical severe accident is one of the important elements of the risk assessment of a nuclear power plant. It is widely known that the meteorological conditions can significantly influence the outcomes of such assessment, since it determines the results of the calculation of the radionuclide environmental transport. This study aims to assess the impacts of the meteorological conditions to the results of the consequence assessment. The consequence assessment code, OSCAAR, of Japan Atomic Energy Agency (JAEA) is used for the assessment. The results of the consequence assessment using Thai meteorological data are compared with those using Japanese meteorological data. The Thai case has following characteristics. Low wind speed made the radionuclides concentrate at the center comparing to the Japanese case. The squalls induced the peaks in the ground concentration distribution. The evacuated land is larger than the Japanese case though the relocated land is smaller, which is attributed to the concentration of the radionuclides near the release point.

  18. PERSPECTIVES ON A DOE CONSEQUENCE INPUTS FOR ACCIDENT ANALYSIS APPLICATIONS

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Thoman, D.C.; Lowrie, J.; Keller, A.

    2008-01-01

    Department of Energy (DOE) accident analysis for establishing the required control sets for nuclear facility safety applies a series of simplifying, reasonably conservative assumptions regarding inputs and methodologies for quantifying dose consequences. Most of the analytical practices are conservative, have a technical basis, and are based on regulatory precedent. However, others are judgmental and based on older understanding of phenomenology. The latter type of practices can be found in modeling hypothetical releases into the atmosphere and the subsequent exposure. Often the judgments applied are not based on current technical understanding but on work that has been superseded. The objective of this paper is to review the technical basis for the major inputs and assumptions in the quantification of consequence estimates supporting DOE accident analysis, and to identify those that could be reassessed in light of current understanding of atmospheric dispersion and radiological exposure. Inputs and assumptions of interest include: Meteorological data basis; Breathing rate; and Inhalation dose conversion factor. A simple dose calculation is provided to show the relative difference achieved by improving the technical bases

  19. Guide for licensing evaluations using CRAC2: A computer program for calculating reactor accident consequences

    International Nuclear Information System (INIS)

    White, J.E.; Roussin, R.W.; Gilpin, H.

    1988-12-01

    A version of the CRAC2 computer code applicable for use in analyses of consequences and risks of reactor accidents in case work for environmental statements has been implemented for use on the Nuclear Regulatory Commission Data General MV/8000 computer system. Input preparation is facilitated through the use of an interactive computer program which operates on an IBM personal computer. The resulting CRAC2 input deck is transmitted to the MV/8000 by using an error-free file transfer mechanism. To facilitate the use of CRAC2 at NRC, relevant background material on input requirements and model descriptions has been extracted from four reports - ''Calculations of Reactor Accident Consequences,'' Version 2, NUREG/CR-2326 (SAND81-1994) and ''CRAC2 Model Descriptions,'' NUREG/CR-2552 (SAND82-0342), ''CRAC Calculations for Accident Sections of Environmental Statements, '' NUREG/CR-2901 (SAND82-1693), and ''Sensitivity and Uncertainty Studies of the CRAC2 Computer Code,'' NUREG/CR-4038 (ORNL-6114). When this background information is combined with instructions on the input processor, this report provides a self-contained guide for preparing CRAC2 input data with a specific orientation toward applications on the MV/8000. 8 refs., 11 figs., 10 tabs

  20. Health effects models for nuclear power plant accident consequence analysis

    International Nuclear Information System (INIS)

    Abrahamson, S.; Bender, M.A.; Boecker, B.B.; Scott, B.R.

    1993-05-01

    The Nuclear Regulatory Commission (NRC) has sponsored several studies to identify and quantify, through the use of models, the potential health effects of accidental releases of radionuclides from nuclear power plants. The Reactor Safety Study provided the basis for most of the earlier estimates related to these health effects. Subsequent efforts by NRC-supported groups resulted in improved health effects models that were published in the report entitled open-quotes Health Effects Models for Nuclear Power Plant Consequence Analysisclose quotes, NUREG/CR-4214, 1985 and revised further in the 1989 report NUREG/CR-4214, Rev. 1, Part 2. The health effects models presented in the 1989 NUREG/CR-4214 report were developed for exposure to low-linear energy transfer (LET) (beta and gamma) radiation based on the best scientific information available at that time. Since the 1989 report was published, two addenda to that report have been prepared to (1) incorporate other scientific information related to low-LET health effects models and (2) extend the models to consider the possible health consequences of the addition of alpha-emitting radionuclides to the exposure source term. The first addendum report, entitled open-quotes Health Effects Models for Nuclear Power Plant Accident Consequence Analysis, Modifications of Models Resulting from Recent Reports on Health Effects of Ionizing Radiation, Low LET Radiation, Part 2: Scientific Bases for Health Effects Models,close quotes was published in 1991 as NUREG/CR-4214, Rev. 1, Part 2, Addendum 1. This second addendum addresses the possibility that some fraction of the accident source term from an operating nuclear power plant comprises alpha-emitting radionuclides. Consideration of chronic high-LET exposure from alpha radiation as well as acute and chronic exposure to low-LET beta and gamma radiations is a reasonable extension of the health effects model

  1. Developments in modelling the economic impact of Off-site accident consequences

    International Nuclear Information System (INIS)

    Haywood, S.M.; Robinson, C.A.; Faude, D.

    1991-01-01

    Models for assessing the economic consequences of accidental releases of radioactivity have application both in accident consequence codes and in decision aiding computer systems for use in emergency response. Such models may be applied in emergency planning, and studies in connection with the siting, design and licensing of nuclear facilities. Several models for predicting economic impact have been developed, in Europe and the US, and these are reviewed. A new model, called COCO-1 (Cost of Consequences Off-site), has been developed under the CEC MARIA programme and the features of the model are summarised. The costs calculated are a measure of the benefit foregone as a result of the accident, and in addition to tangible monetary costs the model attempts to include costs arising from the effect of the accident on individuals, for instance the disruption caused by the loss of homes. COCO-1 includes the cost of countermeasures, namely evacuation, relocation, sheltering, food restrictions and decontamination, and also the cost of health effects in the exposed population. The primary quantity used in COCO-1 to measure the economic value of land subject to restrictions on usage is Gross Domestic Product (GDP). Examples of default data included in the model are presented, as are the results of an illustrative application. The limitations of COCO-1 are discussed, and areas where further data are needed are identified

  2. Analyses of PWR boron dilution consequences with the Arrotta code

    International Nuclear Information System (INIS)

    Johanson, E.; Cheng, H.W.; Sehgal, B.R.

    1998-03-01

    During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced

  3. Integrated severe accident containment analysis with the CONTAIN computer code

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Williams, D.C.; Rexroth, P.E.; Tills, J.L.

    1985-12-01

    Analysis of physical and radiological conditions iunside the containment building during a severe (core-melt) nuclear reactor accident requires quantitative evaluation of numerous highly disparate yet coupled phenomenologies. These include two-phase thermodynamics and thermal-hydraulics, aerosol physics, fission product phenomena, core-concrete interactions, the formation and combustion of flammable gases, and performance of engineered safety features. In the past, this complexity has meant that a complete containment analysis would require application of suites of separate computer codes each of which would treat only a narrower subset of these phenomena, e.g., a thermal-hydraulics code, an aerosol code, a core-concrete interaction code, etc. In this paper, we describe the development and some recent applications of the CONTAIN code, which offers an integrated treatment of the dominant containment phenomena and the interactions among them. We describe the results of a series of containment phenomenology studies, based upon realistic accident sequence analyses in actual plants. These calculations highlight various phenomenological effects that have potentially important implications for source term and/or containment loading issues, and which are difficult or impossible to treat using a less integrated code suite

  4. Severe accident analysis using MARCH 1.0 code

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1987-09-01

    The description and utilization of the MARCH 1.0 computer code, which aim to analyse physical phenomena associated with core meltdown accidents in PWR type reactors, are presented. The primary system is modeled as a single volume which is partitioned into a gas (steam and hydrogen) region and a water region. March predicts blowdown from the primary system in single phase. Based on results of the probabilistic safety analysis for the Zion and Indian Point Nuclear Power Plants, the S 2 HFX sequence accident for Angra-1 reactor is studied. The S 2 HFX sequence means that the loss of coolant accident occurs through small break in primary system with bot total failures of the reactor safety system and containment in yours recirculation modes, leading the core melt and the containment failure due to overpressurization. The obtained results were considered reasonable if compared with the results obtained for the Zion and Indian Point nuclear power plants. (Author) [pt

  5. State of the art for assessing the off-side economic consequences of nuclear accidents

    International Nuclear Information System (INIS)

    Gallego Diaz, E.

    1996-01-01

    The paper is intended to offer a wide perspective on th methodologies for assessing the off-side economic consequences of nuclear accidents. The element which can contribute to the cost are first reviewed, namely the application of countermeasures against radioactive contamination: population movements, decontamination, food bans; together with the resulting health effects if this is the case. The basic characteristics of the existing models and codes are also presented, including the most recent developments and intercomparisons of results. Some applications of this kind of studies in different fields are outlined. (Author) 17 refs

  6. Modeling the consequences of hypothetical accidents for the Titan II system

    International Nuclear Information System (INIS)

    Greenly, G.D.; Sullivan, T.J.

    1981-11-01

    Calculations have been made with the Atmospheric Release Advisory Capability (ARAC) suite of three-dimensional transport and diffusion codes MATHEW/ADPIC to assess the consequences of severe, hypothetical accident scenarios. One set of calculations develops the integrated dose and surface deposition patterns for a non-nuclear, high explosive detonation and dispersal of material. A second set of calculations depicts the time integrated dose and instantaneous concentration patterns for a substantial, continuous leak of the missile fuel oxidizer converted to nitrogen dioxide (NO 2 ). The areas affected and some of the implications for emergency response management are discussed

  7. Accident consequence calculations for project W-058 safetyanalysis

    Energy Technology Data Exchange (ETDEWEB)

    Van Keuren, J.C.

    1997-06-10

    Accident consequence analyses have been performed for Project W-058, the Replacement Cross Site Transfer System. using the assumption and analysis techniques developed for the Tank Remediation Waste system Basis for Interim Operation. most potential accident involving the FISTS are bounded by the TWRS BIO analysis. However, the spray leak and pool leak scenarios require revised analyses since the RCSTS design utilizes larger diameter pipe and higher pressures than those analyzed in the TWRS BIO. Also the volume of diversion box and vent station are larger than that assumed for the valve pits in the TWRS BIO, which effects results of sprays or spills into the pits. the revised analysis for the spray leak is presented in Section 2, for the above ground spill in Section 3, for the presented in Section 2, for the above ground spill in Section 3, for the subsurface spill forming a pool in Section 4, and for the subsurface pool remaining subsurface in Section 5. The conclusion from these sections are summarized below.

  8. Health effects models for nuclear power plant accident consequence analysis

    International Nuclear Information System (INIS)

    Evans, J.S.; Abrahmson, S.; Bender, M.A.; Boecker, B.B.; Scott, B.R.; Gilbert, E.S.

    1993-10-01

    This report is a revision of NUREG/CR-4214, Rev. 1, Part 1 (1990), Health Effects Models for Nuclear Power Plant Accident Consequence Analysis. This revision has been made to incorporate changes to the Health Effects Models recommended in two addenda to the NUREG/CR-4214, Rev. 1, Part 11, 1989 report. The first of these addenda provided recommended changes to the health effects models for low-LET radiations based on recent reports from UNSCEAR, ICRP and NAS/NRC (BEIR V). The second addendum presented changes needed to incorporate alpha-emitting radionuclides into the accident exposure source term. As in the earlier version of this report, models are provided for early and continuing effects, cancers and thyroid nodules, and genetic effects. Weibull dose-response functions are recommended for evaluating the risks of early and continuing health effects. Three potentially lethal early effects -- the hematopoietic, pulmonary, and gastrointestinal syndromes are considered. Linear and linear-quadratic models are recommended for estimating the risks of seven types of cancer in adults - leukemia, bone, lung, breast, gastrointestinal, thyroid, and ''other''. For most cancers, both incidence and mortality are addressed. Five classes of genetic diseases -- dominant, x-linked, aneuploidy, unbalanced translocations, and multifactorial diseases are also considered. Data are provided that should enable analysts to consider the timing and severity of each type of health risk

  9. Offsite radiological consequence analysis for the bounding aircraft crash accident

    International Nuclear Information System (INIS)

    OBERG, B.D.

    2003-01-01

    The purpose of this calculation note is to quantitatively analyze a bounding aircraft crash accident for comparison to the DOE-STD-3009-94, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', Appendix A, Evaluation Guideline of 25 rem. The potential of aircraft impacting a facility was evaluated using the approach given in DOE-STD-3014-96, ''Accident Analysis for Aircraft Crash into Hazardous Facilities''. The following aircraft crash FR-equencies were determined for the Tank Farms in RPP-11736, ''Assessment Of Aircraft Crash FR-equency For The Hanford Site 200 Area Tank Farms'': (1) The total aircraft crash FR-equency is ''extremely unlikely.'' (2) The general aviation crash FR-equency is ''extremely unlikely.'' (3) The helicopter crash FR-equency is ''beyond extremely unlikely.'' (4) For the Hanford Site 200 Areas, other aircraft type, commercial or military, each above ground facility, and any other type of underground facility is ''beyond extremely unlikely.'' As the potential of aircraft crash into the 200 Area tank farms is more FR-equent than ''beyond extremely unlikely,'' consequence analysis of the aircraft crash is required

  10. Testing of an accident consequence assessment model using field data

    International Nuclear Information System (INIS)

    Homma, Toshimitsu; Matsubara, Takeshi; Tomita, Kenichi

    2007-01-01

    This paper presents the results obtained from the application of an accident consequence assessment model, OSCAAR to the Iput dose reconstruction scenario of BIOMASS and also to the Chernobyl 131 I fallout scenario of EMRAS, both organized by International Atomic Energy Agency. The Iput Scenario deals with 137 Cs contamination of the catchment basin and agricultural area in the Bryansk Region of Russia, which was heavily contaminated after the Chernobyl accident. This exercise was used to test the chronic exposure pathway models in OSCAAR with actual measurements and to identify the most important sources of uncertainty with respect to each part of the assessment. The OSCAAR chronic exposure pathway models had some limitations but the refined model, COLINA almost successfully reconstructed the whole 10-year time course of 137 Cs activity concentrations in most requested types of agricultural products and natural foodstuffs. The Plavsk scenario provides a good opportunity to test not only the food chain transfer model of 131 I but also the method of assessing 131 I thyroid burden. OSCAAR showed in general good capabilities for assessing the important 131 I exposure pathways. (author)

  11. Accident consequence calculations for project W-058 safety analysis

    International Nuclear Information System (INIS)

    Van Keuren, J.C.

    1997-01-01

    Accident consequence analyses have been performed for Project W-058, the Replacement Cross Site Transfer System. using the assumption and analysis techniques developed for the Tank Remediation Waste system Basis for Interim Operation. most potential accident involving the FISTS are bounded by the TWRS BIO analysis. However, the spray leak and pool leak scenarios require revised analyses since the RCSTS design utilizes larger diameter pipe and higher pressures than those analyzed in the TWRS BIO. Also the volume of diversion box and vent station are larger than that assumed for the valve pits in the TWRS BIO, which effects results of sprays or spills into the pits. the revised analysis for the spray leak is presented in Section 2, for the above ground spill in Section 3, for the presented in Section 2, for the above ground spill in Section 3, for the subsurface spill forming a pool in Section 4, and for the subsurface pool remaining subsurface in Section 5. The conclusion from these sections are summarized below

  12. Consequences of the nuclear power plant accident at Chernobyl

    International Nuclear Information System (INIS)

    Ginzburg, H.M.; Reis, E.

    1991-01-01

    The Chernobyl Nuclear Power Plant accident, in the Ukrainian Soviet Socialist Republic (SSR), on April 26, 1986, was the first major nuclear power plant accident that resulted in a large-scale fire and subsequent explosions, immediate and delayed deaths of plant operators and emergency service workers, and the radioactive contamination of a significant land area. The release of radioactive material, over a 10-day period, resulted in millions of Soviets, and other Europeans, being exposed to measurable levels of radioactive fallout. Because of the effects of wind and rain, the radioactive nuclide fallout distribution patterns are not well defined, though they appear to be focused in three contiguous Soviet Republics: the Ukrainian SSR, the Byelorussian SSR, and the Russian Soviet Federated Socialist Republic. Further, because of the many radioactive nuclides (krypton, xenon, cesium, iodine, strontium, plutonium) released by the prolonged fires at Chernobyl, the long-term medical, psychological, social, and economic effects will require careful and prolonged study. Specifically, studies on the medical (leukemia, cancers, thyroid disease) and psychological (reactive depressions, post-traumatic stress disorders, family disorganization) consequences of continued low dose radiation exposure in the affected villages and towns need to be conducted so that a coherent, comprehensive, community-oriented plan may evolve that will not cause those already affected any additional harm and confusion

  13. Consequences of the Chernobyl accident for people and the environment

    International Nuclear Information System (INIS)

    2006-01-01

    This report recalls the accident scenario, discusses the dispersion of the radioactive plume, comments the contamination at the vicinity of the power station, discusses and comments data related to radioactive deposits in Europe and in France, comments available information regarding radioactive fallouts in Belarus, Ukraine and Russia (models have been used to assess radioactive deposits). It addresses the issue of food product contamination in these three countries (impact on farm products, on water streams and on forests), but also in France. It comments the health impacts, more particularly on the people who intervened on the site, but also on people who received medium doses. Thyroid cancer data are discussed for the three mainly concerned countries. Other pathologies and non-cancerous effects are also discussed. The mortality induced by the accident is commented. Effects in France are evoked as well as social and economic consequences in Ukraine, Belarus and Russia. The document provides several links to other documents for further and more detailed information

  14. Towards more realistic assessment of reactor accident consequences

    International Nuclear Information System (INIS)

    Tveten, U.

    1985-07-01

    The purpose of the Nordic project described in the report has been to improve the data base used in accident consequence assessments, and also to improve the assessment models in use in the Nordic countries. The following data related questions have been dealt with: Terrestrial transfer factors, the freshwater pathways, comparison of dynamic and static calculation models for fish, and the shielding effect of buildings. The work on terrestrial transfer factors has resulted in the generation of a Nordic fallout data bank. The following experimental investigations have been performed: Natural decontamination of roofs under summer and winter conditions, deposition in urban areas, and the filter effect of buildings. Various aspects of mitigating actions have also been examined

  15. The Nordic safety program on accident consequence assessment

    International Nuclear Information System (INIS)

    Tveten, U.

    1988-01-01

    One important part of Nordic cooperation is partially funded by the Nordic Council of Ministers, namely the work performed within the Nordic Safety Program (often referred to as the NKA projects). NKA is the Nordic abbreviation of the Nordic Liaison Committee on Atomic Energy. One program area in the present four-year period is concerned with problems related to reactor accident consequence assessment, and contains almost twenty projects covering a wide range of subjects. The author is program coordinator for this program area. The program will be completed in 1989. The program was strongly influenced by Chernobyl, and a number of new projects were included in the program in 1986. Involved in the program are these Nordic institutions: Riso National Laboratory (Denmark). Technical Research Centre of Finland. Finnish Centre for Radiation and Nuclear Safety. Finnish Meteorological Institute. Institute for Energy Technology (Norway). Agricultural University of Norway. Meteorological Institute of Norway. Studsvik Energiteknik AB (Sweden). National Defence Research Laboratory (Sweden)

  16. Social and psychological consequences of the Chernobyl accident in Yugoslavia

    Energy Technology Data Exchange (ETDEWEB)

    Milanovic, S; Pavlovic, S [Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1997-09-01

    A day before the accident in Chernobyl, Yugoslavia was the country with nuclear energy programme, one nuclear power plant and strong affiliation towards nuclear fuel cycle. Public relation programs did not existed. The majority of information were classified and public trust was almost undisturbed. It was almost possible to say that the public attitude was indifferent. A month later everything was quite different. The public has been awaken from sleepy unconscious. The public reaction moved from surprise, interest and hunger for information to chronic suspicion. In years later phobic and radiophonic reaction become common place. The final consequence today is huge magnifying lens of public eye, watching carefully everything connected with radiation, even trivial matters, and thus forming strong pressure to decision makers. 2 refs.

  17. Social and psychological consequences of the Chernobyl accident in Yugoslavia

    International Nuclear Information System (INIS)

    Milanovic, S.; Pavlovic, S.

    1997-01-01

    A day before the accident in Chernobyl, Yugoslavia was the country with nuclear energy programme, one nuclear power plant and strong affiliation towards nuclear fuel cycle. Public relation programs did not existed. The majority of information were classified and public trust was almost undisturbed. It was almost possible to say that the public attitude was indifferent. A month later everything was quite different. The public has been awaken from sleepy unconscious. The public reaction moved from surprise, interest and hunger for information to chronic suspicion. In years later phobic and radiophonic reaction become common place. The final consequence today is huge magnifying lens of public eye, watching carefully everything connected with radiation, even trivial matters, and thus forming strong pressure to decision makers

  18. Test Data for USEPR Severe Accident Code Validation

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe

    2007-05-01

    This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: • Fuel Heatup and Melt Progression • Reactor Coolant System (RCS) Thermal Hydraulics • In-Vessel Molten Pool Formation and Heat Transfer • Fuel/Coolant Interactions during Relocation • Debris Heat Loads to the Vessel • Vessel Failure • Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure • Melt Spreading and Coolability • Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

  19. [Thrombosis and post-thrombotic syndrome as a consequence of an accident].

    Science.gov (United States)

    Wahl, U; Hirsch, T

    2015-10-01

    Phlebothromboses represent alarming complications in accident victims since they can cause fatal pulmonary embolisms. More than half of those affected also develop post-thrombotic syndrome in the course of the illness. In addition to making clinical assessments, the traumatologist should also have fundamental knowledge about diagnostic methods and be familiar with interpreting internal findings. Colour-coded duplex sonography plays a central role in diagnosing thrombosis and in assessing functional limitations. Further information can be gathered from various phlebological procedures. The expert evaluation of the immediate, as well as the long-term consequences of an accident frequently require leg swelling to be classified. It is not uncommon for post-thrombotic syndrome to be diagnosed for the first time during this process. An additional vascular appraisal is often required. An appreciation of social-medical and insurance-related aspects means a high degree of responsibility is placed on the expert.

  20. Integrated computer codes for nuclear power plant severe accident analysis

    International Nuclear Information System (INIS)

    Jordanov, I.; Khristov, Y.

    1995-01-01

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs

  1. Integrated computer codes for nuclear power plant severe accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jordanov, I; Khristov, Y [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1996-12-31

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs.

  2. Development of INCTAC code for analyzing criticality accident phenomena

    International Nuclear Information System (INIS)

    Mitake, Susumu; Hayashi, Yamato; Sakurai, Shungo

    2003-01-01

    Aiming at understanding nuclear transients and thermal- and hydraulic-phenomena of the criticality accident, a code named INCTAC has been newly developed at the Institute of Nuclear Safety. The code is applicable to the analysis of criticality accident transients of aqueous homogenous fuel solution system. Neutronic transient model is composed of equations for the kinetics and for the spatial distributions, which are deduced from the time dependent multi-group transport equations with the quasi steady state assumption. Thermal-hydraulic transient model is composed of a complete set of the mass, momentum and energy equations together with the two-phase flow assumptions. Validation tests of INCTAC were made using the data obtained at TRACY, a transient experiment criticality facility of JAERI. The calculated results with INCTAC showed a very good agreement with the experiment data, except a slight discrepancy of the time when the peak of reactor power was attained. But, the discrepancy was resolved with the use of an adequate model for movement and transfer of the void in the fuel solution mostly generated by radiolysis. With a simulation model for the transport of radioactive materials through ventilation systems to the environment, INCTAC will be used as an overall safety evaluation code of the criticality accident. (author)

  3. Validation and application of the system code ATHLET-CD for BWR severe accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

    2016-10-15

    Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

  4. Process criticality accident likelihoods, consequences and emergency planning

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1992-01-01

    Evaluation of criticality accident risks in the processing of significant quantities of fissile materials is both complex and subjective, largely due to the lack of accident statistics. Thus, complying with national and international standards and regulations which require an evaluation of the net benefit of a criticality accident alarm system, is also subjective. A review of guidance found in the literature on potential accident magnitudes is presented for different material forms and arrangements. Reasoned arguments are also presented concerning accident prevention and accident likelihoods for these material forms and arrangements. (Author)

  5. Consequences of Fukushima 11032011 - Radiological consequences from the nuclear accidents in Fukushima on 11 March 2011

    International Nuclear Information System (INIS)

    2011-12-01

    On 11 March 2011 at 14.46 the strongest earthquake ever recorded in Japan struck the Pacific coast in front of Fukushima. The earthquake and the following tsunami damaged the nuclear power plants in Fukushima Dai-ichi to such an extent that the Japanese government declared the state of catastrophic accident with degree 7 according to the International Nuclear and Radiological Event Scale (INES). At Fukushima Dai-ichi there were 6 boiling water reactors (BWR), a storage pool for spent fuel assemblies and a dry cask storage. 12 km apart at Fukushima Dai-ni there were 4 more BWR. At the moment of the earthquake the reactors 1 to 3 of Fukushima Dai-ichi, as well as the 4 reactors at Fukushima Dai-ni, were at full power, while the reactors 4 to 6 of Fukushima Dai-ichi were shut down for revision. From 12 March 2011 on, fairly large quantities of radioactive materials were released from Fukushima Dai-ichi reactors with meaningful consequences on the population in the near neighbourhood. The irradiation from the radioactivity bearing clouds, the ingestion and inhalation, and the deposit of radioactive materials on the ground threatened the population. The inhabitants of large areas had to be evacuated. Furthermore, radioactive materials contaminated the drinking water, the sea water and finally the plants and animals, i.e. the food chain of the people living there. The Swiss Federal Nuclear Safety Inspectorate (ENSI) continuously proceeded with the evaluation of the situation in Japan and a specialists' team made a detailed analysis of the accident, with emphasis on the human and organisational factors and on the lessons learned from this. The present report describes the present knowledge about the radiological consequences of the accident in Fukushima Dai-ichi on the population in the neighbourhood and on the staff at the power plant, until October 2011. First, the unrolling of the accident and its consequences on the plant site are analysed according to international

  6. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  7. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    International Nuclear Information System (INIS)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav

    2017-01-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  8. Steady-state and accident analyses of PBMR with the computer code SPECTRA

    International Nuclear Information System (INIS)

    Stempniewicz, Marek M.

    2002-01-01

    The SPECTRA code is an accident analysis code developed at NRG. It is designed for thermal-hydraulic analyses of nuclear or conventional power plants. The code is capable of analysing the whole power plant, including reactor vessel, primary system, various control and safety systems, containment and reactor building. The aim of the work presented in this paper was to prepare a preliminary thermal-hydraulic model of PBMR for SPECTRA, and perform steady state and accident analyses. In order to assess SPECTRA capability to model the PBMR reactors, a model of the INCOGEN system has been prepared first. Steady state and accident scenarios were analyzed for INCOGEN configuration. Results were compared to the results obtained earlier with INAS and OCTOPUS/PANTHERMIX. A good agreement was obtained. Results of accident analyses with PBMR model showed qualitatively good results. It is concluded that SPECTRA is a suitable tool for analyzing High Temperature Reactors, such as INCOGEN or for example PBMR (Pebble Bed Modular Reactor). Analyses of INCOGEN and PBMR systems showed that in all analyzed cases the fuel temperatures remained within the acceptable limits. Consequently there is no danger of release of radioactivity to the environment. It may be concluded that those are promising designs for future safe industrial reactors. (author)

  9. A study on the estimation of economic consequence of severe accident

    International Nuclear Information System (INIS)

    Hong, Dae Seok; Lee, Kun Jai; Jeong, Jong Tae

    1996-01-01

    A model to estimate economic consequence of severe accident provides some measure of the impact on the accident and enables to know the different effects of the accident described as same terms of cost and combined as necessary. Techniques to assess the consequences of accidents in terms of cost have many applications, for instance in examining countermeasure options, as part of either emergency planning or decision making after an accident. In this study, a model to estimate the accident economic consequence is developed appropriate to our country focused on PWR accident costs from a societal viewpoint. Societal costs are estimated by accounting for losses that directly affect the plant licensee, the public, the nuclear industry, or the electric utility industry after PWR accident

  10. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia; Martins, Marcelo, E-mail: ayabe@ipen.br, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LABRISCO/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco

    2017-07-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  11. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    International Nuclear Information System (INIS)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R.; Giovedi, Claudia; Martins, Marcelo

    2017-01-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  12. The modelling of off-site economic consequences of nuclear accidents

    International Nuclear Information System (INIS)

    Alonso, A.; Gallego, E.; Martin, J.E.

    1991-01-01

    The paper presents a computer model for the probabilistic assessment of the off-site economic risk derived from nuclear accidents. The model is called MECA (Model for Economic Consequence Assessment) and takes into consideration the direct costs caused, following an accident, by the different countermeasures adopted to prevent both the early and chronic exposure of the population to the radionuclides released, as well as the direct costs derived from health damage to the affected population. The model uses site-specific data that are organized in a socio-economic data base; detailed distributions of population, livestock census, agricultural production and farmland use, as well as of employment, salaries, and added value for different economic sectors are included. This data base has been completed for Spain, based on available official statistics. The new code, coupled to a general ACA code, provides capability to complete probabilistic risk assessments from the point of view of the off-site economic consequences, and also to perform cost-effectiveness analysis of the different countermeasures in the field of emergency preparedness

  13. Severe accident tests and development of domestic severe accident system codes

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

  14. Severe accident tests and development of domestic severe accident system codes

    International Nuclear Information System (INIS)

    2013-01-01

    According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

  15. Cesium-137: psychological and social consequences of the Goiania's accident

    International Nuclear Information System (INIS)

    Helou, Suzana; Costa Neto, Sebastiao Benicio da

    1995-01-01

    The book care for radioactive accident occurred in 1987 in Goiania - brazilian city. The accident had origin by the hospitable equipment incorrect handling which contained a stainless steel capsule, in which interior there was cesium-137 chloride. The main boarded aspects are: psychological and social aspects verified after the accident; psychological and social analysis of population of Goiania three years after the accident; essay on the pertinence of Luscher's abbreviate test in psychological evaluation of the radioactive accident victims of Goiania; and psychological and mobile evaluation of intra-uterus children exposed to the radiation with cesium-137

  16. Process criticality accident likelihoods, consequences, and emergency planning

    Energy Technology Data Exchange (ETDEWEB)

    McLaughlin, T.P.

    1991-01-01

    Evaluation of criticality accident risks in the processing of significant quantities of fissile materials is both complex and subjective, largely due to the lack of accident statistics. Thus, complying with standards such as ISO 7753 which mandates that the need for an alarm system be evaluated, is also subjective. A review of guidance found in the literature on potential accident magnitudes is presented for different material forms and arrangements. Reasoned arguments are also presented concerning accident prevention and accident likelihoods for these material forms and arrangements. 13 refs., 1 fig., 1 tab.

  17. The program system UFOMOD for assessing the consequences of nuclear accidents

    International Nuclear Information System (INIS)

    Ehrhardt, J.; Burkart, K.; Hasemann, I.; Matzerath, C.; Panitz, H.J.; Steinhauer, C.

    1988-10-01

    The programm system UFOMOD is a completely new accident consequence assessment (ACA) code. Its structure and modelling is based on the experience gained from applications of the old UFOMOD code during and after the German Risk Study - Phase A, the results of scientific investigations performed within the ongoing Phase B and the CEC-project MARIA, and the requirements resulting from the extended use of ACAs to help in decision-making. One of the most important improvements is the introduction of different trajecotry models for describing atmospheric dispersion in the near range and at larger distances. Emergency actions and countermeasures modelling takes into account recommendations of international commissions. The dosimetric models contain completely new age-, sex- and time-dependent data of dose-conversion factors for external and internal radiation; the ingestion pathway is modelled to consider seasonal dependencies. New dose-risk-relationships for stochastic and non-stochastic health effects are implemented; a special algorithm developed for ACA codes allows individual and collective leukemia and cancer risks to be presented as a function of time after the accident. According to the modular structure of the new program system UFOMOD, an easy access to parameter values and the results of the various submodels exists what facilitates sensitivity and uncertainty analyses. (orig.) [de

  18. Estimation of the Radiological Consequences of Fukushima Dai-ichi Nuclear Power Plant Accident using MACCS2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sora; Min, Byung-Il; Park, Kihyun; Yang, Byung-Mo; Suh, Kyung-suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Three of them have undergone fuel melting and hydrogen explosions. A significant amount of radioactive material was released into the atmosphere from FDNPP and dispersed all over the world. In this study, we assessed the offsite consequences of Fukushima disaster in the region within a 30-km radius of FDNPP using the MELCOR Accident Consequence Code Systems 2(MACCS2) code, which is the Nuclear Regulatory Commission's (NRC's) code. The reflection of the realistic regional characteristics, such as long-term meteorological data, site- and population-specific data, and radiation safety regulatory, is essential to accurately analyze the off-site consequences. The assessment that reflects regional characteristics would contribute to identify main causes of exposure doses and to find the effective countermeasures for minimizing the accidental off-site consequences.

  19. 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

  20. Development of a system of computer codes for severe accident analyses and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1991-12-15

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy.

  1. Development of a system of computer codes for severe accident analyses and its applications

    International Nuclear Information System (INIS)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan

    1991-12-01

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy

  2. The assessment of environmental consequences of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Beattie, J.R.

    1981-01-01

    Thorough measures are taken throughout all stages of design, construction and operation of nuclear power reactors, and therefore no accident producing any significant environmental impact is likely to occur. Nevertheless as a precaution, such accidents have been the subject of intensive scientific predictive studies. After a historical review of theoretical papers on reactor accidents and their imagined environmental impacts and of those accidents that have indeed occurred, this paper gives an outline of fission products or other radioactive substances that may or may not be released by an accident, and of their possible effects after dispersion in the atmosphere. This general introduction is followed by sections describing what are sometimes called 'design basis accidents' for four of the main reactor types (magnox, AGR, PWR and CDFR), the precautions against these accidents and the probable degree of environmental impact likely. The paper concludes with a reference to those very low probability accidents which might have more serious environmental impacts, and proceeds from there to show that both the individual and community risks from such accidents are numerically moderate compared to other risks apparently accepted by society. A brief reflection on the relevance of numerical values and perceived risk concludes the paper. (author)

  3. Preliminary Analysis of a Loss of Condenser Vacuum Accident Using the MARS-KS Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jieun Kim; Bang, Young Seok; Oh, Deog Yeon; Kim, Kap; Woo, Sweng-Wong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    In accordance with revision of NUREG-0800 of USNRC, the area of review for loss of condenser vacuum(LOCV) accident has been expanded to analyze both peak pressures of primary and secondary system separately. Currently, the analysis of LOCV accident, which is caused by malfunction of condenser, has been focused to fuel cladding integrity and peak pressure in the primary system. In this paper, accident analysis for LOCV using MARS-KS code were conducted to support the licensing review on transient behavior of secondary system pressure of APR1400 plant. The accident analysis for the loss of condenser vacuum (LOCV) of APR1400 was conducted with the MARS-KS code to support the review on the pressure behavior of primary and secondary system. Total four cases which have different combination of availability of offsite power and the pressurizer spray are considered. The preliminary analysis results shows that the initial conditions or assumptions which concludes the severe consequence are different for each viewpoint, and in some cases, it could be confront with each viewpoint. Therefore, with regard to the each acceptance criteria, figuring out and sensitivity analysis of the initial conditions and assumptions for system pressure would be necessary.

  4. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment. Volume 3, Appendices C, D, E, F, and G

    Energy Technology Data Exchange (ETDEWEB)

    Harper, F.T.; Young, M.L.; Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States)] [and others

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the third of a three-volume document describing the project and contains descriptions of the probability assessment principles; the expert identification and selection process; the weighting methods used; the inverse modeling methods; case structures; and summaries of the consequence codes.

  5. Probabilistic accident consequence uncertainty analysis: Dispersion and deposition uncertainty assessment. Volume 3, Appendices C, D, E, F, and G

    International Nuclear Information System (INIS)

    Harper, F.T.; Young, M.L.; Miller, L.A.

    1995-01-01

    The development of two new probabilistic accident consequence codes, MACCS and COSYMA, completed in 1990, estimate the risks presented by nuclear installations based on postulated frequencies and magnitudes of potential accidents. In 1991, the US Nuclear Regulatory Commission (NRC) and the Commission of the European Communities (CEC) began a joint uncertainty analysis of the two codes. The objective was to develop credible and traceable uncertainty distributions for the input variables of the codes. Expert elicitation, developed independently, was identified as the best technology available for developing a library of uncertainty distributions for the selected consequence parameters. The study was formulated jointly and was limited to the current code models and to physical quantities that could be measured in experiments. To validate the distributions generated for the wet deposition input variables, samples were taken from these distributions and propagated through the wet deposition code model along with the Gaussian plume model (GPM) implemented in the MACCS and COSYMA codes. Resulting distributions closely replicated the aggregated elicited wet deposition distributions. Project teams from the NRC and CEC cooperated successfully to develop and implement a unified process for the elaboration of uncertainty distributions on consequence code input parameters. Formal expert judgment elicitation proved valuable for synthesizing the best available information. Distributions on measurable atmospheric dispersion and deposition parameters were successfully elicited from experts involved in the many phenomenological areas of consequence analysis. This volume is the third of a three-volume document describing the project and contains descriptions of the probability assessment principles; the expert identification and selection process; the weighting methods used; the inverse modeling methods; case structures; and summaries of the consequence codes

  6. Source term and radiological consequence evaluation for nuclear accidents using a 'hand type' methodology

    International Nuclear Information System (INIS)

    Margeanu, Sorin; Tatiana, Angelescu

    2005-01-01

    In the last decades, hand type calculations have been replaced by computerized solutions, which are much more accurate, but the preparation of an input to run the code can be a time consuming process and can require a laborious work. This is why, a place for hand calculation based on nomograms still exist in some areas. An example is emergency response to an accidental release of radioactive contaminants when the health of persons close to the accident site might be at risk. In this case, results from computerized accident consequences assessment models may be delayed due to the equipment malfunction or the time required developing minimal input files and performing the calculations (typically more than five minutes). A simple nomogram (developed using computerized dispersion model calculations) can provide dispersion and dose estimates within a minute. The paper presents the methodology used for these 'hand type' calculation and the nomograms, figures and tables used to evaluate the dose to an individual close to the release point. In order to illustrate the use of methodology, a hypothetical severe accident scenario involving 14-MW INR-TRIGA research reactor was considered. (authors)

  7. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Brown, J. [National Radiological Protection Board (United Kingdom); Goossens, L.H.J.; Kraan, B.C.P. [Delft Univ. of Technology (Netherlands)] [and others

    1997-06-01

    This volume is the second of a two-volume document that summarizes a joint project by the US Nuclear Regulatory and the Commission of European Communities to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This two-volume report, which examines mechanisms and uncertainties of transfer through the food chain, is the first in a series of five such reports. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain transfer that affect calculations of offsite radiological consequences. Seven of the experts reported on transfer into the food chain through soil and plants, nine reported on transfer via food products from animals, and two reported on both. The expert judgment elicitation procedure and its outcomes are described in these volumes. This volume contains seven appendices. Appendix A presents a brief discussion of the MAACS and COSYMA model codes. Appendix B is the structure document and elicitation questionnaire for the expert panel on soils and plants. Appendix C presents the rationales and responses of each of the members of the soils and plants expert panel. Appendix D is the structure document and elicitation questionnaire for the expert panel on animal transfer. The rationales and responses of each of the experts on animal transfer are given in Appendix E. Brief biographies of the food chain expert panel members are provided in Appendix F. Aggregated results of expert responses are presented in graph format in Appendix G.

  8. Probabilistic accident consequence uncertainty analysis: Food chain uncertainty assessment. Volume 2: Appendices

    International Nuclear Information System (INIS)

    Brown, J.; Goossens, L.H.J.; Kraan, B.C.P.

    1997-06-01

    This volume is the second of a two-volume document that summarizes a joint project by the US Nuclear Regulatory and the Commission of European Communities to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and magnitudes of potential accidents. This two-volume report, which examines mechanisms and uncertainties of transfer through the food chain, is the first in a series of five such reports. A panel of sixteen experts was formed to compile credible and traceable uncertainty distributions for food chain transfer that affect calculations of offsite radiological consequences. Seven of the experts reported on transfer into the food chain through soil and plants, nine reported on transfer via food products from animals, and two reported on both. The expert judgment elicitation procedure and its outcomes are described in these volumes. This volume contains seven appendices. Appendix A presents a brief discussion of the MAACS and COSYMA model codes. Appendix B is the structure document and elicitation questionnaire for the expert panel on soils and plants. Appendix C presents the rationales and responses of each of the members of the soils and plants expert panel. Appendix D is the structure document and elicitation questionnaire for the expert panel on animal transfer. The rationales and responses of each of the experts on animal transfer are given in Appendix E. Brief biographies of the food chain expert panel members are provided in Appendix F. Aggregated results of expert responses are presented in graph format in Appendix G

  9. 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

  10. Chernobyl accident: Causes, consequences and problems of radiation measurements

    International Nuclear Information System (INIS)

    Kortov, V.; Ustyantsev, Yu.

    2013-01-01

    General description of Chernobyl accident is given in the review. The accident causes are briefly described. Special attention is paid to radiation situation after the accident and radiation measurements problems. Some data on Chernobyl disaster are compared with the corresponding data on Fukushima accident. It is noted that Chernobyl and Fukushima lessons should be taken into account while developing further measures on raising nuclear industry safety. -- Highlights: ► The short comparative analysis of accidents at Chernobyl and Fukushima is given. ► We note the great effect of β-radiation on the radiation situation at Chernobyl. ► We discuss the problems of radiation measurements under these conditions. ► The impact of shelter on the radiation situation near Chernobyl NPS is described

  11. A comparison of the consequences of the design basis accident of the Greek Research Reactor with those of a serious realistic accident

    International Nuclear Information System (INIS)

    Kollas, J.G.; Anoussis, J.N.

    1985-12-01

    An analysis of the radiological consequences of the design basis and the coolant flow blockage accidents of the Greek Research Reactor is presented. The results indicate that the consequences of the coolant flow blockage accident are practically trivial being 1-2 orders of magnitude lower than the corresponding consequences of the design basis accident. (author)

  12. A dynamic food-chain model and program for predicting the radiological consequences of nuclear accident

    International Nuclear Information System (INIS)

    Hu Erbang; Gao Zhanrong; Zhang Heyuan; Wei Weiqiang

    1996-12-01

    A dynamic food-chain model and program, DYFOM-95, for predicting the radiological consequences of nuclear accident has been developed, which is not only suitable to the West food-chain but also to Chinese food chain. The following processes, caused by accident release which will make an impact on radionuclide concentration in the edible parts of vegetable are considered: dry and wet deposition interception and initial retention, translocation, percolation, root uptake and tillage. Activity intake rate of animals, effects of processing and activity intake of human through ingestion pathway are also considered in calculations. The effects of leaf area index LAI of vegetable are considered in dry deposition model. A method for calculating the contribution of rain with different period and different intensity to total wet deposition is established. The program contains 1 main code and 5 sub-codes to calculate dry and wet deposition on surface of vegetable and soil, translocation of nuclides in vegetable, nuclide concentration in the edible parts of vegetable and in animal products and activity intake of human and so on. (24 refs., 9 figs., 11 tabs.)

  13. The consequences from liquid pathways after a reactor meltdown accident

    Energy Technology Data Exchange (ETDEWEB)

    Niemczyk, S J; Adams, K G; Murfin, W B; Ritchie, L T; Eppel, E W; Johnson, J D

    1981-06-01

    The potential radiological impact of a core-melt accident on the human population has been investigated. In particular, the radiation dose received from radioactivity which could reach the population via liquid pathways has been considered. Radioactivity could be released directly to the hydrosphere after a core-melt accident as a result of melt-through of the containment basemat followed by any of three processes: (1) leaching of the melt debris; 2 escape of sumpwater through the hole formed by the melt (or from passage out of the containment by an alternate route); and 3) depressurization of the containment atmosphere through the melt hole. The three types of releases would differ primarily in their rates, their magnitudes and their radioactive compositions. Both the containment atmosphere and the sumpwater releases would occur relatively rapidly. However, most of the radionuclides present in these two releases in substantial quantities would be expected to be rather short-lived. Therefore, such releases could have a significant impact at a specific site only if the travel times of the important radionuclides to the human population were small. In contrast, leaching of radionuclides from the melt debris would be expected to occur relatively slowly. Most of the long-lived isotopes would be expected to be found primarily in the melt debris. Consequently, even though this release occurred relatively slowly, the impact could still be significant. In contrast to the situation for releases to the atmosphere, accidents corresponding to the most probable RSS (Reactor Safety Study) meltdown categories would result in the largest releases to the hydrosphere. Furthermore, substantial amounts of radioactivity would generally be expected to be released to the hydrosphere during any meltdown accident involving complete melt-through of the containment basemat. On the basis of subsurface hydrologies alone, sites range from those that essentially preclude any impacts to the human

  14. The consequences from liquid pathways after a reactor meltdown accident

    International Nuclear Information System (INIS)

    Niemczyk, S.J.; Adams, K.G.; Murfin, W.B.; Ritchie, L.T.; Eppel, E.W.; Johnson, J.D.

    1981-06-01

    The potential radiological impact of a core-melt accident on the human population has been investigated. In particular, the radiation dose received from radioactivity which could reach the population via liquid pathways has been considered. Radioactivity could be released directly to the hydrosphere after a core-melt accident as a result of melt-through of the containment basemat followed by any of three processes: (1) leaching of the melt debris; 2 escape of sumpwater through the hole formed by the melt (or from passage out of the containment by an alternate route); and 3) depressurization of the containment atmosphere through the melt hole. The three types of releases would differ primarily in their rates, their magnitudes and their radioactive compositions. Both the containment atmosphere and the sumpwater releases would occur relatively rapidly. However, most of the radionuclides present in these two releases in substantial quantities would be expected to be rather short-lived. Therefore, such releases could have a significant impact at a specific site only if the travel times of the important radionuclides to the human population were small. In contrast, leaching of radionuclides from the melt debris would be expected to occur relatively slowly. Most of the long-lived isotopes would be expected to be found primarily in the melt debris. Consequently, even though this release occurred relatively slowly, the impact could still be significant. In contrast to the situation for releases to the atmosphere, accidents corresponding to the most probable RSS (Reactor Safety Study) meltdown categories would result in the largest releases to the hydrosphere. Furthermore, substantial amounts of radioactivity would generally be expected to be released to the hydrosphere during any meltdown accident involving complete melt-through of the containment basemat. On the basis of subsurface hydrologies alone, sites range from those that essentially preclude any impacts to the human

  15. A computer code for analysis of severe accidents in LWRs

    International Nuclear Information System (INIS)

    2001-01-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  16. A computer code for analysis of severe accidents in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  17. A computer code for analysis of severe accidents in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  18. Prevention of the causes and consequences of a criticality accident - measures adopted in France

    International Nuclear Information System (INIS)

    Fruchard, Y.; Lavie, J.M.

    1966-01-01

    The question of safety in regard to criticality accident risks has two aspects: prevention of the cause and limitation of the consequences. These two aspects are closely connected. The effort devoted to prevention of the causes depends on the seriousness of the possible human psychologic and economic consequences of the accident. The criticality accidents which have occurred in the nuclear industry, though few in number, do reveal the imperfect nature of the techniques adopted to prevent the causes, and also constitute the only available realistic basis for evaluating the consequences and developing measures to limit them. The authors give a analysis of the known causes and consequences of past criticality accidents and on this basis make a number of comments concerning: the validity of traditional safety criteria, the probability of accidents for different types of operations, characteristic accidents which can serve as models, and the extent of possible radiological consequences. The measures adopted in France to limit the consequences of a possible criticality accident under the headings: location, design and lay-out of the installations, accident detection, and dosimetry for the exposed personnel, are briefly described after a short account of the criteria used in deciding on them. (author) [fr

  19. On the sequence and consequences of the Chernobyl reactor accident

    Energy Technology Data Exchange (ETDEWEB)

    Hennies, H H

    1986-01-01

    A serious reactor accident occurred on April 26, 1986 at Chernobyl near Kiev (Soviet Union) where, after melting of the core, there was a considerable release of radioactivity to the environment and to the atmosphere. The radioactivity release caused irradiation of the operating staff, which led to 24 deaths by June 1986. Hardly anything is known about the irradiation of the environment of the reactor plant, but the population within a radius of 30 km was evacuated. The radioactivity released into the atmosphere spread all over Europe, and Germany was affected a few days after the accident. The article gives a short description of the plant which suffered the accident, one tries to describe the course of the accident and to discuss the applicability to German plants.

  20. Review of severe accidents and the results of accident consequence assessment in different energy systems (Contract research)

    International Nuclear Information System (INIS)

    Matsuki, Yoshio; Muramatsu, Ken

    2008-05-01

    The cases of severe accidents and the consequence assessments in different energy systems, Coal, Oil, Gas, Hydro and Nuclear, were collected, and then they were further analyzed. In this report, the information on the accidents in various energy systems were collected from the sources of the Paul Scherrer Institute (hereinafter, 'PSI') and the International Atomic Energy Agency (hereinafter, 'IAEA'). The information on the severe accidents of nuclear power plants were collected from the report of the US Presidential Commission on Catastrophic Nuclear Accidents and several relevant reports issued in the countries of the European Union, together with the reports of the PSI and the IAEA. To analyze the collected information, several parameters, which are numbers of fatalities, injuries, evacuees and the costs of the damages, were chosen to characterize those accidents in different energy systems. And then, upon the comparison of these characteristics of different accidents, the impacts of the accidents in nuclear and other energy systems were compared. Upon the results of the analysis, it is pointed out that the cost caused by the Chernobyl Accident, the severe accident in nuclear energy, tends to be higher than in the other energy systems. On the other hand, from the aspects of fatalities and injuries, it is not confirmed that the damages of the Chernobyl Accident are larger than in the other energy systems. However, it is also recognized, as the specific characteristics of the severe nuclear accident, that the impacts of the accident spread in a wider area, and stay for a longer period, in comparison with the ones in the other energy systems. (author)

  1. Development and assessment of ASTEC code for severe accident simulation

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Pignet, S.; Seropian, C.; Montanelli, T.; Giordano, P.; Jacq, F.; Schwinges, B.

    2005-01-01

    Full text of publication follows: The ASTEC integral code, jointly developed by IRSN and GRS since several years for evaluation of source term during a severe accident (SA) in a Light Water Reactor, will play a central role in the SARNET network of excellence of the 6. Framework Programme (FwP) of the European Commission which started in spring 2004. It should become the reference European SA integral code in the next years. The version V1.1, released in June 2004, allows to model most of the main physical phenomena (except steam explosion) near or at the state of the art. In order to allow to study a great number of scenarios, a compromise must be found between precision of results and calculation time: one day of accident time usually takes less than one day of real time to be simulated on a PC computer. Important efforts are being made on validation by covering more than 30 reference experiments, often International Standard Problems from OECD (CORA, LOFT, PACTEL, BETA, VANAM, ACE-RTF, Phebus.FPT1...). The code is also used for the detailed interpretation of all the integral Phebus.FP experiments. Eighteen European partners performed a first independent evaluation of the code capabilities in 2000-03 within the frame of the EVITA 5. FwP project on one hand by comparison to experiments and on another hand by benchmarking with MAAP4 and MELCOR integral codes on plant applications on PWR and VVER. Their main conclusions were the needs of improvement of code robustness (especially the 2 new modules CESAR and DIVA simulating respectively circuit thermal hydraulics and core degradation) and of post-processing tools. Some improvements have already been achieved in the latest version V 1.1 on these two aspects. A new module MEDICIS devoted to Molten Core Concrete Interaction (MCCI) is implemented in this version, with a tight coupling to the containment thermal hydraulics module CPA. The paper presents a detailed analysis of a TMLB sequence on a French 900 MWe PWR, from

  2. Radioecological and dosimetric consequences of Chernobyl accident in France; Consequences radioecologiques et dosimetriques de l`accident de Tchernobyl en France

    Energy Technology Data Exchange (ETDEWEB)

    Renaud, Ph; Beaugelin, K; Maubert, H; Ledenvic, Ph

    1998-12-31

    After ten years and the taking in account of numerous data, it can be affirmed that the dosimetric consequences of Chernobyl accident will have been limited in France. for the period 1986-2046, the individual middle efficient dose commitment, for the area the most reached by depositing is inferior to 1500 {mu}Sv, that represents about 1% of middle natural exposure in the same time. but mountains and forests can have more important surface activities than in plain. Everywhere else, it can be considered that the effects of Chernobyl accident are disappearing. the levels of cesium 137 are now often inferior to what they were before the accident. (N.C.)

  3. Radioecological and dosimetric consequences of Chernobyl accident in France; Consequences radioecologiques et dosimetriques de l`accident de Tchernobyl en France

    Energy Technology Data Exchange (ETDEWEB)

    Renaud, Ph.; Beaugelin, K.; Maubert, H.; Ledenvic, Ph

    1997-12-31

    After ten years and the taking in account of numerous data, it can be affirmed that the dosimetric consequences of Chernobyl accident will have been limited in France. for the period 1986-2046, the individual middle efficient dose commitment, for the area the most reached by depositing is inferior to 1500 {mu}Sv, that represents about 1% of middle natural exposure in the same time. but mountains and forests can have more important surface activities than in plain. Everywhere else, it can be considered that the effects of Chernobyl accident are disappearing. the levels of cesium 137 are now often inferior to what they were before the accident. (N.C.)

  4. The Fukushima accident: radiological consequences and first lessons. Proceedings; L'accident de Fukushima: consequences radiologiques et premiers enseignements. Recueil des presentations

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-02-15

    This document brings together the available presentations given at the conference organised by the French society of radiation protection about the Fukushima accident, its radiological consequences and the first lessons learnt. Sixteen presentations (slides) are compiled in this document and deal with: 1 - Accident progress and first actions (Thierry Charles, IRSN); 2 - Conditions and health monitoring of the Japanese intervention teams (Bernard Le Guen, EDF); 3 - The Intra Group action after the Fukushima accident (Michel Chevallier, Groupe Intra; Frederic Mariotte, CEA); 4 - Processing of effluents (Georges Pagis, Areva); 5 - Fukushima accident: impact on the terrestrial environment in Japan (Didier Champion, IRSN); 6 - Consequences of the Fukushima accident on the marine environment (Dominique Boust, IRSN); 7 - Territories decontamination perspectives (Pierre Chagvardieff, CEA); 8 - Actions undertaken by Japanese authorities (Florence Gallay, ASN); 9 - Japanese population monitoring and health stakes (Philippe Pirard, InVS); 10 - Citizen oversight actions implemented in Japan (David Boilley, ACRO); 11 - Implementation of ICRP's (International Commission on Radiological Protection) recommendations by Japanese authorities: first analysis (Jacques Lochard, CIPR); 12 - Control of Japan imported food stuff (David Brouque, DGAL); 13 - Questions asked by populations in France and in Germany (Florence-Nathalie Sentuc, GRS; Pascale Monti, IRSN); 14 - Labour law applicable to French workers working abroad (Thierry Lahaye, DGT); 15 - Protection of French workers working in Japan, Areva's experience (Patrick Devin, Areva); 16 - Fukushima accident experience feedback and post-accident nuclear doctrine (Jean-Luc Godet, ASN)

  5. Genetic consequences of the Chernobyl accident for Belarus republic

    International Nuclear Information System (INIS)

    Lazjuk, G.; Nikolaev, D.; Novikova, I.; Satow, Yukio

    1998-01-01

    various uncertainties. Only direct methods, which count the final effect, with all their drawbacks, can provide accurate information on genetic losses. We have estimated possible genetic consequences for the residents of Belarus Republic due to the Chernobyl accident by studying malformations found in legal medical abortuses and by counting congenital anomalies in fetuses and newborns. (J.P.N.)

  6. Genetic consequences of the Chernobyl accident for Belarus republic

    Energy Technology Data Exchange (ETDEWEB)

    Lazjuk, G.; Nikolaev, D.; Novikova, I. [Belarus Institute for Hereditary Diseases, Minsk (Belarus); Satow, Yukio

    1998-03-01

    various uncertainties. Only direct methods, which count the final effect, with all their drawbacks, can provide accurate information on genetic losses. We have estimated possible genetic consequences for the residents of Belarus Republic due to the Chernobyl accident by studying malformations found in legal medical abortuses and by counting congenital anomalies in fetuses and newborns. (J.P.N.)

  7. Uncertainty analysis with a view towards applications in accident consequence assessments

    International Nuclear Information System (INIS)

    Fischer, F.; Erhardt, J.

    1985-09-01

    Since the publication of the US-Reactor Safety Study WASH-1400 there has been an increasing interest to develop and apply methods which allow to quantify the uncertainty inherent in probabilistic risk assessments (PRAs) and accident consequence assessments (ACAs) for installations of the nuclear fuel cycle. Research and development in this area is forced by the fact that PRA and ACA are more and more used for comparative, decisive and fact finding studies initiated by industry and regulatory commissions. This report summarizes and reviews some of the main methods and gives some hints to do sensitivity and uncertainty analyses. Some first investigations aiming at the application of the method mentioned above to a submodel of the ACA-code UFOMOD (KfK) are presented. Sensitivity analyses and some uncertainty studies an important submodel of UFOMOD are carried out to identify the relevant parameters for subsequent uncertainty calculations. (orig./HP) [de

  8. Accidental beam loss in superconducting accelerators: Simulations, consequences of accidents and protective measures

    International Nuclear Information System (INIS)

    Drozhdin, A.; Mokhov, N.; Parker, B.

    1994-02-01

    The consequences of an accidental beam loss in superconducting accelerators and colliders of the next generation range from the mundane to rather dramatic, i.e., from superconducting magnet quench, to overheating of critical components, to a total destruction of some units via explosion. Specific measures are required to minimize and eliminate such events as much as practical. In this paper we study such accidents taking the Superconducting Supercollider complex as an example. Particle tracking, beam loss and energy deposition calculations were done using the realistic machine simulation with the Monte-Carlo codes MARS 12 and STRUCT. Protective measures for minimizing the damaging effects of prefire and misfire of injection and extraction kicker magnets are proposed here

  9. Improved atmospheric dispersion modelling in the new program system UFOMOD for accident consequence assessments

    International Nuclear Information System (INIS)

    Panitz, H.J.

    1988-01-01

    An essential aim of the improvements of the new program system UFOMOD for Accident Consequence Assessments (ACAs) was to substitute the straightline Gaussian plume model conventionally used in ACA models by more realistic atmospheric dispersion models. To identify improved models which can be applied in ACA codes and to quantify the implications of different concepts of dispersion modelling on the results of an ACA, probabilistic comparative calculations with different atmospheric dispersion models have been carried out. The study showed that there are trajectory models available which can be applied in ACAs and that these trajectory models provide more realistic results of ACAs than straight-line Gaussian models. This led to a completly novel concept of atmospheric dispersion modelling which distinguish between two different distance ranges of validity: the near range ( 50 km). The two ranges are assigned to respective trajectory models

  10. Investigation of alpha experiment by severe accident analysis code SAMPSON

    International Nuclear Information System (INIS)

    Baglietto, Emilio; Ninokata, Hisashi; Naitoh, Masanori

    2006-01-01

    The severe accident analysis code SAMPSON is adopted in this work to evaluate its capability of reproducing the complex gap cooling phenomenon. The ALPHA experiment is adopted for validation, where molten aluminum oxide (Al 2 O 3 ) produced by a thermite reaction is poured into a water filled hemispherical vessel at the ambient pressure of approximately 1.3 MPa. The spreading and cooling of the debris that has relocated into the pressure vessel lower plenum are simulated, including the analysis of the RPV failure. The model included in the core to mimic the water penetration inside the gap is evaluated and improvements are proposed. The importance of the introduction of some mechanistic approach to describe the gap formation and evolution is underlined, where the results show its necessity in order to correctly reproduce the experimental trends. (author)

  11. Model to predict radiological consequences of transportation accidents involving dispersal of radioactive material in urban areas

    International Nuclear Information System (INIS)

    Taylor, J.M.; Daniel, S.L.

    1978-01-01

    The analysis of accidental releases of radioactive material which may result from transportation accidents in high-density urban areas is influenced by several urban characteristics which make computer simulation the calculational method of choice. These urban features fall into four categories. Each of these categories contains time- and location-dependent parameters which must be coupled to the actual time and location of the release in the calculation of the anticipated radiological consequences. Due to the large number of dependent parameters a computer model, METRAN, has been developed to quantify these radiological consequences. Rather than attempt to describe an urban area as a single entity, a specific urban area is subdivided into a set of cells of fixed size to permit more detailed characterization. Initially, the study area is subdivided into a set of 2-dimensional cells. A uniform set of time-dependent physical characteristics which describe the land use, population distribution, traffic density, etc., within that cell are then computed from various data sources. The METRAN code incorporates several details of urban areas. A principal limitation of the analysis is the limited availability of accurate information to use as input data. Although the code was originally developed to analyze dispersal of radioactive material, it is currently being evaluated for use in analyzing the effects of dispersal of other hazardous materials in both urban and rural areas

  12. Aspects of risk analysis application to estimation of nuclear accidents and tests consequences and intervention management

    International Nuclear Information System (INIS)

    Demin, V.F.; Hedemann-Jensen, P.; Rolevich, I.V.; Schneider, T.S.; Sobolev, B.G.

    1996-01-01

    For assessment of accident consequences and a post-accident management a risk analysis methodology and data bank (BARD) with allowance for radiation and non-radiation risk causes should be developed and used. Aspects of these needs and developments are considered. Some illustrative results of health risk estimation made with BARD for the Bryansk region territory with relatively high radioactive contamination from the Chernobyl accident are presented

  13. The consequences of the Chernobyl nuclear accident in Greece - Report No. 2

    International Nuclear Information System (INIS)

    1986-12-01

    In this report a realistic estimate of the radioactive fallout on Greece from the Chernobyl nuclear accident is described. The measurements performed on environmental samples and samples of the food chain, as well as some realistic estimations for the population doses and the expected consequences of the accident are presented. The analysis has shown that the radiological impact of the accident in Greece can be considered minor. (J.K.)

  14. Evaluation of radiological dispersion/consequence codes supporting DOE nuclear facility SARs

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Paik, I.K.; Chung, D.Y.

    1996-01-01

    Since the early 1990s, the authorization basis documentation of many U.S. Department of Energy (DOE) nuclear facilities has been upgraded to comply with DOE orders and standards. In this process, many safety analyses have been revised. Unfortunately, there has been nonuniform application of software, and the most appropriate computer and engineering methodologies often are not applied. A DOE Accident Phenomenology and Consequence (APAC) Methodology Evaluation Program was originated at the request of DOE Defense Programs to evaluate the safety analysis methodologies used in nuclear facility authorization basis documentation and to define future cost-effective support and development initiatives. Six areas, including source term development (fire, spills, and explosion analysis), in-facility transport, and dispersion/ consequence analysis (chemical and radiological) are contained in the APAC program. The evaluation process, codes considered, key results, and recommendations for future model and software development of the Radiological Dispersion/Consequence Working Group are summarized in this paper

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

    International Nuclear Information System (INIS)

    Behling, S.R.

    1985-05-01

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

  16. Chernobyl accident consequences in Germany: Nuclear safety and radiation protection

    International Nuclear Information System (INIS)

    Edelhauser, H.; Wendling, R.D.; Weiss, W.; Klonk, H.; Weil, L.

    1997-01-01

    A working Programme of the Federal Government was initiated on 26 May 1986 to cover all aspects of nuclear safety and public health, including research and public affairs in the light of the European and international activities resulting from the accident

  17. Radiological consequence of Chernobyl nuclear power accident in Japan

    International Nuclear Information System (INIS)

    Uchiyama, Masafumi; Nakamura, Yuji; Kankura, Takako; Iwasaki, Tamiko; Fujimoto, Kenzo; Kobayashi, Sadayoshi.

    1988-03-01

    Two years have elapsed since the accident in Chernobyl nuclear power station shocked those concerned with nuclear power generation. The effect that this accident exerted on human environment has still continued directly and indirectly, and the reports on the effect have been made in various countries and by international organizations. In Japan, about the exposure dose of Japanese people due to this accident, the Nuclear Safety Commission and Japan Atomic Energy Research Institute issued the reports. In this report, the available data concerning the envrionmental radioactivity level in Japan due to the Chernobyl accident are collected, and the evaluation of exposure dose which seems most appropriate from the present day scientific viewpoint was attempted by the detailed analysis in the National Institute of Radiological Sciences. The enormous number of the data observed in various parts of Japan were different in sampling, locality, time and measuring method, so difficulty arose frequently. The maximum concentration of I-131 in floating dust was 2.5 Bq/m 3 observed in Fukui, and the same kinds of radioactive nuclides as those in Europe were detected. (Kako, I.)

  18. Prevention of "simple accidents at work" with major consequences

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten

    2016-01-01

    broadly. This review identifies gaps in the prevention of simple accidents, relating to safety barriers for risk control and the management processes that need to be in place to deliver those risk controls in a continuingly effective state. The article introduces the ‘‘INFO cards’’ as a tool......The concept ‘‘simple accidents’’ is understood as traumatic events with one victim. In the last 10 years many European countries have seen a decline in the number of fatalities, but there still remain many severe accidents at work. In the years 2009–2010 in European countries 2.0–2.4 million...... occupational accidents a year were notified leading to 4500 fatalities and 90,000 permanent disabilities each year. The article looks at the concept ‘‘accident’’ to find similarities and distinctions between major and simple accident characteristics. The purpose is to find to what extent the same kinds...

  19. 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

  20. Consequences and countermeasures in a nuclear power accident: Chernobyl experience.

    Science.gov (United States)

    Kirichenko, Vladimir A; Kirichenko, Alexander V; Werts, Day E

    2012-09-01

    Despite the tragic accidents in Fukushima and Chernobyl, the nuclear power industry will continue to contribute to the production of electric energy worldwide until there are efficient and sustainable alternative sources of energy. The Chernobyl nuclear accident, which occurred 26 years ago in the former Soviet Union, released an immense amount of radioactivity over vast territories of Belarus, Ukraine, and the Russian Federation, extending into northern Europe, and became the most severe accident in the history of the nuclear industry. This disaster was a result of numerous factors including inadequate nuclear power plant design, human errors, and violation of safety measures. The lessons learned from nuclear accidents will continue to strengthen the safety design of new reactor installations, but with more than 400 active nuclear power stations worldwide and 104 reactors in the Unites States, it is essential to reassess fundamental issues related to the Chernobyl experience as it continues to evolve. This article summarizes early and late events of the incident, the impact on thyroid health, and attempts to reduce agricultural radioactive contamination.

  1. Structure shielding from cloud and fallout gamma ray sources for assessing the consequences of reactor accidents

    International Nuclear Information System (INIS)

    Burson, Z.G.; Profio, A.E.

    1975-12-01

    Radiation shielding provided by transportation vehicles and structures typical of where people live and work were estimated for cloud and fallout gamma-ray sources resulting from a hypothetical reactor accident. Dose reduction factors are recommended for a variety of situations for realistically assessing the consequences of reactor accidents

  2. Fuel solution criticality accident studies with the SILENE reactor: phenomenology, consequences and simulated intervention

    International Nuclear Information System (INIS)

    Barbry, F.

    1984-01-01

    After defining the content and the objectives of criticality accident studies, the SILENE reactor, a means of studying fuel solution criticality accidents, is presented. Information obtained from the CRAC and SILENE experimental programs are then presented; they concern power excursion phenomenology, radiological consequences, and finally guide-lines for current and future programs

  3. Development of a severe accident training simulator using a MELCOR code

    International Nuclear Information System (INIS)

    Kim, Ko Ryu; Jeong, Kwang Sub; Ha, Jae Joo; Jung, Won Dae

    2002-03-01

    Nuclear power plants' severe accidents are, despite of their rareness, very important in safety aspects, because of their huge damages when occurred. For the appropriate execution of severe accident strategy, more information for decision-making are required because of the uncertainties included in severe accidents. Earlier NRC raised concerns over severe accident training in the report NUREC/CR-477, and accordingly, developing effective training tools for severe accident were emphasized. In fact the training tools were requested from industrial area, nevertheless, few training tools were developed due to the uncertainties in severe accidents, lacks of analysis computer codes and technical limitations. SATS, the severe accident training simulator, is developed as a multi-purpose tools for severe accident training. SATS uses the calculation results of MELCOR, an integral severe accident analysis code, and with the help of SL-GMS graphic tools, provides dynamic displays of severe accident phenomena on the terminal of IBM PC. It aimed to have two main features: one is to provide graphic displays to represent severe accident phenomena and the other is to process and simulate severe accident strategy given by plant operators and TSC staffs. Severe accident strategies are basically composed of series of operations of available pumps, valves and other equipments. Whenever executing strategies with SATS, the trainee should follow the HyperKAMG, the on line version of the recently developed severe accident guidance (KAMG). Severe accident strategies are closely related to accidents scenarios. TLOFW and LOCA , two representative severe accident scenarios of Uljin 3,4, are developed as a built-in scenarios of SATS. Although SATS has some minor problems at this time, we expect SATS will be a good severe accident training tool after the appropriate addition of accident scenarios. Moreover, we also expect SATS will be a good advisory tool for the severe accident research

  4. Feasibility study on the rod ejection accident analysis with RETRAN-MASTER code system

    International Nuclear Information System (INIS)

    Kim, Y. H.; Lee, C. S.

    2003-01-01

    KEPRI has been developed the in-house methodology for non-LOCA safety analyses based on the codes and methodologies of vendors and EPRI. Using the methodology, the rod ejection accident, which is classified into the generic accident analysis category of reactivity insertion accident in primary system, has been analyzed with RETRAN-MASTER code system. And the feasibility of the coupled code system has been verified by the review of the results. Furthermore, to assess the important parameters to the accident, the sensitivity analyses have been carried out over some parameters

  5. Development of system of computer codes for severe accident analysis and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Jang, H S; Jeon, M H; Cho, N J. and others [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1992-01-15

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in nuclear power plants. This system of codes is necessary to conduct Individual Plant Examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident-resistance. Severe accident can be mitigated by the proper accident management strategies. Some operator action for mitigation can lead to more disastrous result and thus uncertain severe accident phenomena must be well recognized. There must be further research for development of severe accident management strategies utilizing existing plant resources as well as new design concepts.

  6. Development of system of computer codes for severe accident analysis and its applications

    International Nuclear Information System (INIS)

    Jang, H. S.; Jeon, M. H.; Cho, N. J. and others

    1992-01-01

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in nuclear power plants. This system of codes is necessary to conduct Individual Plant Examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident-resistance. Severe accident can be mitigated by the proper accident management strategies. Some operator action for mitigation can lead to more disastrous result and thus uncertain severe accident phenomena must be well recognized. There must be further research for development of severe accident management strategies utilizing existing plant resources as well as new design concepts

  7. The international conference ''one decade after Chernobyl: Summing up the consequences of the accident''

    International Nuclear Information System (INIS)

    1996-01-01

    An International Conference entitled ''One decade after Chernobyl: Summing up the consequences of the accident'' was held at the Austria Center Vienna from 8 to 12 April 1996, the aim being to seek a common and conclusive understanding of the nature and magnitude of the consequences of the Chernobyl accident. The Conference was attended by 845 participants and observers from 71 countries and 20 organizations and covered by 208 journalists from 31 countries and two organizations

  8. Accident-generated radioactive particle source term development for consequence assessment of nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Sutter, S.L.; Ballinger, M.Y.; Halverson, M.A.; Mishima, J.

    1983-04-01

    Consequences of nuclear fuel cycle facility accidents can be evaluated using aerosol release factors developed at Pacific Northwest Laboratory. These experimentally determined factors are compiled and consequence assessment methods are discussed. Release factors can be used to estimate the fraction of material initially made airborne by postulated accident scenarios. These release fractions in turn can be used in models to estimate downwind contamination levels as required for safety assessments of nuclear fuel cycle facilities. 20 references, 4 tables

  9. Safety analysis of the IAEA reference research reactor during loss of flow accident using the code MERSAT

    International Nuclear Information System (INIS)

    Hainoun, A.; Ghazi, N.; Abdul-Moaiz, B. Mansour

    2010-01-01

    Using the thermal hydraulic code MERSAT detailed model including primary and secondary loop was developed for the IAEA's reference research reactor MTR 10 MW. The developed model enables the simulation of expected neutronic and thermal hydraulic phenomena during normal operation, reactivity and loss of flow accidents. Two different loss of flow accident (LOFA) have been simulated using slow and fast decrease time of core mass flow. In both cases the expected flow reversal from downward forced to upward natural circulation has been successfully simulated. The results indicate that in both accidents the limit of onset of subcooled boiling was not arrived and consequently no exceed of design limits in term of thermal hydraulic instability or DNB is observed. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermal hydraulic system codes.

  10. The Chernobyl reactor accident and its consequences. 3. rev. ed.

    International Nuclear Information System (INIS)

    1987-01-01

    The report presents a comprehensive survey of measured data explaining the radiation exposure in the Land Hessen, and a chronological survey of the decisions and measures taken by the Hessian regional government in response to the Chernobyl reactor accident. The measures for instance included selection of appropriate measuring methods and sites, checking of various environmental material, waste disposal surveillance, and dose assessments, and a range of monitoring programmes. (PW) [de

  11. Radiation accidents: occurrence, types, consequences, medical management, and the lessons to be learned

    International Nuclear Information System (INIS)

    Turai, I.; Veress, K.

    2001-01-01

    The paper reviews the frequency, causes and occurrence of radiation accidents with some significant exposure to human. More detailed information is provided in tabulated form on the health consequences of those twenty severe radiation accidents that occurred in 1986-2000, world-wide. Reference is given to the very low cumulative incidence of significant radiation accidents, as during the last 57 years there were, in average, seven registered accidents annually in all countries of the world. Thus, the chance for most of the physicians to meet a patient with symptoms of acute radiation injury during their professional career is very low

  12. Guidelines for calculation of atmospheric dispersion and radiological consequences of design basis reactor accidents - Severe accident calculation guidelines, EPR

    International Nuclear Information System (INIS)

    Martens, R.; Schmitz, B.M.; Horn, M.

    1999-01-01

    The activities carried out within the (reduced) project period (1. Sept. until 31. Dec. 1998) for coordinated harmonization between France and Germany, of guidelines for calculation of the radiological consequences of a severe reactor accident, are summarized. (orig./CB) [de

  13. Problems of softening the Chernobyl accident consequences. Proceedings of the International seminar. Pt. 1

    International Nuclear Information System (INIS)

    1993-01-01

    Proceedings of the International seminar on the Problems to soften the Chernobyl accident consequences held by the International Association of Dissemination of Knowledge and the Russian branch of the Society on the Dissemination of Knowledge in Bryansk in 1993. The proceedings of the seminar deal with the study of scientific and practical activity linked with the elimination of the Chernobyl accident effects. Main theoretical concepts used as the basis of the elaborated regulations are presented, as well; ways and techniques to soften the consequences of the Chernobyl accident to decontaminate the affected territories and to protect the population health are discussed

  14. Improvement of Severe Accident Analysis Computer Code and Development of Accident Management Guidance for Heavy Water Reactor

    International Nuclear Information System (INIS)

    Park, Soo Yong; Kim, Ko Ryu; Kim, Dong Ha; Kim, See Darl; Song, Yong Mann; Choi, Young; Jin, Young Ho

    2005-03-01

    The objective of the project is to develop a generic severe accident management guidance(SAMG) applicable to Korean PHWR and the objective of this 3 year continued phase is to construct a base of the generic SAMG. Another objective is to improve a domestic computer code, ISAAC (Integrated Severe Accident Analysis code for CANDU), which still has many deficiencies to be improved in order to apply for the SAMG development. The scope and contents performed in this Phase-2 are as follows: The characteristics of major design and operation for the domestic Wolsong NPP are analyzed from the severe accident aspects. On the basis, preliminary strategies for SAM of PHWR are selected. The information needed for SAM and the methods to get that information are analyzed. Both the individual strategies applicable for accident mitigation under PHWR severe accident conditions and the technical background for those strategies are developed. A new version of ISAAC 2.0 has been developed after analyzing and modifying the existing models of ISAAC 1.0. The general SAMG applicable for PHWRs confirms severe accident management techniques for emergencies, provides the base technique to develop the plant specific SAMG by utility company and finally contributes to the public safety enhancement as a NPP safety assuring step. The ISAAC code will be used inevitably for the PSA, living PSA, severe accident analysis, SAM program development and operator training in PHWR

  15. Improvement of Severe Accident Analysis Computer Code and Development of Accident Management Guidance for Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Kim, Ko Ryu; Kim, Dong Ha; Kim, See Darl; Song, Yong Mann; Choi, Young; Jin, Young Ho

    2005-03-15

    The objective of the project is to develop a generic severe accident management guidance(SAMG) applicable to Korean PHWR and the objective of this 3 year continued phase is to construct a base of the generic SAMG. Another objective is to improve a domestic computer code, ISAAC (Integrated Severe Accident Analysis code for CANDU), which still has many deficiencies to be improved in order to apply for the SAMG development. The scope and contents performed in this Phase-2 are as follows: The characteristics of major design and operation for the domestic Wolsong NPP are analyzed from the severe accident aspects. On the basis, preliminary strategies for SAM of PHWR are selected. The information needed for SAM and the methods to get that information are analyzed. Both the individual strategies applicable for accident mitigation under PHWR severe accident conditions and the technical background for those strategies are developed. A new version of ISAAC 2.0 has been developed after analyzing and modifying the existing models of ISAAC 1.0. The general SAMG applicable for PHWRs confirms severe accident management techniques for emergencies, provides the base technique to develop the plant specific SAMG by utility company and finally contributes to the public safety enhancement as a NPP safety assuring step. The ISAAC code will be used inevitably for the PSA, living PSA, severe accident analysis, SAM program development and operator training in PHWR.

  16. The Fukushima accident and its consequences. Facts, explanations and comments; L'accident de Fukushima et ses consequences. Faits, explications et commentaires

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-03-06

    This document proposes an overview of the present situation in the different reactors of the Fukushima power station and discusses its control by the operator. It also describes what went on, the causes of the accident, and what occurred on the accident day (earthquake, tsunami, flooding). It discusses whether some mistakes regarding the design and the protection of reactors could explain the accident. It presents the various measures which have been immediately implemented to protect the populations and to confine the accident. It proposes an assessment of damages for the ground and marine environment in terms of contamination. It addresses the consequences of the released radioactivity on population health and on personnel intervening within the site. It discusses the restoration perspectives for contaminated areas and the possible return of evacuated population. Then, it describes the different phases for the station dismantling. It evokes the issue of fallouts beyond Japan and in Europe, outlines some lessons learned from the accident and new safety measures to be implemented in France. It discusses how nuclear risk management is organised in France and its efficiency. It addresses the consequences for the development of nuclear energy in the world

  17. COSYMA, a mainframe and PC program package for assessing the consequences of hypothetical accidents

    International Nuclear Information System (INIS)

    Jones, J.A.; Hasemann, I.; Steen, J. van der

    1996-01-01

    COSYMA (Code System from MARIA) is a program package for assessing the off-site consequences of accidental releases of radioactive material to atmosphere, developed as part of the European Commission's MARIA programme (Methods for Assessing the Radiological Impact of Accidents). COSYMA represents a fusion of ideas and modules from the Forschungszetrum Karlsruhe program system UFOMOD, the National Radiological Protection Board program MARC and new model developments together with data libraries from other MARIA contractors. Mainframe and PC versions of COSYMA are distributed to interested users by arrangement with the European Commission. The system was first released in 1990, and has subsequently been updated. The program system uses independent modules for the different parts of the analysis, and so permits a flexible problem-oriented application to different sites, source terms, emergency plans and the needs of users in the various parts of Europe. Users of the mainframe system can choose the most appropriate combination of modules for their particular application. The PC version includes a user interface which selects the required modules for the endpoints specified by the user. This paper describes the structure of the mainframe and PC versions of COSYMA, and summarises the models included in them. The mainframe or PC versions of COSYMA have been distributed to more than 100 organisations both inside and outside the European Union, and have been used in a wide variety of applications. These range from full PRA level 3 analyses of nuclear power and research reactors to investigations on advanced containment concepts and the preplanning of off-site emergency actions. Some of the experiences from these applications are described in the paper. An international COSYMA user group has been established to stimulate communication between the owners, developers and users of the code and to serve as a reference point for questions relating to the code. The group produces

  18. One decade after Chernobyl. Summing up the consequences of the accident. Proceedings of an international conference

    International Nuclear Information System (INIS)

    1996-01-01

    The consequences attributed to the disastrous accident that occurred at the Chernobyl nuclear power plant on 26 April 1986 have been subjected to extensive scientific examination; however, they are still viewed with widely differing perspectives. It is fitting then that, ten years after the accident, the European Commission (EC), the International Atomic Energy Agency (IAEA) and the World Health Organization (WHO) should jointly sponsor an international conference to review the consequences of the accident and to seek a common and conclusive understanding of their nature and magnitude. The International Conference on One Decade after Chernobyl: Summing up the Consequences of the Accident was held at the Austria Center, Vienna, on 8-12 April 1996. Refs, figs, tabs

  19. Cancer consequences of the Chernobyl accident: 20 years on

    International Nuclear Information System (INIS)

    Cardis, Elisabeth; Howe, Geoffrey; Ron, Elaine

    2006-01-01

    26 April 2006 marks the 20th anniversary of the Chernobyl accident. On this occasion, the World Health Organization (WHO), within the UN Chernobyl Forum initiative, convened an Expert Group to evaluate the health impacts of Chernobyl. This paper summarises the findings relating to cancer. A dramatic increase in the incidence of thyroid cancer has been observed among those exposed to radioactive iodines in childhood and adolescence in the most contaminated territories. Iodine deficiency may have increased the risk of developing thyroid cancer following exposure to radioactive iodines, while prolonged stable iodine supplementation in the years after exposure may reduce this risk. Although increases in rates of other cancers have been reported, much of these increases appear to be due to other factors, including improvements in registration, reporting and diagnosis. Studies are few, however, and have methodological limitations. Further, because most radiation-related solid cancers continue to occur decades after exposure and because only 20 years have passed since the accident, it is too early to evaluate the full radiological impact of the accident. Apart from the large increase in thyroid cancer incidence in young people, there are at present no clearly demonstrated radiation-related increases in cancer risk. This should not, however, be interpreted to mean that no increase has in fact occurred: based on the experience of other populations exposed to ionising radiation, a small increase in the relative risk of cancer is expected, even at the low to moderate doses received. Although it is expected that epidemiological studies will have difficulty identifying such a risk, it may nevertheless translate into a substantial number of radiation-related cancer cases in the future, given the very large number of individuals exposed. (review)

  20. Cancer consequences of the Chernobyl accident: 20 years on

    Energy Technology Data Exchange (ETDEWEB)

    Cardis, Elisabeth [International Agency for Research on Cancer, 150 Cours Albert Thomas, 69372 Lyon CEDEX 08 (France); Howe, Geoffrey [Department of Epidemiology, Mailman School of Public Health, Columbia University, 722 W. 168th Street, Room 1104, New York, NY 10032 (United States); Ron, Elaine [Radiation Epidemiology Branch, Division of Epidemiology and Genetics, National Cancer Institute, Building EPS, MS 7238, Rockville, MD 20852 (United States)] (and others)

    2006-06-15

    26 April 2006 marks the 20th anniversary of the Chernobyl accident. On this occasion, the World Health Organization (WHO), within the UN Chernobyl Forum initiative, convened an Expert Group to evaluate the health impacts of Chernobyl. This paper summarises the findings relating to cancer. A dramatic increase in the incidence of thyroid cancer has been observed among those exposed to radioactive iodines in childhood and adolescence in the most contaminated territories. Iodine deficiency may have increased the risk of developing thyroid cancer following exposure to radioactive iodines, while prolonged stable iodine supplementation in the years after exposure may reduce this risk. Although increases in rates of other cancers have been reported, much of these increases appear to be due to other factors, including improvements in registration, reporting and diagnosis. Studies are few, however, and have methodological limitations. Further, because most radiation-related solid cancers continue to occur decades after exposure and because only 20 years have passed since the accident, it is too early to evaluate the full radiological impact of the accident. Apart from the large increase in thyroid cancer incidence in young people, there are at present no clearly demonstrated radiation-related increases in cancer risk. This should not, however, be interpreted to mean that no increase has in fact occurred: based on the experience of other populations exposed to ionising radiation, a small increase in the relative risk of cancer is expected, even at the low to moderate doses received. Although it is expected that epidemiological studies will have difficulty identifying such a risk, it may nevertheless translate into a substantial number of radiation-related cancer cases in the future, given the very large number of individuals exposed. (rev0009i.

  1. 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.

  2. The radioecological consequences of Chernobyl accident for fish

    International Nuclear Information System (INIS)

    Ryabov, I.N.

    1997-01-01

    The estimate of dynamics of radionuclides concentration in muscles of some game-fish from Kiev reservoir and likes in Bryansk region for period after Chernobyl accident was carried out. The concentration of 137 Cs in fish has not exceeded the admissible concentration (600 Bq/kg ww) since 1993. The exceptions are the cooling-pond of Chernobyl NPP and Kozlanovskoe Lake where the concentration of 137 Cs in fish's muscles exceeded the admissible level more than 5-6 times even in 1995. It was concluded that chronic irradiation of game-fish in water bodies outside 30-km zone would not affect the volume of fishing

  3. Uncertainty analysis in calculations of a road accident consequences

    International Nuclear Information System (INIS)

    Bonnefous, S.; Brenot, J.; Hubert, P.

    1995-01-01

    This paper develops a concrete situation witch is the search for an evacuation distance in case of a road accident implying a chlorine tank. The methodological aspect is how implementing uncertainty analysis in deterministic models with random parameters. The study demonstrates a great dispersion in the results. It allows to establish satisfactory decision rules and a hierarchy on parameters witch is useful to define priorities in the search for information and to improve the treatment of these parameters. (authors). 8 refs., 1 fig., 2 tabs

  4. The Fukushima Accident: A Station Blackout and the Consequences

    International Nuclear Information System (INIS)

    Schäfer, F.; Tusheva, P.; Kliem, S.

    2012-01-01

    Lessons learned from Fukushima: • Underestimation of the role of the natural hazards • Insufficient protection of the emergency power and service water systems • Protection of fuel assembly storage pools insufficient • Safety review for Station Blackout and seismic evaluation needed • Diverse power supply systems, diverse sources for water delivery • Role of passive safety systems, they must work in a real passive manner and without electricity to open valves • Backup systems for reactor parameters monitoring • Revision of Severe Accident Management Guidelines and countermeasures for specific “rare” events • Early/late phase operators’ actions / Effectiveness of the operators’ actions

  5. Development status of Severe Accident Analysis Code SAMPSON

    International Nuclear Information System (INIS)

    Iwashita, Tsuyoshi; Ujita, Hiroshi

    2000-01-01

    The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology' project Phase 1 have been completed. The verification study of Severe Accident Analysis Code SAMPSON prototype developed in Phase 1 was conducted in two steps. First, each analysis module was run independently and analysis results were compared and verified against separate-effect test data with good results. Test data are as follows: CORA-13 (FZK) for the Core Heat-up Module; VI-3 of HI/VI Test (ORNL) for the FP Release from Fuel Module; KROTOS-37 (JRC-ISPRA) for the Molten Core Relocation Module; Water Spread Test (UCSB) for the Debris Spreading Model and Benard's Melting Test for Natural Convection Model in the Debris Cooling Module; Hydrogen Burning Test (NUPEC) for the Ex-Vessel Thermal Hydraulics Module; PREMIX, PM10 (FZK) for the Steam Explosion Module; and SWISS-2 (SNL) for the Debris-Concrete Interaction Module. Second, with the Simulation Supervisory System, up to 11 analysis modules were executed concurrently in the parallel environment (currently, NUPEC uses IBM-SP2 with 72 process elements), to demonstrate the code capability and integrity. The target plant was Surry as a typical PWR and the initiation events were a 10-inch cold leg failure. The analysis is divided to two cases; one is in-vessel retention analysis when the gap cooling is effective (In-vessel scenario test), the other is analysis of phenomena event is extended to ex-vessel due to the Reactor Pressure Vessel failure when the gap cooling is not sufficient (Ex-vessel scenario test). The system verification test has confirmed that the full scope of the scenarios can be analyzed and phenomena occurred in scenarios can be simulated qualitatively reasonably considering the physical models used for the situation. The Ministry of International Trade and Industry, Japan sponsors this work. (author)

  6. Consequences of the Fukushima accident on Germany's energy policy

    International Nuclear Information System (INIS)

    Guidez, J.

    2011-01-01

    Germany is the country that most quickly reacted to the Fukushima accident by adopting a 3 month long moratory on its own law on reactor service life extension. This moratory led to the immediate stopping of the 7 oldest nuclear power reactors. An ethical commission, set just after the accident, concluded in end May, that Germany has the opportunity to replace nuclear energy with technologies less hazardous for the society and recommended a complete withdrawal from nuclear power in 10 years. The first resistance tests performed on German reactors showed that the 7 reactors concerned by the moratory have an inferior standard of protection against plane crashes. An energy law passed on the 30. June 2011 imposes that: -) the 7 reactors concerned by the moratory will be definitely closed, -) renewable energies will be developed massively (particularly offshore wind power), and -) new gas and coal power plant will have to assure the transition. German nuclear industry faces a huge reorganization and expects a loss of 11.000 jobs. (A.C.)

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

    International Nuclear Information System (INIS)

    Tentner, A.M.; Parma, E.; Wei, T.; Wigeland, R.

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-01

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

  9. A simple assessment scheme for severe accident consequences using release parameters

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Kampanart, E-mail: kampanarts@tint.or.th [Thailand Institute of Nuclear Technology, 16 Vibhavadi-Rangsit Rd., Latyao, Chatuchak, 10900 (Thailand); Okamoto, Koji [The University of Tokyo, 7-3-1 Hongo, Bunkyo, Tokyo 113-8654 (Japan)

    2016-08-15

    Highlights: • Nuclear accident consequence index can assess overall consequences of an accident. • Correlations between the index and release parameters are developed. • Relation between the index and release amount follows power function. • The exponent of the power function is the key to the relation. - Abstract: Nuclear accident consequence index (NACI) which can assess the overall consequences of a severe accident on people and the environment is developed based on findings from previous studies. It consists of three indices: radiation effect index, relocation index and decontamination index. Though the NACI can cover large range of consequences, its assessment requires extensive resources. The authors then attempt to simplify the assessment, by investigating the relations between the release parameters and the NACI, in order to use the release parameters for severe accident consequence assessment instead of the NACI. NACI and its components increase significantly when the release amount is increased, while the influences of the release period and the release starting time on the NACI are nearly negligible. Relations between the release amount and the NACI and its components follow simple power functions (y = ax{sup b}). The exponent of the power functions seems to be the key to the relations. The exponent of the relation between the release amount and the NACI was around 0.8–1.0 when the release amount is smaller than 100 TBq, and it increased to around 1.3–1.4 when the release amount is equal to or larger than 100 TBq.

  10. Consequences in Norway after a hypothetical accident at Sellafield - Predicted impacts on the environment.

    Energy Technology Data Exchange (ETDEWEB)

    Thoerring, H.; Liland, A.

    2010-12-15

    This report deals with the environmental consequences in Norway after a hypothetical accident at Sellafield. The investigation is limited to the terrestrial environment, and focus on animals grazing natural pastures, plus wild berries and fungi. Only 137Cs is considered. The predicted consequences are severe, in particular for mutton and goat milk production. (Author)

  11. Consequences in Norway after a hypothetical accident at Sellafield - Predicted impacts on the environment

    International Nuclear Information System (INIS)

    Thoerring, H.; Liland, A.

    2010-12-01

    This report deals with the environmental consequences in Norway after a hypothetical accident at Sellafield. The investigation is limited to the terrestrial environment, and focus on animals grazing natural pastures, plus wild berries and fungi. Only 137Cs is considered. The predicted consequences are severe - in particular for mutton and goat milk production. (Author)

  12. Small chances - great consequences or the consequences of a large-scale accident in a nuclear power plant

    International Nuclear Information System (INIS)

    Dijk, G. van; Smit, W.A.

    1977-01-01

    This report is a sequel to the previous Boerderijcahier (no. 7502) which discussed long-term effects of soil contamination in case of a nuclear power plant accident. In this report the short-term health effects are discussed. Models describing the local consequences of a severe accident are developed, taking into account the possible weather conditions (meteorological model), the evacuation possibilities and the inhabitability of certain areas. In each case long-term and short-term effects are discussed. The safety studies by various departments of the Netherlands' government and the Rasmussen report are commented on

  13. SHOCK WAVE ANALYSIS OF THE CONSEQUENCES OF A REACTOR ACCIDENT

    Energy Technology Data Exchange (ETDEWEB)

    Klickman, A E; Nicholson, R B; Nims, J B

    1963-06-15

    The solution to the problem of transmission and attenuation of the shock wave resulting from a large reactor accident is demonstrated for a configuration typical of many reactors. The particular configuration is that of a spherical gas bubble surrounded by one or more concentric regions of compressible material. A systematic parameter study was made in which the physical characteristics of the compressible shield regions and the expansion characteristics of a gas were assumed to be parameters. Results for seven cases are shown, and similar cases with only one important difference are compared. From these comparisons it was concluded that under certain conditions alternative materials can be substituted for reactor materials in model experiments and TNT can be used as an energy source instead of uranium. In the outer crushable region the total mass of material is the important factor. (A.G.W.)

  14. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject F. Contributions to code validation using BWR data and to evaluation and optimization of accident management measures. Final report

    International Nuclear Information System (INIS)

    Di Marcello, Valentino; Imke, Uwe; Sanchez Espinoza, Victor

    2016-09-01

    The exact knowledge of the transient course of events and of the dominating processes during a severe accident in a nuclear power station is a mandatory requirement to elaborate strategies and measures to minimize the radiological consequences of core melt. Two typical experiments using boiling water reactor assemblies were modelled and simulated with the severe accident simulation code ATHLET-CD. The experiments are related to the early phase of core degradation in a boiling water reactor. The results reproduce the thermal behavior and the hydrogen production due to oxidation inside the bundle until relocation of material by melting. During flooding of the overheated assembly temperatures and hydrogen oxidation are under estimated. The deviations from the experimental results can be explained by the missing model to simulate bore carbide oxidation of the control rods. On basis of a hypothetical loss of coolant accident in a typical German boiling water reactor the effectivity of flooding the partial degraded core is investigated. This measure of mitigation is efficient and prevents failure of the reactor pressure vessel if it starts before molten material is relocated into the lower plenum. Considerable amount of hydrogen is produced by oxidation of the metallic components.

  15. Experience with psychological consequences of the Chernobyl nuclear plant accident

    International Nuclear Information System (INIS)

    Buzunov, V.A.; Druzhinin, A.M.; Druzhinina, E.S.

    1993-01-01

    The paper describes the image of radiation menance. Basic differences in image parameters are revealed for some population groups. The psychological levels of the image are regarded as psychological phenomena. Some specific psychological consequences of mental regression are outlined in the paper

  16. One decade after Chernobyl: Summing up the consequences of the accident. Poster presentations

    International Nuclear Information System (INIS)

    1997-09-01

    The consequences attributed to the disastrous accident that occurred at the Chernobyl nuclear power plant on 26 April 1986 have been subjected to extensive scientific examination; however, they are still viewed with widely differing perspectives. It is fitting then that, ten years after the accident, the European Commission (EC), the International Atomic Energy Agency (IAEA) and the World Health Organization (WHO) should jointly sponsor an international conference to review the consequences of the accident and to seek a common and conclusive understanding of their nature and magnitude. The International Conference on One Decade after Chernobyl: Summing up the Consequences of the Accident was held at the Austria Center, Vienna, on 8-12 April 1996. To facilitate the discussions of the Conference, background papers were prepared for the Technical Symposium by teams of scientists from around the world, who collaborated over a period of months to ascertain, consolidate and present the current state of knowledge in six key areas: clinically observed effects; thyroid effects; long term health effects; other health related effects; consequences for the environment; and the consequences in perspective: prognosis for the future. A background paper on the social, economic, institutional and political impact of the accident was prepared by Belarus, the Russian Federation and Ukraine. The conclusions of the Forum on Nuclear Safety Aspects served as a background paper on this topic

  17. One decade after Chernobyl: Summing up the consequences of the accident. Poster presentations

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    The consequences attributed to the disastrous accident that occurred at the Chernobyl nuclear power plant on 26 April 1986 have been subjected to extensive scientific examination; however, they are still viewed with widely differing perspectives. It is fitting then that, ten years after the accident, the European commission (EC), the International Atomic Energy Agency (IAEA) and the World Health Organization (WHO) should jointly sponsor an international conference to review the consequences of the accident and to seek a common and conclusive understanding of their nature and magnitude. The International Conference on One Decade after Chernobyl: Summing up the Consequences of the Accident was held at the Austria Center, Vienna, on 8-12 April 1996. To facilitate the discussions of the Conference, background papers were prepared for the Technical Symposium by teams of scientists from a round the world, who collaborated over a period of months to ascertain, consolidate and present the current state of knowledge in six key areas: clinically observed effects; thyroid effects; long term health effects; other health related effects; consequences for the environment; and the consequences in perspective: prognosis for the future. A background paper on the social, economic, institutional and political impact of the accident was prepared by Belarus, the Russian Federation and Ukraine. The conclusions of the Forum on Nuclear Safety Aspects served as a background paper on this topic. Refs, figs, tabs.

  18. One decade after Chernobyl: Summing up the consequences of the accident. Poster presentations

    International Nuclear Information System (INIS)

    1997-09-01

    The consequences attributed to the disastrous accident that occurred at the Chernobyl nuclear power plant on 26 April 1986 have been subjected to extensive scientific examination; however, they are still viewed with widely differing perspectives. It is fitting then that, ten years after the accident, the European commission (EC), the International Atomic Energy Agency (IAEA) and the World Health Organization (WHO) should jointly sponsor an international conference to review the consequences of the accident and to seek a common and conclusive understanding of their nature and magnitude. The International Conference on One Decade after Chernobyl: Summing up the Consequences of the Accident was held at the Austria Center, Vienna, on 8-12 April 1996. To facilitate the discussions of the Conference, background papers were prepared for the Technical Symposium by teams of scientists from a round the world, who collaborated over a period of months to ascertain, consolidate and present the current state of knowledge in six key areas: clinically observed effects; thyroid effects; long term health effects; other health related effects; consequences for the environment; and the consequences in perspective: prognosis for the future. A background paper on the social, economic, institutional and political impact of the accident was prepared by Belarus, the Russian Federation and Ukraine. The conclusions of the Forum on Nuclear Safety Aspects served as a background paper on this topic. Refs, figs, tabs

  19. One decade after Chernobyl: Summing up the consequences of the accident. Poster presentations

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    The consequences attributed to the disastrous accident that occurred at the Chernobyl nuclear power plant on 26 April 1986 have been subjected to extensive scientific examination; however, they are still viewed with widely differing perspectives. It is fitting then that, ten years after the accident, the European Commission (EC), the International Atomic Energy Agency (IAEA) and the World Health Organization (WHO) should jointly sponsor an international conference to review the consequences of the accident and to seek a common and conclusive understanding of their nature and magnitude. The International Conference on One Decade after Chernobyl: Summing up the Consequences of the Accident was held at the Austria Center, Vienna, on 8-12 April 1996. To facilitate the discussions of the Conference, background papers were prepared for the Technical Symposium by teams of scientists from around the world, who collaborated over a period of months to ascertain, consolidate and present the current state of knowledge in six key areas: clinically observed effects; thyroid effects; long term health effects; other health related effects; consequences for the environment; and the consequences in perspective: prognosis for the future. A background paper on the social, economic, institutional and political impact of the accident was prepared by Belarus, the Russian Federation and Ukraine. The conclusions of the Forum on Nuclear Safety Aspects served as a background paper on this topic. Ref, figs, tabs.

  20. Consequences of radioactive fallout - Experience acquired from the Chernobyl accident

    International Nuclear Information System (INIS)

    Lutzko Alexandre

    1997-01-01

    The paper presents considerations on the nature of protection measures which were taken following the Chernobyl accident. Six methods used in Belarus aiming at reducing the radionuclide concentration in soils are listed which resulted in a reduction by a factor of 1.5 to 2 of the contamination levels in different agricultural products after 1986. A table is given showing the continuous decrease between 1986 and 1991 in the amount of sold meat and milk on market having doses exceeding the admissible values. In 1994 an increase in the ratio of contaminated products is explained as being due to investment reduction and to augmentation of transfer coefficients produced by an extremely hot summer. In conclusion it is stressed that lifting and removing the superficial layers of contaminated soils proved to be an inefficient method of decontaminating vast areas. Certain lands will not be cultivated for long as no protective action can be conducted to get them returned to agriculture. If one forgives for an instant the natural decay it seems that never one can get rid of the radioactivity. An artificial reduction methods should be supported by an international experience in the field

  1. Consequences of the Chernobyl accident for reindeer husbandry in Sweden

    International Nuclear Information System (INIS)

    Åhman, G.; Rydberg, A.; Åhman, B.

    1990-01-01

    Large parts of the reindeer hearding area in Sweden were contaminated with radioactive caesium from the Chernobyl fallout. During the first year after the accident no food with activity concentrations exceeding 300 Bq/kg was allowed to be sold in Sweden. This meant that about 75% of all reindeer meat produced in Sweden during the autumn and winter 1986/87 were rejected because of too high caesium activités. In May 1987 the maximum level for Cs-137 in reindeer, game and fresh-water fish was raised to 1500 Bq/kg. During the last two year, 1987/88 and 1988/89, about 25% of the slaughtered reindeer has had activities exceeding this limit. The effective long-time halflife or radiocaesium in reindeer after the nuclear weapon tests in the sixties was about 7 years. If this halflife is correct also for the Chernobyl fallout it will take about 35 years before most of the reinder in Sweden are below the current limit 1500 Bq/kg in the winter. However, by feeding the animals uncontaminated food for about two months, many reindeer can be saved for human consumption

  2. Consequences of the Chernobyl accident for reindeer husbandry in Sweden

    Directory of Open Access Journals (Sweden)

    Gustaf Åhman

    1990-09-01

    Full Text Available Large parts of the reindeer hearding area in Sweden were contaminated with radioactive caesium from the Chernobyl fallout. During the first year after the accident no food with activity concentrations exceeding 300 Bq/kg was allowed to be sold in Sweden. This meant that about 75% of all reindeer meat produced in Sweden during the autumn and winter 1986/87 were rejected because of too high caesium activités. In May 1987 the maximum level for Cs-137 in reindeer, game and fresh-water fish was raised to 1500 Bq/kg. During the last two year, 1987/88 and 1988/89, about 25% of the slaughtered reindeer has had activities exceeding this limit. The effective long-time halflife or radiocaesium in reindeer after the nuclear weapon tests in the sixties was about 7 years. If this halflife is correct also for the Chernobyl fallout it will take about 35 years before most of the reinder in Sweden are below the current limit 1500 Bq/kg in the winter. However, by feeding the animals uncontaminated food for about two months, many reindeer can be saved for human consumption.

  3. OFFSITE RADIOLOGICAL CONSEQUENCE CALCULATION FOR THE BOUNDING MIXING OF INCOMPATIBLE MATERIALS ACCIDENT

    International Nuclear Information System (INIS)

    SANDGREN, K.R.

    2006-01-01

    This document quantifies the offsite radiological consequence of the bounding mixing of incompatible materials accident for comparison with the 25 rem Evaluation Guideline established in Appendix A of DOE-STD-3009. The bounding accident is an inadvertent addition of acid to a waste tank. The calculated offsite dose does not challenge the Evaluation Guideline. Revision 4 updates the analysis to consider bulk chemical additions to single shell tanks (SSTs)

  4. Impact of source terms on distances to which reactor accident consequences occur

    International Nuclear Information System (INIS)

    Ostmeyer, R.M.

    1982-01-01

    Estimates of the distances over which reactor accident consequences might occur are important for development of siting criteria and for emergency response planning. This paper summarizes the results of a series of CRAC2 calculations performed to estimate these distances. Because of the current controversy concerning the magnitude of source terms for severe accidents, the impact of source term reductions upon distance estimates is also examined

  5. Accident on the Chernobyl nuclear power plant. Getting over the consequences and lessons learned

    International Nuclear Information System (INIS)

    Nosovskij, A.V.; Vasil'chenko, V.N.; Klyuchnikov, A.A.; Prister, B.S.

    2006-01-01

    The book is devoted to the 20 anniversary of the accident on the 4th Power Unit of the Chernobyl NPP. The power plant construction history, accident reasons, its consequences, the measures on its liquidation are represented. The current state of activity on the Chernobyl power unit decommission, the 'Shelter' object conversion into the ecologically safe system is described. The future of the Chernobyl NPP site and disposal zone is discussed

  6. Confidence level in the calculations of HCDA consequences using large codes

    International Nuclear Information System (INIS)

    Nguyen, D.H.; Wilburn, N.P.

    1979-01-01

    The probabilistic approach to nuclear reactor safety is playing an increasingly significant role. For the liquid-metal fast breeder reactor (LMFBR) in particular, the ultimate application of this approach could be to determine the probability of achieving the goal of a specific line-of-assurance (LOA). Meanwhile a more pressing problem is one of quantifying the uncertainty in a calculated consequence for hypothetical core disruptive accident (HCDA) using large codes. Such uncertainty arises from imperfect modeling of phenomenology and/or from inaccuracy in input data. A method is presented to determine the confidence level in consequences calculated by a large computer code due to the known uncertainties in input invariables. A particular application was made to the initial time of pin failure in a transient overpower HCDA calculated by the code MELT-IIIA in order to demonstrate the method. A probability distribution function (pdf) for the time of failure was first constructed, then the confidence level for predicting this failure parameter within a desired range was determined

  7. ASSESSMENT OF THE FUKUSIMA NUCLEAR POWER PLANT ACCIDENT CONSEQUENCES BY THE POPULATION IN THE FAR EAST

    Directory of Open Access Journals (Sweden)

    G. V. Arkhangelskaya

    2012-01-01

    Full Text Available The article analyzes the attitude of the population in the five regions of the Far East to the consequences of the accident at the Fukushimai nuclear power plant, as well as the issues of informing about the accident. The analysis of public opinion is based on the data obtained by anonymous questionnaire survey performed in November 2011. In spite of the rather active informing and objective information on the absence of the contamination, most of the population of the Russian Far East believes that radioactive contamination is presented in the areas of their residence, and the main cause of this contamination is the nuclear accident in Japan.

  8. [Health-related consequences of obstructive sleep apnea: daytime sleepiness, accident risk and legal aspects].

    Science.gov (United States)

    Orth, M; Kotterba, S

    2012-04-01

    Daytime sleepiness for any reason leads to impairment of daytime performance and an increased accident rate. The consequences are an increase of illness- and accident-related costs for the health system. Obstructive sleep apnea (OSA) is one of the major reasons for increased daytime sleepiness, especially in professional drivers. The accident frequency in OSA can be significantly reduced by adequate continuous positive airway pressure (CPAP) therapy. Up till now there are no uniform legal regulations about the handling of OSAS patients or patients with daytime sleepiness due to other diseases as far as driving ability is concerned.

  9. Comparison of the foodchain transport models of WASH-1400 and MARC using the accident consequence model UFOMOD

    International Nuclear Information System (INIS)

    Steinhauer, C.

    1985-04-01

    Within the frame of the contract with the European Community 'Methods for Assessing the Radiological Impact of Accidents' (CEC-MARIA) comparative accident consequence assessments were performed with the computer code UFOMOD, replacing the currently implemented foodchain transport model of the WASH-1400 study by the dynamic transport model of the MARC methodology. The calculations were based on the release category FK2 of the German Risk Study with meteorological data representing four different regions of the Federal Republic of Germany. The study of seasonal variations was carried out with the MARC data for four representative times of deposition with an agricultural practice adopted in the UK. In this report the differences are presented which are observed in the potential doses due to ingestion, the areas affected by food-bans and the late health effects when using both models and taking the influence of seasonal effects into account. (orig.) [de

  10. A few seconds to have an accident, a long time to recover: consequences for road accident victims from the ESPARR cohort 2 years after the accident.

    Science.gov (United States)

    Tournier, Charlène; Charnay, Pierrette; Tardy, Hélène; Chossegros, Laetitia; Carnis, Laurent; Hours, Martine

    2014-11-01

    The aim of the present study was to describe the consequences of a road accident in adults, taking account of the type of road user, and to determine predictive factors for consequences at 2 years. Prospective follow-up study. The cohort was composed of 1168 victims of road traffic accidents, aged ≥16 years. Two years after the accident, 912 victims completed a self-administered questionnaire. Weighted logistic regression models were implemented to compare casualties still reporting impact related to the accident versus those reporting no residual impact. Five outcomes were analysed: unrecovered health status, impact on occupation or studies, on familial or affective life, on leisure or sport activities and but also the financial difficulties related to the accident. 46.1% of respondents were motorised four-wheel users, 29.6% motorised two-wheel (including quad) users, 13.3% pedestrians (including inline skate and push scooter users) and 11.1% cyclists. 53.3% reported unrecovered health status, 32.0% persisting impact on occupation or studies, 25.2% on familial or affective life, 46.9% on leisure or sport activities and 20.2% still had accident-related financial difficulties. Type of user, adjusted on age and gender, was linked to unrecovered health status and to impact on leisure or sport activities. When global severity (as measured by NISS) was integrated in the previous model, type of user was also associated with impact on occupation or studies. Type of user was further associated with impact on occupation or studies and on leisure or sport activities when global severity and the sociodemographic data obtained at inclusion were taken into account. It was not, however, related to any of the outcomes studied here, when the models focused on the injured body region. Finally, type of road user did not seem, on the various predictive models, to be related to financial difficulties due to the accident or to impact on familial or affective life. Overall, victims

  11. RADIOLOGICAL AND MEDICAL CONSEQUENCES OF THE CHERNOBYL ACCIDENT

    Directory of Open Access Journals (Sweden)

    V. G. Bebeshko

    2012-01-01

    Full Text Available From the position of a 25-years’ experience to overcome the health effects of Chernobyl the dynamics of the radiation environment, the first summarizing at the international level (1988, the results of completed research and practical monitoring are analyzed. Cohort of acute radiation syndrome (ARS survivors under medical observation at the S.I. "Research Center for Radiation Medicine of the National Academy of Medical Sciences of Ukraine" is the largest. Within the 25 years the functional state of the major organs and body systems, and metabolic homeostasis for this category of persons were studied, a comprehensive assessment of their health, mental and physical performance were given, and risk factors and peculiarities of stochastic and non-stochastic pathology courses were identified, as well as a system of rehabilitation patients after ARS was developed. ARS survivors are suffering from chronic diseases of internal organs and systems (from 5-7 to 10-12 diagnoses at the same time. A correlation between acute radiation effects and specific HLA phenotypes were revealed. The dynamics of the immune system recovery after irradiation was studied. The role and prognostic value of telomere length and programmed cell death of lymphocytes in the formation of the cellular effects of ionizing radiation were determined for the first time. Differences between spontaneous and radiation-induced acute myeloid leukemias were found. Dose-dependent neuropsychiatric, neurophysiological, neuropsychological and neuroimaging deviations were identified after irradiation at doses above 0.3 Sv. It was shown that the lymphocytes of Chernobyl clean-up workers with doses 350 – 690 mGy can induce "the bystander effect" in the non-irradiated cells even after 19 years after exposure. The rates of cancer incidence and mortality of victims, the lessons and key problems to be solved in the third decade after the Chernobyl accident are considered.

  12. Estimates of the financial consequences of nuclear-power-reactor accidents

    International Nuclear Information System (INIS)

    Strip, D.R.

    1982-09-01

    This report develops preliminary techniques for estimating the financial consequences of potential nuclear power reactor accidents. Offsite cost estimates are based on CRAC2 calculations. Costs are assigned to health effects as well as property damage. Onsite costs are estimated for worker health effects, replacement power, and cleanup costs. Several classes of costs are not included, such as indirect costs, socio-economic costs, and health care costs. Present value discounting is explained and then used to calculate the life cycle cost of the risks of potential reactor accidents. Results of the financial consequence estimates for 156 reactor-site combinations are summarized, and detailed estimates are provided in an appendix. The results indicate that, in general, onsite costs dominate the consequences of potential accidents

  13. The importance of trajectory modelling in accident consequence assessments

    International Nuclear Information System (INIS)

    Jones, J.A.; Williams, J.A.; Hill, M.D.

    1988-01-01

    Most atmospheric dispersion models used at present or probabilistic risk assessment (PRA) are linear: they take account of the wind speed but not the direction after the first hour. Therefore, the trajectory model is a more realistic description of the cloud's behaviour. However, the extra complexity means that the computing costs increase. This is an important factor for the MARIA code which is intended to be run on computers of varying power. The numbers of early effects predicted by a linear model and a trajectory model in a probabilistic risk assessment were compared to see which model should be preferred. The trajectory model predicted about 25% fewer expected early deaths and 30% more people evacuated than the linear model. However, the trajectory model took about ten times longer to calculate its results. The choice between the two models may depend on the speed of the computer available

  14. Economic consequences of major accidents in the industrial plants: The case of a nuclear power plant

    International Nuclear Information System (INIS)

    Fraix, J.

    1989-09-01

    These last years, newspapers head-lines have reported various accidents (Mexico City, Bhopal, Chernobyl, ...) which have drawn attention to the fact that the major technological risk is now a reality and that, undoubtedly, industrial decision-makers ought to integrate it into their preoccupations. In addition to the sometimes considerable human problems such accidents engender, their economic consequences may be such that they become significant on a national or even international scale. The aim of the present paper is to analyse these economic effects by using the particular context of a nuclear power plant. The author has deliberately limited his subject to the consequences of a major accident, that is to say a sudden event, theoretically unforeseen and beyond man's control. The qualification major means an accident of which the consequences extend far beyond the industrial plant itself. The direct and indirect economic consequences are analysed from the responsibility point of view as well as from the national and international community's point of view. A paragraph explains how the coverage of the costs can rely on the cooperation of a number of parties: responsible company, state, insurers, customers, etc. The study is broadly based on the experience resulting from the two major accidents which happened in the nuclear industry these last years (Three Mile Island in 1979 and Chernobyl in 1986) and makes use of more theoretical considerations, for example in the field of the economic evaluation of human life. (author). 58 refs, 2 figs, 12 tabs

  15. Chernobyl accident: causes and consequences (expert conclusion). Part 3. Chernobyl accident effect on Belarus

    International Nuclear Information System (INIS)

    Nesterenko, V.B.

    1992-01-01

    Expert conclusion is presented on the Chernobyl accident effect on Belarus. Problems of ground and food contamination, medical and biological radiation effects on the population are considered. Attention is paid to the radiation monitoring and radiometric gages. Scale of the damage for forestry and agriculture is described and recommendations on the agriculture is described and recommendations on the agricultural production and forest utilization at contaminated areas are given. 24 refs.; 4 figs.; 24 tabs

  16. Simulation of small break loss of coolant accident using relap 5/ MOD 2 computer code

    International Nuclear Information System (INIS)

    Megahed, M.M.

    1992-01-01

    An assessment of relap 5 / MOD 2/Cycle 36.05 best estimate computer code capabilities in predicting the thermohydraulic response of a PWR following a small break loss of coolant accident is presented. The experimental data base for the evaluation is the results of Test S-N H-3 performed in the semi scale MOD-2 c Test facility which modeled a 0.5% small break loss of coolant accident with an accompanying failure of the high pressure injection emergency core cooling system. A conclusion was reached that the code is capable of making small break loss of coolant accident calculations efficiently. However, some of the small break loss of coolant accident related phenomena were not properly predicted by the code, suggesting a need for code improvement.9 fig., 3 tab

  17. A thermo mechanical benchmark calculation of a hexagonal can in the BTI accident with INCA code

    International Nuclear Information System (INIS)

    Zucchini, A.

    1988-01-01

    The thermomechanical behaviour of an hexagonal can in a benchmark problem (simulating the conditions of a BTI accident in a fuel assembly) is examined by means of the INCA code and the results systematically compared with those of ADINA

  18. Simulation of severe accident using March-3 computer code

    International Nuclear Information System (INIS)

    Fernandes, A.; Nakata, H.

    1991-01-01

    The severe accident sensitivity analysis utilizing the March-3 approximate modelization options has been performed. The reference results against which the present results have been compared were obtained from the best published results for the most representative accident sequences: TMLU, S sub(2)DC sub(r) and S sub(2)DCF sub(r) for the Zion-1 reactor. The results of the present sensitivity analysis revealed the presence of very crude modelizations, in the March-3 program, to represent the critical phenomenologies involved in the severe accident sequences considered, even though large uncertainties must still be taken into account due primarily to the scarcity of the integral benchmark data. (author)

  19. Information on the Chernobyl NPP accident and its consequencies prepared for IAEA

    Energy Technology Data Exchange (ETDEWEB)

    1986-11-01

    The information on the accident at the 4th power unit of the Chernobyl NPP and its consequences prepared for IAEA on the basis of the conclusions made by the Government commission constituted for investigating the accident causes and implementing the necessary emergency and reconstruction measures is given. The accident with reactor core disruption and partial destruction of the building Lappened on 26.04.86 at 1 hour and 23 minutes. The accident occurred before reactor shut-down for planned repairs during the testing of one of turbogenerators. The design features of the RBMK-1000 reactor plant, its main physical characteristics and parameters of the NPP safety system are considered. The chronology of the accident development and the results of analysis carried out using a mathematical model are given. The causes of the accident are analyzed. The measures for preventing the accident development and lessening its consequences as well as those for the environment radioactive contamination control and sanitary provisions are described in detail. The conclusion is made that the original cause of the accident is highly improbable combination of disorder and errors in operational conditions made by the personnel of the power unit. It is emphasized that development of the world nuclear engineering, besides advantages in the field of power supply and natural resources conservation, incurs also damages of international character. Among these are transboundary radioactivity transport, in particular, during serious radiation accidents and the danger of international terrorism and specific radiation hazard of nuclear objects under war conditions. All this defines the key necessity of deep international cooperation in the field of nuclear power engineering and its safeguarding.

  20. Minimization of the occupational doses during the liquidation of the radiation accident consequences

    International Nuclear Information System (INIS)

    Kuryndina, Lidia; Stroganov, Anatoly; Kuryndin, Anton

    2008-01-01

    Full text: As known the accident on the Chernobylskaya npp is the heaviest one in the nuclear energy history. It showed how considerable can be radiation levels on the breakdown nuclear facility. Nevertheless Russian specialists on radiation protection worked out and successfully realized a conception of the working in such conditions during the liquidation of the accident consequences. The conception based out on using ALARA principle, included the methods of radiation fields structure analysis and allowed to minimize of the occupational doses at operations of the accident consequences liquidation. The main idea of the conception is in strongly dependence between the radiation dose of the personnel performing the liquidation operations and concrete sequence of these operations. Also it is necessary from time to time to receive the experimental information about radiation situation dynamics on the breakdown facility and to make variant calculations for optimizing for the successful implementation of such approach. The structure of these calculations includes variable fraction for the actual state of the facility before the accident and after one and not variable fraction depend on the geometric and protection characteristics of the facility. And the second part is more complicated and bigger. Therefore the most part of these calculations required for the any successful liquidation of the accident consequences can be made on the facility projecting stage. If it will be made the following tasks can be solved in case of the accident: 1) To estimate a distribution of the contamination source using the radiation control system indications; 2) To determine a contribution from each source to the dose rate for any contaminated area; 3) To estimate the radiation doses of the personnel participated in the accident consequences liquidation; 4) To select and to realize the sequence of the liquidation operations giving the minimal doses. The paper will overview the description

  1. EDPUFF- a Gaussian dispersion code for consequence analysis

    International Nuclear Information System (INIS)

    Oza, R.B.; Bapat, V.N.; Nair, R.N.; Hukkoo, R.K.; Krishnamoorthy, T.M.

    1995-01-01

    EDPUFF- Equi Distance Puff is a Gaussian dispersion code in FORTRAN language to model atmospheric dispersion of instantaneous or continuous point source releases. It is designed to incorporate the effect of changing meteorological conditions and source release rates on the spatial distribution profiles and its consequences. Effects of variation of parameters like puff spacing, puff packing, averaging schemes are discussed and the choice of the best values for minimum errors and minimum computer CPU time are identified. The code calculates the doses to individual receptors as well as average doses for population zones from internal and external routes over the area of interest. Internal dose computations are made for inhalation and ingestion pathways while the doses from external route consists of cloud doses and doses from surface deposited activity. It computes inhalation and ingestion dose (milk route only) for critical group (1 yr old child). In case of population zones it finds out maximum possible doses in a given area along with the average doses discussed above. Report gives the doses from various pathways for unit release of fixed duration. (author). 7 refs., figs., 7 appendixes

  2. Assessing the consequences in a nuclear accident scenario at Cernavoda NPP

    International Nuclear Information System (INIS)

    Margeanu, Sorin; Angelescu, Tatiana

    2004-01-01

    Having in view a possible nuclear incident, considerable planning is necessary to reduce at manageable levels the types of decisions leading to effective responses concerning the public protection. One of the most important parts of an emergency response plan is the computerized system which allows to predict the radiological impact of the accident and to provide information in a manageable and effective form for evaluating alternative countermeasure strategies in the various stages of the accident. In this paper the PC-COSYMA results for early containment failure of a CANDU reactor are presented. The deterministic health effects arising in nuclear accident situation are also presented. As source term we have used the core inventory obtained with ORIGEN computer code. The essential input parameters for PC-COSYMA computer code are also done. (authors)

  3. Cohort formation for epidemiological study of medical consequences of the Chernobyl accident

    International Nuclear Information System (INIS)

    Rozhko, A.V.; Masyakin, V.B.; Vlasova, N.G.

    2008-01-01

    Belarus State Registry of the Chernobyl-affected population contains information about 276 000 residents of the Republic of Belarus exposed due to the Chernobyl NPP accident. Evidently, the population who lived in the evacuation zone was exposed mostly to radiation and also people participating in the liquidation of the Chernobyl accident consequences (emergency workers) within this zone in early post accident period of the catastrophe. Taking into account this criterion, we singled out the group out of all data files including all people who stayed in the evacuation zone not later than on May 31, 1986. The total number of the group made up 39 548 people including 4251 people who were under 18 at the moment of the accident. By preliminary estimation the number of person-years taking into account the deceased and left out of observation made up at the beginning of 2007- 735 600. During the period since 1986 there was detected 2671 cases of malignant tumors in the cohort and among people who were children and adolescents in 1986 there was registered 106 cases of malignant tumors (82% -thyroid cancer). Among 7483 of the deceased, malignant tumors is the cause of death at 1260 people. At present the real number of alive and remained subjects under observation makes up 25359 people including 2321 people who were under 18 at the moment of the accident. This group will form the base for further prospective research aiming at assessment of medical consequences of the Chernobyl NPP accident. (author)

  4. Formation of decontamination cost calculation model for severe accident consequence assessment

    International Nuclear Information System (INIS)

    Silva, Kampanart; Promping, Jiraporn; Okamoto, Koji; Ishiwatari, Yuki

    2014-01-01

    In previous studies, the authors developed an index “cost per severe accident” to perform a severe accident consequence assessment that can cover various kinds of accident consequences, namely health effects, economic, social and environmental impacts. Though decontamination cost was identified as a major component, it was taken into account using simple and conservative assumptions, which make it difficult to have further discussions. The decontamination cost calculation model was therefore reconsidered. 99 parameters were selected to take into account all decontamination-related issues, and the decontamination cost calculation model was formed. The distributions of all parameters were determined. A sensitivity analysis using the Morris method was performed in order to identify important parameters that have large influence on the cost per severe accident and large extent of interactions with other parameters. We identified 25 important parameters, and fixed most negligible parameters to the median of their distributions to form a simplified decontamination cost calculation model. Calculations of cost per severe accident with the full model (all parameters distributed), and with the simplified model were performed and compared. The differences of the cost per severe accident and its components were not significant, which ensure the validity of the simplified model. The simplified model is used to perform a full scope calculation of the cost per severe accident and compared with the previous study. The decontamination cost increased its importance significantly. (author)

  5. Critical analysis of accident scenario and consequences modelling applied to light-water reactor power plants for accident categories beyond the design basis accident (DBA)

    International Nuclear Information System (INIS)

    Brofferio, C.; Cagnetti, P.; Ferrara, V.; Manilia, E.; Pietrangeli, G.; Sennis, C.

    1985-01-01

    A critical analysis and sensitivity study of the modelling of accident scenarios and environmental consequences are presented, for light-water reactor accident categories beyond the standard design-basis-accident category. The first chapter, on ''source term'' deals with the release of fission products from a damaged core inventory and their migration within the primary circuit and the reactor containment. Particular attention is given to the influence of engineering safeguards intervention and of the chemical forms of the released fission products. The second chapter deals with their release to the atmosphere, transport and wet or dry deposition, outlining relevant partial effects and confronting short-duration or prolonged releases. The third chapter presents a variability analysis, for environmental contamination levels, for two extreme hypothetical scenarios, evidencing the importance of plume rise. A numerical plume rise model is outlined

  6. 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.

  7. ILK statement on the consequences of the Chernobyl accident. Taking stock after twenty years

    International Nuclear Information System (INIS)

    2006-01-01

    The Chernobyl reactor accident was the consequence of a reactor design which was not inherently safe, and of a lack of 'safety culture'. The RBMK-type reactor (a Russian graphite-moderated light water reactor design: reaktor bolshoi moshnosty kanalny=high-power channel reactor) had not been designed to a satisfactory safety level, and the operating staff were not informed on the weak spots in plant design. The combination of these factors caused the worst nuclear accident, completely destroying the reactor. The consequences may be seen as the product of two severe accidents superimposed upon each other: the explosion of the reactor, and core melt-down associated with an intense, persistent fire of the graphite moderator. The Statement contains analyses of these points: Release, Propagation and Deposition of Radioactive Materials; Protective Measures; Impact on the Environment and Agriculture; Assessment of Radiation Exposure; Health Impact; Psychological and Societal Impacts; Potential Residual Risks. (orig.)

  8. In Vienna about Chernobyl. Summing up the consequences of the accident

    International Nuclear Information System (INIS)

    Latek, S.

    1996-01-01

    The Joint EC/IAEA/WHO International Conference ''One Decade after Chernobyl: Summing up the consequences of the accident'' has been held in Vienna, 8-12 April 1996. The most important subjects of the conference was: assessment of total releases and deposits, radiation doses, clinically observed effects, thyroid effects, longer term health effects, psychological and environmental consequences, social economic, institutional and political impact, nuclear safety, sarcophagus, perspective and prognosis

  9. Frozen Accident Pushing 50: Stereochemistry, Expansion, and Chance in the Evolution of the Genetic Code.

    Science.gov (United States)

    Koonin, Eugene V

    2017-05-23

    Nearly 50 years ago, Francis Crick propounded the frozen accident scenario for the evolution of the genetic code along with the hypothesis that the early translation system consisted primarily of RNA. Under the frozen accident perspective, the code is universal among modern life forms because any change in codon assignment would be highly deleterious. The frozen accident can be considered the default theory of code evolution because it does not imply any specific interactions between amino acids and the cognate codons or anticodons, or any particular properties of the code. The subsequent 49 years of code studies have elucidated notable features of the standard code, such as high robustness to errors, but failed to develop a compelling explanation for codon assignments. In particular, stereochemical affinity between amino acids and the cognate codons or anticodons does not seem to account for the origin and evolution of the code. Here, I expand Crick's hypothesis on RNA-only translation system by presenting evidence that this early translation already attained high fidelity that allowed protein evolution. I outline an experimentally testable scenario for the evolution of the code that combines a distinct version of the stereochemical hypothesis, in which amino acids are recognized via unique sites in the tertiary structure of proto-tRNAs, rather than by anticodons, expansion of the code via proto-tRNA duplication, and the frozen accident.

  10. On the assessment of adverse consequences of Chernobyl accident

    International Nuclear Information System (INIS)

    Burlakova, E.B.

    2007-01-01

    of the damages. With this kind of low-level irradiation, the reparative systems either are not initiated (induced), or function inadequately, or are initiated with a delay, i.e., when the exposed object has already received radiation damages. Recently, the absence of reparation at low irradiation doses was verified on the cell level, and the complex character of the dose dependence was confirmed. Previously, we published a similar scheme of dependence of damages on irradiation dose, which was different for different dose ranges. According to the scheme, the quantitative characteristics were similar for the doses that differed by several orders of magnitude; in a certain dose range, the effect may have an opposite sign.The results obtained and supported by numerous experiments are important because the above dose dependences made it possible to come to conclusion about a radiogenic or non-radiogenic character of changes observed in an irradiated organism. The indisputable conclusion that if the effect increases with the dose it is evidence for its radiogenic nature is by no means in favor of an opposite statement, i.e., that the absence of a direct dose-effect dependence but its nonmonotonic character is evidence for the absence of a relation of the effect to irradiation. The controversial conclusions of International and Russian organizations stem mainly from the underestimation and misunderstanding of the effects of ultra-low and low irradiation doses, reluctance to apply other criteria to assess the consequences of irradiation on human health, and conviction (groundless) that low doses cause either no damages or such minor damages that they may be neglected and disregarded. In the lecture, data that elucidate the above controversies will be presented.

  11. Prevention of the causes and consequences of a criticality accident - measures adopted in France; Prevention des causes et des consequences d'un accident de criticite - solutions adoptees en France

    Energy Technology Data Exchange (ETDEWEB)

    Fruchard, Y.; Lavie, J.M

    1966-07-01

    The question of safety in regard to criticality accident risks has two aspects: prevention of the cause and limitation of the consequences. These two aspects are closely connected. The effort devoted to prevention of the causes depends on the seriousness of the possible human psychologic and economic consequences of the accident. The criticality accidents which have occurred in the nuclear industry, though few in number, do reveal the imperfect nature of the techniques adopted to prevent the causes, and also constitute the only available realistic basis for evaluating the consequences and developing measures to limit them. The authors give a analysis of the known causes and consequences of past criticality accidents and on this basis make a number of comments concerning: the validity of traditional safety criteria, the probability of accidents for different types of operations, characteristic accidents which can serve as models, and the extent of possible radiological consequences. The measures adopted in France to limit the consequences of a possible criticality accident under the headings: location, design and lay-out of the installations, accident detection, and dosimetry for the exposed personnel, are briefly described after a short account of the criteria used in deciding on them. (author) [French] La surete relative aux risques d'accidents de criticite presente deux aspects: la prevention des causes et les parades aux consequences. Ces deux aspects sont tres lies. L'effort consenti a la prevention des causes decoule de l'importance des consequences humaines economiques et psychologiques possibles d'un eventuel accident. Les accidents de criticite survenus dans l'industrie nucleaire, malgre leur rarete, d'une part devoilent les imperfections des techniques de prevention des causes, d'autre part constituent la seule base realiste disponible d'evaluation des consequences et de mise au point des parades a ces consequences

  12. Prevention of the causes and consequences of a criticality accident - measures adopted in France; Prevention des causes et des consequences d'un accident de criticite - solutions adoptees en France

    Energy Technology Data Exchange (ETDEWEB)

    Fruchard, Y; Lavie, J M

    1966-07-01

    The question of safety in regard to criticality accident risks has two aspects: prevention of the cause and limitation of the consequences. These two aspects are closely connected. The effort devoted to prevention of the causes depends on the seriousness of the possible human psychologic and economic consequences of the accident. The criticality accidents which have occurred in the nuclear industry, though few in number, do reveal the imperfect nature of the techniques adopted to prevent the causes, and also constitute the only available realistic basis for evaluating the consequences and developing measures to limit them. The authors give a analysis of the known causes and consequences of past criticality accidents and on this basis make a number of comments concerning: the validity of traditional safety criteria, the probability of accidents for different types of operations, characteristic accidents which can serve as models, and the extent of possible radiological consequences. The measures adopted in France to limit the consequences of a possible criticality accident under the headings: location, design and lay-out of the installations, accident detection, and dosimetry for the exposed personnel, are briefly described after a short account of the criteria used in deciding on them. (author) [French] La surete relative aux risques d'accidents de criticite presente deux aspects: la prevention des causes et les parades aux consequences. Ces deux aspects sont tres lies. L'effort consenti a la prevention des causes decoule de l'importance des consequences humaines economiques et psychologiques possibles d'un eventuel accident. Les accidents de criticite survenus dans l'industrie nucleaire, malgre leur rarete, d'une part devoilent les imperfections des techniques de prevention des causes, d'autre part constituent la seule base realiste disponible d'evaluation des consequences et de mise au point des parades a ces consequences. Les auteurs presentent une analyse des

  13. Application of GIS in prediction and assessment system of off-site accident consequence for NPP

    International Nuclear Information System (INIS)

    Wang Xingyu; Shi Zhongqi

    2002-01-01

    The assessment and prediction software system of off-site accident consequence for Guangdong Nuclear Power Plant (GNARD2.0) is a GIS-based software system. The spatial analysis of radioactive materials and doses with geographic information is available in this system. The structure and functions of the GNARD system and the method of applying ArcView GIS are presented

  14. Medical and biological aspects of ionizing radiation influence in consequence with accident at ChNPP

    International Nuclear Information System (INIS)

    Shidlovs'ka, T.A.

    2011-01-01

    This monograph presents the issues on systematic influence of ionizing radiation on the biological systems. The results of personal complex studies because of influence of ionizing radiation in consequence with accident at ChNPP on auditory analyzer, creating voice, cardiovascular system and central nervous system are submitted.

  15. Environmental decision support system on base of geoinformational technologies for the analysis of nuclear accident consequences

    International Nuclear Information System (INIS)

    Haas, T.C.; Maigan, M.; Arutyunyan, R.V.; Bolshov, L.A.; Demianov, V.V.

    1996-01-01

    The report deals with description of the concept and prototype of environmental decision support system (EDSS) for the analysis of late off-site consequences of severe nuclear accidents and analysis, processing and presentation of spatially distributed radioecological data. General description of the available software, use of modem achievements of geostatistics and stochastic simulations for the analysis of spatial data are presented and discussed

  16. Users guide for NRC145-2 accident assessment computer code

    International Nuclear Information System (INIS)

    Pendergast, M.M.

    1982-08-01

    An accident assessment computer code has been developed for use at the Savannah River Plant. This computer code is based upon NRC Regulatory Guide 1.145 which provides guidence for accident assessements for power reactors. The code contains many options so that the user may utilize the code for many different assessments. For example the code can be used for non-nuclear assessments such as Sulpher Dioxide which may be required by the EPA. A discription of the code is contained in DP-1646. This document is a compilation of step-by-step instructions on how to use the code on the SRP IBM 3308 computer. This document consists of a number of tables which contain copies of computer listings. Some of the computer listings are copies of input; other listings give examples of computer output

  17. Consequences in the pumps operation during a large loss of coolant accident

    International Nuclear Information System (INIS)

    Santos, G.A. dos; Sabundjian, G.

    1991-08-01

    The event of living on or turning off the operation of the Reactor Cooling Pumps - RCPs, in the case of a Loss of Coolant Accident - LOCA, has been a reason of a lot of studies after the Three Mile Island 2 accident. Thus, it was investigated a large break LOCA in the cold leg of Angra 1, with the RELAP4/MOD5 Code during the blowdown. The attained results indicated that the best performance of the core was in the case where the RCPs had been turned off in the beginning of the transient, when compared with different operation conditions of the RCPs. (author)

  18. ASTEC V2 severe accident integral code: Fission product modelling and validation

    International Nuclear Information System (INIS)

    Cantrel, L.; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

    2014-01-01

    One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology. ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase. In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R and D knowledge. This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix

  19. The Fukushima Daiichi Accident. Technical Volume 4/5. Radiological Consequences. Annexes

    International Nuclear Information System (INIS)

    2015-08-01

    The Fukushima Daiichi Accident consists of a Report by the IAEA Director General and five technical volumes. It is the result of an extensive international collaborative effort involving five working groups with about 180 experts from 42 Member States with and without nuclear power programmes and several international bodies. It provides a description of the accident and its causes, evolution and consequences, based on the evaluation of data and information from a large number of sources available at the time of writing. The Fukushima Daiichi Accident will be of use to national authorities, international organizations, nuclear regulatory bodies, nuclear power plant operating organizations, designers of nuclear facilities and other experts in matters relating to nuclear power, as well as the wider public. The set contains six printed parts and five supplementary CD-ROMs. Contents: Report by the Director General; Technical Volume 1/5, Description and Context of the Accident; Technical Volume 2/5, Safety Assessment; Technical Volume 3/5, Emergency Preparedness and Response; Technical Volume 4/5, Radiological Consequences; Technical Volume 5/5, Post-accident Recovery; Annexes. The Report by the Director General is available separately in Arabic, Chinese, English, French, Russian, Spanish and Japanese

  20. The accident at the Chernobyl' nuclear power plant and its consequences

    International Nuclear Information System (INIS)

    1986-08-01

    The material is taken from the conclusions of the Government Commission on the causes of the accident at the fourth unit of the Chernobyl' nuclear power plant and was prepared by a team of experts appointed by the USSR State Committee on the Utilization of Atomic Energy. It contains general material describing the accident, its causes, the action taken to contain the accident and to alleviate its consequences, the radioactive contamination and health of the population and some recommendations for improving nuclear power safety. 7 annexes are devoted to the following topics: water-graphite channel reactors and operating experience with RBMK reactors, design of the reactor plant, elimination of the consequences of the accident and decontamination, estimate of the amount, composition and dynamics of the discharge of radioactive substances from the damaged reactor, atmospheric transport and radioactive contamination of the atmosphere and of the ground, expert evaluation and prediction of the radioecological state of the environment in the area of the radiation plume from the Chernobyl' nuclear power station, medical-biological problems. A separate abstract was prepared for each of these annexes. The slides presented at the post-accident review meeting are grouped in two separate volumes

  1. Code package {open_quotes}SVECHA{close_quotes}: Modeling of core degradation phenomena at severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Veshchunov, M.S.; Kisselev, A.E.; Palagin, A.V. [Nuclear Safety Institute, Moscow (Russian Federation)] [and others

    1995-09-01

    The code package SVECHA for the modeling of in-vessel core degradation (CD) phenomena in severe accidents is being developed in the Nuclear Safety Institute, Russian Academy of Science (NSI RAS). The code package presents a detailed mechanistic description of the phenomenology of severe accidents in a reactor core. The modules of the package were developed and validated on separate effect test data. These modules were then successfully implemented in the ICARE2 code and validated against a wide range of integral tests. Validation results have shown good agreement with separate effect tests data and with the integral tests CORA-W1/W2, CORA-13, PHEBUS-B9+.

  2. The assessment of containment codes by experiments simulating severe accident scenarios

    International Nuclear Information System (INIS)

    Karwat, H.

    1992-01-01

    Hitherto, a generally applicable validation matrix for codes simulating the containment behaviour under severe accident conditions did not exist. Past code applications have shown that most problems may be traced back to inaccurate thermalhydraulic parameters governing gas- or aerosol-distribution events. A provisional code-validation matrix is proposed, based on a careful selection of containment experiments performed during recent years in relevant test facilities under various operating conditions. The matrix focuses on the thermalhydraulic aspects of the containment behaviour after severe accidents as a first important step. It may be supplemented in the future by additional suitable tests

  3. Effects of rainstorms and runoff on consequences of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Brown, W.D.; Wayland, J.R.

    1976-10-01

    A preliminary model describing the effects of washout and runoff on the consequences of a nuclear reactor accident is presented. The most important new feature of this stratified model relative to the model in WASH-1400 is the spatial structure of rainstorms and runoff consisting of four levels of rain activity that are normalized by rain gage data. The predicted concentrations of radioactivity and resultant health consequences of the stratified model are compared with those of the model in WASH-1400 for simplified rainstorms with fixed meteorological conditions, an actual rainstorm, and a stratified sample run consisting of 91 separate reactor accidents. In the case of individual storms, runoff and the spatial structure of the rain in the new model can result in health consequences that are significantly different from those of the WASH-1400 model. The differences between the predictions of the two models are small for the stratified sample run

  4. Consequences to health of the Chernobyl accident; Helbredsmaessige konsekvenser af reaktorulykken i Tjernobyl

    Energy Technology Data Exchange (ETDEWEB)

    Sewerin, I. [Royal Dental College, Dept. of Radiology, Copenhagen (Denmark)

    2001-07-01

    The Chernobyl accident in 1986 has been and still is the subject of great interest. Journalistic reports often contain exaggerations and undocumented statements and much uncertainty about the true consequences of the accident prevails in the population. This article reviews the current literature with the focus on reports from official commissions and documentation in the form of controlled studies. The fatal deterministic consequences comprise about 30 victims. The most important outcome is a marked increase in the incidence of thyroid cancer in children and adolescents in the most heavily contaminated area. Furthermore, pronounced psychosocial problems are dominant in the population of the contaminated area. Other significant and documented health consequences are not seen. (au)

  5. CEC workshop on methods for assessing the offsite radiological consequences of nuclear accidents

    International Nuclear Information System (INIS)

    Luykx, F.; Sinnaeve, J.

    1986-01-01

    On Apr 15-19, 1985, in Luxembourg, the Commission of the European Communities (CEC), in collaboration with the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, and the National Radiological Protection Board (NRPB), United Kingdom, presented a workshop on methods for assessing the offsite radiological consequences of nuclear accidents. The program consisted of eight sessions. The main conclusions, which were presented in the Round Table Session by the individual Session Chairmen, are summarized. Session topics are as follows: Session I: international developments in the field of accident consequence assessment (ACA); Session II: atmospheric dispersion; Session III: food chain models; Session IV: urban contamination; Session V: demographic and land use data; Session VI: dosimetry, health effects, economic and counter measure models; Session VII: uncertainty analysis; and Session VIII: application of probabilistic consequence models as decision aids

  6. Impact of rainstorm and runoff modeling on predicted consequences of atmospheric releases from nuclear reactor accidents

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Brown, W.D.; Wayland, J.R.

    1980-05-01

    A general temperate latitude cyclonic rainstorm model is presented which describes the effects of washout and runoff on consequences of atmospheric releases of radioactive material from potential nuclear reactor accidents. The model treats the temporal and spatial variability of precipitation processes. Predicted air and ground concentrations of radioactive material and resultant health consequences for the new model are compared to those of the original WASH-1400 model under invariant meteorological conditions and for realistic weather events using observed meteorological sequences. For a specific accident under a particular set of meteorological conditions, the new model can give significantly different results from those predicted by the WASH-1400 model, but the aggregate consequences produced for a large number of meteorological conditions are similar

  7. 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)

  8. Proceedings of the first international conference 'The radiological consequences of the Chernobyl accident'

    Energy Technology Data Exchange (ETDEWEB)

    Karaoglou, A; Desmet, G; Kelly, G N; Menzel, H G [European Commission, Brussels (Belgium)

    1996-07-01

    Five main objectives were assigned to the EC/CIS scientific collaborative programme: improvement of the knowledge of the relationship between doses and radiation-induced health effects; updating of the arrangements for off-site emergency management response (shot- and medium term)in the even of a future nuclear accident; assisting the relevant CIS Ministries alleviate the consequences of the Chernobyl accident, in particular in the field of restoration of contaminated territories; elaboration of a scientific basis to definite the content of Community assistance programmes; updating of the local technical infrastructure, and implementation of a large programme of exchange of scientists between both Communities. The topics addressed during the Conference mainly reflect the content of the joint collaborative programme: environmental transfer and decontamination, risk assessment and management, health related issues including dosimetry. The main aims of the Conference are to present the major achievements of the joint EC/CIS collaborative research programme (1992-1995) of the consequences of the Chernobyl accident, and to promote an objective evaluation of them by the international scientific community. The Conference is taking place close to the 10{sup th} anniversary of the accident and we hope it will contribute to more objective communication of the health and environmental consequences of the Chernobyl accident, and how these may be mitigated in future. The Conference is expected to be an important milestone in the series of meetings which will take place internationally around the 10{sup th} anniversary of the nuclear accident. It also provides a major opportunity for all participants to become acquainted with software developed within the framework of the collaborative programme, namely: Geographical Information Systems displaying contamination levels and dose-commitments; Decision Support Systems for the management of contaminated territories; Decision Support

  9. Proceedings of the first international conference 'The radiological consequences of the Chernobyl accident'

    International Nuclear Information System (INIS)

    Karaoglou, A.; Desmet, G.; Kelly, G.N.; Menzel, H.G.

    1996-01-01

    Five main objectives were assigned to the EC/CIS scientific collaborative programme: improvement of the knowledge of the relationship between doses and radiation-induced health effects; updating of the arrangements for off-site emergency management response (shot- and medium term)in the even of a future nuclear accident; assisting the relevant CIS Ministries alleviate the consequences of the Chernobyl accident, in particular in the field of restoration of contaminated territories; elaboration of a scientific basis to definite the content of Community assistance programmes; updating of the local technical infrastructure, and implementation of a large programme of exchange of scientists between both Communities. The topics addressed during the Conference mainly reflect the content of the joint collaborative programme: environmental transfer and decontamination, risk assessment and management, health related issues including dosimetry. The main aims of the Conference are to present the major achievements of the joint EC/CIS collaborative research programme (1992-1995) of the consequences of the Chernobyl accident, and to promote an objective evaluation of them by the international scientific community. The Conference is taking place close to the 10 th anniversary of the accident and we hope it will contribute to more objective communication of the health and environmental consequences of the Chernobyl accident, and how these may be mitigated in future. The Conference is expected to be an important milestone in the series of meetings which will take place internationally around the 10 th anniversary of the nuclear accident. It also provides a major opportunity for all participants to become acquainted with software developed within the framework of the collaborative programme, namely: Geographical Information Systems displaying contamination levels and dose-commitments; Decision Support Systems for the management of contaminated territories; Decision Support Systems

  10. The accident at the Chernobyl' nuclear power plant and its consequences. Pt. 1. General material

    International Nuclear Information System (INIS)

    1986-01-01

    The report contains a presentation of the Chernobyl' nuclear power station and of the RBMK-1000 reactor, including its principal physical characteristics, the safety systems and a description of the site and of the surrounding region. After a chronological account of the events which led to the accident and an analysis of the accident using a mathematical model it is concluded that the prime cause of the accident was an extremely improbable combination of violations of instructions and operating rules committed by the staff of the unit. Technical and organizational measures for improving the safety of nuclear power plants with RBMK reactors have been taken. A detailed description of the actions taken to contain the accident and to alleviate its consequences is given and includes the fire fighting at the nuclear power station, the evaluation of the state of the fuel after the accident, the actions taken to limit the consequences of the accident in the core, the measures taken at units 1, 2 and 3 of the nuclear power station, the monitoring and diagnosis of the state of the damaged unit, the decontamination of the site and of the 30 km zone and the long-term entombment of the damaged unit. The measures taken for environmental radioactive contamination monitoring, starting by the assessment of the quantity, composition and dynamics of fission products release from the damaged reactor are described, including the main characteristics of the radioactive contamination of the atmosphere and of the ground, the possible ecological consequences and data on the exposure of plant and emergency service personnel and of the population in the 30 km zone around the plant. The last part of the report presents some recommendations for improving nuclear power safety, including scientific, technical and organizational aspects and international measures. Finally, an overview of the development of nuclear power in the USSR is given

  11. Joint CEC/OECD(NEA) workshop on recent advances in reactor accident consequence assessment

    International Nuclear Information System (INIS)

    Olast, M.; Sinnaeve, J.

    1988-01-01

    The workshop on probabilistic accident consequence assessment techniques and their applications aims at a review of the present knowledge of all the work in this field. This includes the atmospheric dispersion and deposition modelling, with comparison of the different approaches, the exposure pathways with emphasis on post-deposition processes, the health effects with emphasis on the consequences of the Hiroshima and Nagasaki data re-evaluation, the countermeasures and their economic consequences, the uncertainty analysis of the models and finally the applications of these models as aids to decision making

  12. Radioecological consequences of a potential accident during transport of spent nuclear fuel along an Arctic coastline

    International Nuclear Information System (INIS)

    Iosjpe, M.; Reistad, O.; Amundsen, I.B.

    2009-01-01

    This article presents results pertaining to a risk assessment of the potential consequences of a hypothetical accident occurring during the transportation by ship of spent nuclear fuel (SNF) along an Arctic coastline. The findings are based on modelling of potential releases of radionuclides, radionuclide transport and uptake in the marine environment. Modelling work has been done using a revised box model developed at the Norwegian Radiation Protection Authority. Evaluation of the radioecological consequences of a potential accident in the southern part of the Norwegian Current has been made on the basis of calculated collective dose to man, individual doses for the critical group, concentrations of radionuclides in seafood and doses to marine organisms. The results of the calculations indicate a large variability in the investigated parameters above mentioned. On the basis of the calculated parameters the maximum total activity ('accepted accident activity') in the ship, when the parameters that describe the consequences after the examined potential accident are still in agreement with the recommendations and criterions for protection of the human population and the environment, has been evaluated

  13. Level of health of cleaners taking part in the Chernobyl accident consequences

    International Nuclear Information System (INIS)

    Margine, L.; Vicol, C.

    2009-01-01

    During the period of 1986-1988 about 3,000 Moldova citizens took part in Chernobyl NPP accident consequences elimination. In this article the level of morbidity, disability and mortality among Chernobyl accident consequences liquidation participants is analyzed. As a result of analysis of medical documentation and statistical data was revealed that the sickness rate among disaster fighters 2,3 times higher than general sickness rate of the population in Moldova. Disability in this category is at average of 73 per cent as opposed to the overall index for the population of Moldova - 4,4%, this means it is 17 times higher. Mortality among the participants of the accident at Chernobyl NPP is 6 times higher of general data. The participants of the breakdown elimination of Chernobyl accident consequences are equal in their right with the participants and invalids of war and with the disabled workers. Medical and social security of this group is regulated by the legislation of the Republic of Moldova

  14. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  15. Thermohydraulic analysis of nuclear power plant accidents by computer codes

    International Nuclear Information System (INIS)

    Petelin, S.; Stritar, A.; Istenic, R.; Gregoric, M.; Jerele, A.; Mavko, B.

    1982-01-01

    RELAP4/MOD6, BRUCH-D-06, CONTEMPT-LT-28, RELAP5/MOD1 and COBRA-4-1 codes were successful y implemented at the CYBER 172 computer in Ljubljana. Input models of NPP Krsko for the first three codes were prepared. Because of the high computer cost only one analysis of double ended guillotine break of the cold leg of NPP Krsko by RELAP4 code has been done. BRUCH code is easier and cheaper for use. Several analysis have been done. Sensitivity study was performed with CONTEMPT-LT-28 for double ended pump suction break. These codes are intended to be used as a basis for independent safety analyses. (author)

  16. Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code

    International Nuclear Information System (INIS)

    Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

    2017-02-01

    Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

  17. Evaluation of the General Atomic codes TAP and RECA for HTGR accident analyses

    International Nuclear Information System (INIS)

    Ball, S.J.; Cleveland, J.C.; Sanders, J.P.

    1978-01-01

    The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, and the susceptibility of the codes to misuse and input errors is discussed. A critique of code verification plans is also included. The several cases where direct comparisons could be made between TAP/RECA calculations and those based on other independently developed codes indicated generally good agreement, thus contributing to the credibility of the codes

  18. Development of a dose assessment computer code for the NPP severe accident at intermediate level - Korean case

    International Nuclear Information System (INIS)

    Cheong, J.H.; Lee, K.J.; Cho, H.Y.; Lim, J.H.

    1993-01-01

    A real-time dose assessment computer code named RADCON (RADiological CONsequence analysis) has been developed. An approximation method describing the distribution of radionuclides in a puff was proposed and implemented in the code. This method is expected to reduce the time required to calculate the cloud shine (external dose from radioactive plumes). RADCON can simulate an NPP emergency situation by considering complex topography and continuous washout phenomena and provide a function of effective emergency planning. To verify the code results, RADCON has been compared with RASCAL, which was developed for the U.S. NRC by ORNL, for eight hypothetical accident scenarios. Sensitivity analysis was also performed for the important input parameters. (2 tabs., 3 figs.)

  19. Levels of endogenous regulatory factors in liquidators of consequences of the Chernobyl accident

    International Nuclear Information System (INIS)

    Liasko, L.I.; Souchkevitch, G.N.; Tsyb, A.F.

    1997-01-01

    Dynamics of endogenous regulatory factor levels was studied in liquidators of consequences of the Chernobyl accident (mean age - 42 years). Irradiation dose for 90% of examined individuals was within 100 mSv range. We observed a decreased level of synthesis of intracellular processes regulators (cAMP, cGMP) and biased ratio of arachidonic acid metabolites (TxB2, 6-Keto-PGF1α) in persons worked in the zone of accident at different time during the period of 1986-1988. The parameters measured were preserved even 4 years later and the changes apparently did not depend on the individual's age and work conditions. However they were most pronounced in liquidators of 1986 and in those who stayed in the Chernobyl accident zone for a long time. There was no evident connection between the dose and extent of the parameter alterations. (author)

  20. Development of information resources package for the Chernobyl accident and its consequences by INIS

    International Nuclear Information System (INIS)

    Negeri, B.; Tolstenkov, A.; Rieder, S.

    2006-01-01

    The Chernobyl accident was a global catastrophe that captured global attention and as such literature on the Chernobyl accident and its consequences is an important subject covered by the International Nuclear Information System (INIS) Database. The INIS Database contains about 21000 bibliographic records and 9000 full text documents on this subject from 1986 up to August 2006. Based on these extensive resources INIS released a DVD that contained bibliographic references and full text documents as well a bibliometric study of the Chernobyl references on the occasion of the International Conference entitled 'Chernobyl: Looking Back to Go Forwards' held in Vienna on 6 and 7 September 2005. Subsequently, INIS decided to release Revision 1 of the DVD in August 2006 for the twentieth anniversary of the Chernobyl accident with additional value added information sources. This paper briefly discusses the bibliometric parameters of the references, the contents of DVD and the activities undertaken to produce the Chernobyl information resources package

  1. Two decades of radiological accidents direct causes, roots causes and consequences

    Directory of Open Access Journals (Sweden)

    Rozental Jose de Julio

    2002-01-01

    Full Text Available Practically all Countries utilize radioisotopes in medicine, industry, agriculture and research. The extent to which ionizing radiation practices are employed varies considerably, depending largely upon social and economic conditions and the level of technical skills available in the country. An overview of the majority of practices and the associated hazards will be found in the Table IV to VII of this document. The practices in normal and abnormal operating conditions should follow the basic principles of radiation protection and the Safety of Radiation Sources, considering the IAEA Radiation Protection and the Safety of Radiation Sources, Safety Series 120 and the IAEA Recommendation of the Basic Safety Standards for Radiation Protection, Safety Series Nº 115. The Standards themselves underline the necessity to be able to predict the radiological consequences of emergency conditions and the investigations that should need to be done. This paper describes the major accidents that had happened in the last two decades, provides a methodology for analyses and gives a collection of lessons learned. This will help the Regulatory Authority to review the reasons of vulnerabilities, and to start a Radiation safety and Security Programme to introduce measurescapable to avoid the recurrence of similar events. Although a number of accidents with fatalities have caught the attention of the public in recent year, a safety record has accompanied the widespread use of radiation sources. However, the fact that accidents are uncommon should not give grounds for complacency. No radiological accident is acceptable. From a radiation safety and security of the sources standpoint, accident investigation is necessary to determine what happened, why, when, where and how it occurred and who was (were involved and responsible. The investigation conclusion is an important process toward alertness and feedback to avoid careless attitudes by improving the comprehension

  2. Consequences of tritium release to water pathways from postulated accidents in a DOE production reactor

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Olson, R.L.; Hamby, D.M.

    1991-01-01

    A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 x 10 -3 person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 x 10 -2 person-Sv/Pbq (2.4 x 10 -4 person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This study suggests that application of regional models (1610 km radius) will indicate larger dose consequences from short-term tritium release to the atmosphere than from comparable tritium source terms to water pathways. However, the water pathways assessment is clearly site-specific, and the overall aqueous dose will be dependent on downstream receptor populations and uses of the river

  3. A commentary on the current status and the future role of the European accident code

    International Nuclear Information System (INIS)

    Butland, A.T.D.

    1990-01-01

    This paper describes the history of the project to produce the European Accident code (EAC), leading to the planned release of a version of EAC-2 at the end of 1989. The requirements of a computer code to model the initiation phase of Hypothetical Core Disruptive Accidents (HCDAs) are discussed, paying particular attention to the lessons learnt in the CABRI project. The current status and content of the EAC-2 code are examined in relation to these requirements, noting how the sophisticated modelling plans for EAC-2 make it a benchmark code. The validation status of EAC-2 and future plans are discussed, noting that currently it consists solely of stand-alone validation of the modules used in EAC-2, rather than validation of the combined code. The future role of EAC-2 is briefly discussed in relation to the fast reactor plans in the EEC countries. (author)

  4. Effect of parameter variation of reactor coolant pump on loss of coolant accident consequence

    International Nuclear Information System (INIS)

    Dang Gaojian; Huang Daishun; Gao Yingxian; He Xiaoqiang

    2015-01-01

    In this paper, the analyses were carried out on Ling'ao nuclear power station phase II to study the consequence of the loss of coolant accident when the homologous characteristic curves and free volumes of the reactor coolant pump changed. Two different pumps used in the analysis were 100D (employed on Ling'ao nuclear power station phase II) and ANDRITZ. The thermal characteristics in the large break LOCA accident were analyzed using CATHRE GB and CONPATE4, and the reactor coolant system hydraulics load during blow-clown phase of LOCA accident was analyzed using ATHIS and FORCET. The calculated results show that the homologous characteristic curves have great effect on the thermal characteristics of reactor core during the reflood phase of the large break LOCA accident. The maximum cladding surface temperatures are quite different when the pump's homologous characteristic curves change. On the other hand, the pump's free volume changing results in the variation of the LOCA rarefaction wave propagation, and therefore, the reactor coolant system hydraulic load in LOCA accident would be different. (authors)

  5. Validation of the metal fuel version of the SAS4A accident analysis code

    International Nuclear Information System (INIS)

    Tentner, A.M.

    1991-01-01

    This paper describes recent work directed towards the validation of the metal fuel version of the SAS4A accident analysis code. The SAS4A code system has been developed at Argonne National Laboratory for the simulation of hypothetical severe accidents in Liquid Metal-Cooled Reactors (LMR), designed to operate in a fast neutron spectrum. SAS4A was initially developed for the analysis of oxide-fueled liquid metal-cooled reactors and has played an important role in the simulation and assessment of the energetics potential for postulated severe accidents in these reactors. Due to the current interest in the metal-fueled liquid metal-cooled reactors, a metal fuel version of the SAS4A accident analysis code is being developed in the Integral Fast Reactor program at Argonne. During such postulated accident scenarios as the unprotected (i.e. without scram) loss-of-flow and transient overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling, and fuel and cladding melting and relocation. Due to strong neutronic feedbacks these events can significantly influence the reactor power history in the accident progression. The paper presents the results of a recent SAS4A simulation of the M7 TREAT experiment. 6 refs., 5 figs

  6. Analysis of eventual accidents in a water experimental loop, using the Relap 4 computer code

    International Nuclear Information System (INIS)

    Fernandes Filho, T.L.

    1981-01-01

    Transients caused by accidents as (1) loss of coolant, (2) failure in the principal pump and (3) power excursions were analysed. In the accident simulation, the Relap 4/Mod 3 computer code was used. The results obtained with the steady state model showed to be consistent with the project-and operation data of the experimental loop. For all the accidents analysed that considered the performance of safety systems, the highest temperature of the heating rods in the testing section did not exceed the permissible temperature. (E.G.) [pt

  7. Radiation accidents with global consequences for the population. Problems of risk evaluation

    International Nuclear Information System (INIS)

    Vasilev, G.; Doncheva, B.; Stoilova, S.; Miloslavov, V.; Tsenova, T.; Novkirishki, V.

    1987-01-01

    The theoretical problems concerning the delayed impacts as a result of considerable radiation accidents are discussed. The attention is paid to the maximum individual doses which are relatively low but many people are affected. In these cases, the risk evaluation is based on the cancerogenesis, genetic and teratogenetic consequences among the concerned population. The main equation of the linear threshold-free model 'dose effect' is subjected to analysis. Considering the real prognostic importance of this equation the following recommendations are made: further observation on epidemic diseases; investigation of teratogenetic consequences in connection with the radiation doses obtained during the antenatal development; radiation-hygienic standardization of the oral absorbtion of radionuclides for short and long periods of time; effective equivalent dose determination according to the irradiated organ or tissue (mammary glands, lungs, red marrow, gonads, skin); necessity of national system for in time announcement of radiation accidents, as well as suitable control of the internal and the external irradiation

  8. Final report of the accident phenomenology and consequence (APAC) methodology evaluation. Spills Working Group

    Energy Technology Data Exchange (ETDEWEB)

    Brereton, S.; Shinn, J. [Lawrence Livermore National Lab., CA (United States); Hesse, D [Battelle Columbus Labs., OH (United States); Kaninich, D. [Westinghouse Savannah River Co., Aiken, SC (United States); Lazaro, M. [Argonne National Lab., IL (United States); Mubayi, V. [Brookhaven National Lab., Upton, NY (United States)

    1997-08-01

    The Spills Working Group was one of six working groups established under the Accident Phenomenology and Consequence (APAC) methodology evaluation program. The objectives of APAC were to assess methodologies available in the accident phenomenology and consequence analysis area and to evaluate their adequacy for use in preparing DOE facility safety basis documentation, such as Basis for Interim Operation (BIO), Justification for Continued Operation (JCO), Hazard Analysis Documents, and Safety Analysis Reports (SARs). Additional objectives of APAC were to identify development needs and to define standard practices to be followed in the analyses supporting facility safety basis documentation. The Spills Working Group focused on methodologies for estimating four types of spill source terms: liquid chemical spills and evaporation, pressurized liquid/gas releases, solid spills and resuspension/sublimation, and resuspension of particulate matter from liquid spills.

  9. The usefulness of time-dependent reactor accident consequence modelling for emergency response planning

    International Nuclear Information System (INIS)

    Paretzke, H.G.; Jacob, P.; Mueller, H.; Proehl, G.

    1989-01-01

    After major releases of radionuclides into the atmosphere fast reaction of authorities will be necessary to inform the public of potential consequences and to consider and optimize mitigating actions. These activities require availability of well designed computer models, adequate and fast measurements and prior training of responsible persons. The quantitative assessment models should be capable of taking into account of actual atmospheric dispersion conditions, actual deposition situation (dry, rain, snow, fog), seasonal status of the agriculture, food processing and distribution pathways, etc. In this paper the usefulness of such models will be discussed, their limitations, the relative importance of exposure pathways and a selection of important methods to decrease the activity in food products after an accident. Real-time reactor accident consequence models should be considered as a condition sine qua non for responsible use of nuclear power for electricity production

  10. German offsite accident consequence model for nuclear facilities: further development and application

    International Nuclear Information System (INIS)

    Bayer, A.

    1985-01-01

    The German Offsite Accident Consequence Model - first applied in the German Risk Study for nuclear power plants with light water reactors - has been further developed with the improvement of several important submodels in the areas of atmospheric dispersion, shielding effects of houses, and the foodchains. To aid interpretation, the presentation of results has been extended with special emphasis on the presentation of the loss of life expectancy. The accident consequence model has been further developed for application to risk assessments for other nuclear facilities, e.g., the liquid metal fast breeder reactor (SNR-300) and the high temperature gas cooled reactor. Moreover the model have been further developed in the area of optimal countermeasure strategies (sheltering, evacuation, etc.) in the case of the Central European conditions. Preliminary considerations has been performed in connection with safety goals on the basis of doses

  11. Environmental consequences of the Chernobyl accident and their remediation: Twenty years of experience

    International Nuclear Information System (INIS)

    Anspaugh, L.R.

    2005-01-01

    The explosion on 26 April 1986 at the Chernobyl Nuclear Power Plant located just 100 km from the city of Kyiv in what was then the Soviet Union and now is Ukraine, and consequent ten days' reactor fire resulted in an unprecedented release of radiation and unpredicted adverse consequences both for the public and the environment. Indeed, the IAEA has characterized the event as the 'foremost nuclear catastrophe in human history' and the largest regional release of radionuclides into the atmosphere. Massive radioactive contamination forced the evacuation of more than 100,000 people from the affected region during 1986, and the relocation, after 1986, of another 200,000 from Belarus, the Russian Federation and Ukraine. Some five million people continue to live in areas contaminated by the accident and have to deal with its environmental, health, social and economic consequences. The national governments of the three affected countries, supported by international organizations, have undertaken costly efforts to remedy contamination, provide medical services and restore the region's social and economic well-being. The accident's consequences were not limited to the territories of Belarus, Russia and Ukraine but resulted in substantial transboundary atmospheric transfer and subsequent contamination of numerous European countries that also encountered problems of radiation protection of their populations, although to less extent than the three more affected countries. Although the accident occurred nearly two decades ago, controversy still surrounds the impact of the nuclear disaster. Therefore the IAEA, in cooperation with FAO, UNDP, UNEP, UNOCHA, UNSCEAR, WHO and The World Bank, as well as the competent authorities of Belarus, the Russian Federation and Ukraine, established the Chernobyl Forum in 2003. The mission of the Forum was - through a series of managerial and expert meetings to generate 'authoritative consensual statements' on the environmental consequences and

  12. An assessment of the radiological consequences of accidents in research reactors

    International Nuclear Information System (INIS)

    Ferreira, N.L.D.

    1992-01-01

    This work analyses the radiological consequences of accidents in two types of research reactors: a 5 MWt open pool reactor and a 50 MWt PWR reactor. Two siting cases have been considered: the reactor located near to a large population center and sited in a rural area. The influence of several factors such as source term, meteorological conditions and population distribution have been considered in the present analysis. (author)

  13. Measures for reduction of severe accident consequences: Comprehensive evaluation of the results sponsored by the BMI

    International Nuclear Information System (INIS)

    Bracht, K.; Friedrichs, H.G.

    1989-01-01

    A number of analytical studies were initial in the past by the Federal Ministry of Interior (BMI) of FRG, to investigate the potential of additional constructive measures for risk reduction. Those measures were proposed especially against uncontrolled overpressurization of the containment due to continuous gas/steam generation, penetration of the foundation of the reactor building by melt-concrete interaction, and failure of the containment due to violent hydrogen combustion. This report gives an overview about those studies and summarizes their results. Concerning uncontrolled overpressurization, only filtered venting may be a reasonable measure, while it seems to make not much sense, to look at measures against penetration of the foundation like 'core-catcher' in further detail. To prevent hydrogen combustion with severe consequences, several potential possibilities exist, but none of them can be considered as a safe measure. Additional analysis concerning hydrogen distribution and combustion in a multi-compartment containment are necessary. All studies mentioned in this report, deal with additional constructive measures to mitigate the consequences of severe accidents. Up to day in FRG, the potential of accident prevention and mitigation of its consequences by still or again operable and already existing systems of a plant have not been investigated in detail. As indicated by first results, the use of those systems in the frame of an appropriate accident management may have a large potential for risk reduction. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  15. Evaluation of postpone Ch NPP accident consequences. Health state of children born in families of cleaners taking part in the Chernobyl accident consequences

    International Nuclear Information System (INIS)

    Koretskaya, L.S.; Kornesku, A.V.; Koretskaya, L.I.; Moldovanu, M.; Samotyya, E.E.; Etsko, L.A.

    2009-01-01

    The analysis of the results of the medical investigation during 1996-2008 years of the health of the about 200 children (107 girls, 115 boys 5-17 years old), which parents took part in liquidation of C NPP accident have been performed. Clinical, immunological disturbance and increased chromosome mutagenesis intensity in somatic cells of the investigated children gave reason to classify them among those with the increased risk of probability of the pathology with genetic component. This may be considerate as a well founded conclusion for elaboration of separated registry of the children born in families of liquidators. The immunological disturbance of children probably reflects the disturbance of the differential processes and maturity of cells of thymus and can be the consequences of the factors influence related with the clinic pathologies. The results permitted to select the group for feature cytogenesis and clinic investigations (authors)

  16. CITADEL: a computer code for the analysis of iodine behavior in steam generator tube rupture accidents

    International Nuclear Information System (INIS)

    1982-04-01

    The computer code CITADEL was written to analyze iodine behavior during steam generator tube rupture accidents. The code models the transport and deposition of iodine from its point of escape at the steam generator primary break until its release to the environment. This report provides a brief description of the code including its input requirements and the nature and form of its output. A user's guide describing the manner in which the input data are required to be set up to run the code is also provided

  17. Two codes used in analysis of rod ejection accident for Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Zhu Xinguan

    1987-12-01

    Two codes were developed to analyse rod ejection accident for Qinshan Nuclear Power Plant. One was based on point model with temperature reactivity feedback. In this code, the worth of ejected rod was obtained under'adiabatic' approximation. In the other code, the Nodal Green's Function Method was used to solve space-time dependent neutron diffusion equation. Using these codes, the transient core-power have been calculated for two rod ejection cases at beginning of core-life in Qinshan Nuclear Power Plant

  18. Evaluation of methods to compare consequences from hazardous materials transportation accidents

    International Nuclear Information System (INIS)

    Rhoads, R.E.; Franklin, A.L.; Lavender, J.C.

    1986-10-01

    This report presents the results of a project to develop a framework for making meaningful comparisons of the consequences from transportation accidents involving hazardous materials. The project was conducted in two phases. In Phase I, methods that could potentially be used to develop the consequence comparisons for hazardous material transportation accidents were identified and reviewed. Potential improvements were identified and an evaluation of the improved methods was performed. Based on this evaluation, several methods were selected for detailed evaluation in Phase II of the project. The methods selected were location-dependent scenarios, figure of merit and risk assessment. This evaluation included application of the methods to a sample problem which compares the consequences of four representative hazardous materials - chlorine, propane, spent nuclear fuel and class A explosives. These materials were selected because they represented a broad class of hazardous material properties and consequence mechanisms. The sample case aplication relied extensively on consequence calculations performed in previous transportation risk assessment studies. A consultant was employed to assist in developing consequence models for explosives. The results of the detailed evaluation of the three consequence comparison methods indicates that methods are available to perform technically defensible comparisons of the consequences from a wide variety of hazardous materials. Location-dependent scenario and risk assessment methods are available now and the figure of merit method could be developed with additional effort. All of the methods require substantial effort to implement. Methods that would require substantially less effort were identified in the preliminary evaluation, but questions of technical accuracy preclude their application on a scale. These methods may have application to specific cases, however

  19. Uncertainty and sensitivity analysis of food pathway results with the MACCS Reactor Accident Consequence Model

    International Nuclear Information System (INIS)

    Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the food pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 87 imprecisely-known input variables on the following reactor accident consequences are studied: crop growing season dose, crop long-term dose, milk growing season dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, area dependent cost, crop disposal cost, milk disposal cost, condemnation area, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: fraction of cesium deposition on grain fields that is retained on plant surfaces and transferred directly to grain, maximum allowable ground concentrations of Cs-137 and Sr-90 for production of crops, ground concentrations of Cs-134, Cs-137 and I-131 at which the disposal of milk will be initiated due to accidents that occur during the growing season, ground concentrations of Cs-134, I-131 and Sr-90 at which the disposal of crops will be initiated due to accidents that occur during the growing season, rate of depletion of Cs-137 and Sr-90 from the root zone, transfer of Sr-90 from soil to legumes, transfer of Cs-137 from soil to pasture, transfer of cesium from animal feed to meat, and the transfer of cesium, iodine and strontium from animal feed to milk

  20. Uncertainty and sensitivity analysis of food pathway results with the MACCS reactor accident consequence model

    International Nuclear Information System (INIS)

    Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the food pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 87 imprecisely-known input variables on the following reactor accident consequences are studied: crop growing-season dose, crop long-term dose, milk growing-season dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, area dependent cost, crop disposal cost, milk disposal cost, condemnation area, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: fraction of cesium deposition on grain fields that is retained on plant surfaces and transferred directly to grain, maximum allowable ground concentrations of Cs-137 and Sr-90 for production of crops, ground concentrations of Cs-134, Cs-137 and I-131 at which the disposal of milk will be initiated due to accidents that occur during the growing season, ground concentrations of Cs-134, I-131 and Sr-90 at which the disposal of crops will be initiated due to accidents that occur during the growing season, rate of depletion of Cs-137 and Sr-90 from the root zone, transfer of Sr-90 from soil to legumes, transfer of Cs-137 from soil to pasture, transfer of cesium from animal feed to meat, and the transfer of cesium, iodine and strontium from animal feed to milk

  1. Code portability and data management considerations in the SAS3D LMFBR accident-analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1981-01-01

    The SAS3D code was produced from a predecessor in order to reduce or eliminate interrelated problems in the areas of code portability, the large size of the code, inflexibility in the use of memory and the size of cases that can be run, code maintenance, and running speed. Many conventional solutions, such as variable dimensioning, disk storage, virtual memory, and existing code-maintenance utilities were not feasible or did not help in this case. A new data management scheme was developed, coding standards and procedures were adopted, special machine-dependent routines were written, and a portable source code processing code was written. The resulting code is quite portable, quite flexible in the use of memory and the size of cases that can be run, much easier to maintain, and faster running. SAS3D is still a large, long running code that only runs well if sufficient main memory is available

  2. SECAN: system for the evaluation of the consequences of nuclear accidents

    International Nuclear Information System (INIS)

    Avila, R.; Cardenas, G.; Sanchez, J.L.; Magarino, R.

    1993-01-01

    The SECAN code is an useful mean to evaluate operatively accidents that cause radioactive releases to the atmosphere and its impact on the man and the environment. The calculation of external radiation doses to the whole body because of the cloud and the deposit, as well as the doses in isolated organs by inhalation, superficial and volumetric concentration values, and external doses rate is found among its possibilities. This code also offers criteria to take measures for the protection of the population according to the zone covered and the amount of peoples involved depending on the selected intervention level. The main output ways are presented in tables, graphic and automatized map of the site of Juragua nuclear power plant, where is clearly shown the radiological situation. This code has been elaborated in Turbo Pascal 6.0 for IBM-compatible computers

  3. Effects of spent fuel types on offsite consequences of hypothetical accidents

    International Nuclear Information System (INIS)

    Courtney, J. C.; Dwight, C. C.; Lehto, M. A.

    2000-01-01

    Argonne National Laboratory (ANL) conducts experimental work on the development of waste forms suitable for several types of spent fuel at its facility on the Idaho National Engineering and Environmental Laboratory (INEEL) located 48 km West of Idaho Falls, ID. The objective of this paper is to compare the offsite radiological consequences of hypothetical accidents involving the various types of spent nuclear fuel handled in nonreactor nuclear facilities. The highest offsite total effective dose equivalents (TEDEs) are estimated at a receptor located about 5 km SSE of ANL facilities. Criticality safety considerations limit the amount of enriched uranium and plutonium that could be at risk in any given scenario. Heat generated by decay of fission products and actinides does not limit the masses of spent fuel within any given operation because the minimum time elapsed since fissions occurred in any form is at least five years. At cooling times of this magnitude, fewer than ten radionuclides account for 99% of the projected TEDE at offsite receptors for any credible accident. Elimination of all but the most important nuclides allows rapid assessments of offsite doses with little loss of accuracy. Since the ARF (airborne release fraction), RF (respirable fraction), LPF (leak path fraction) and atmospheric dilution factor (χ/Q) can vary by orders of magnitude, it is not productive to consider nuclides that contribute less than a few percent of the total dose. Therefore, only 134 Cs, 137 Cs- 137m Ba, and the actinides significantly influence the offsite radiological consequences of severe accidents. Even using highly conservative assumptions in estimating radiological consequences, they remain well below current Department of Energy guidelines for highly unlikely accidents

  4. 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.

  5. Qualification of ARROTTA code for LWR accident analysis

    International Nuclear Information System (INIS)

    Huang, P.-H.; Peng, K.Y.; Lin, W.-C.; Wu, J.-Y.

    2004-01-01

    This paper presents the qualification efforts performed by TPC and INER for the 3-D spatial kinetics code ARROTTA for LWR core transient analysis. TPC and INER started a joint 5 year project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multi-dimensional kinetics calculations such as rod ejection for PWR and rod drop for BWR. To qualify ARROTTA for analysis of FSAR licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicated that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include NEACRP rod ejection problem, 3-D LMW LWR rod withdrawal/insertion problem, and 3-D LRA BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared to other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multi-dimensional core transient for LWRs. (author)

  6. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    International Nuclear Information System (INIS)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor

    2013-01-01

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

  7. Comparative analysis of unprotected loss-of-flow accidents for the 1.0 m EFR-LVC core using different computer codes

    International Nuclear Information System (INIS)

    Royl, P.; Frizonnet, J.M.; Moran, J.

    1993-02-01

    A comparative analysis of the unprotected loss of flow (ULOF) accident has been performed for the LVC core (Lower Void Core) of the European Fast Reactor EFR with the FRAX5B and FRAX5C codes from the AEA-T, the PHYSURAC code from CEA and the SAS4A REF92 code system developed jointly between KfK, CEA and PNC. The accident is triggered by the run down of the coolant pumps with failure to trip the reactor by the primary and/or secondary shutdown system. Only a limited amount of mitigating reactivity from the third shutdown line was considered so that the accident can progress into boiling and core disruption. This code outlines the important modelling differences and compares the different simulations. The discussion of the rather wide spectrum of calculated accident progressions identifies the generic differences, relates them to the applied models, and summarizes the key points that are responsible for the different progressions. A comparison of the consequence spectrum from all simulations indicates zero work energies for the majority of the calculations. All simulations show up the need for a continued accident analysis into the early and late transition phase

  8. Radiological consequences of accidents during disposal of spent nuclear fuel in a deep borehole

    International Nuclear Information System (INIS)

    Grundfelt, Bertil

    2013-07-01

    In this report, an analysis of the radiological consequences of potential accidents during disposal of spent nuclear fuel in deep boreholes is presented. The results presented should be seen as coarse estimates of possible radiological consequences of a canister being stuck in a borehole during disposal rather than being the results of a full safety analysis. In the concept for deep borehole disposal of spent nuclear fuel developed by Sandia National Laboratories, the fuel is assumed to be encapsulated in mild steel canisters and stacked between 3 and 5 km depth in boreholes that are cased with perforated mild steel casing tubes. The canisters are joined together by couplings to form strings of 40 canisters and lowered into the borehole. When a canister string has been emplaced in the borehole, a bridge plug is installed above the string and a 10 metres long concrete plug is cast on top of the bridge plug creating a floor for the disposal of the next sting. In total 10 canister strings, in all 400 canisters, are assumed to be disposed of at between 3 and 5 kilometres depth in one borehole. An analysis of potential accidents during the disposal operations shows that the potentially worst accident would be that a canister string is stuck above the disposal zone of a borehole and cannot be retrieved. In such a case, the borehole may have to be sealed in the best possible way and abandoned. The consequences of this could be that one or more leaking canisters are stuck in a borehole section with mobile groundwater. In the case of a leaking canister being stuck in a borehole section with mobile groundwater, the potential radiological consequences are likely to be dominated by the release of the so-called Instant Release Fraction (IRF) of the radionuclide inventory, i.e. the fraction of the radionuclides that as a consequence of the in-core conditions are present in the annulus between the fuel pellets and the cladding or on the grain boundaries of the UO 2 matrix. The

  9. Radiological consequences of accidents during disposal of spent nuclear fuel in a deep borehole

    Energy Technology Data Exchange (ETDEWEB)

    Grundfelt, Bertil [Kemakta Konsult AB, Stockholm (Sweden)

    2013-07-15

    In this report, an analysis of the radiological consequences of potential accidents during disposal of spent nuclear fuel in deep boreholes is presented. The results presented should be seen as coarse estimates of possible radiological consequences of a canister being stuck in a borehole during disposal rather than being the results of a full safety analysis. In the concept for deep borehole disposal of spent nuclear fuel developed by Sandia National Laboratories, the fuel is assumed to be encapsulated in mild steel canisters and stacked between 3 and 5 km depth in boreholes that are cased with perforated mild steel casing tubes. The canisters are joined together by couplings to form strings of 40 canisters and lowered into the borehole. When a canister string has been emplaced in the borehole, a bridge plug is installed above the string and a 10 metres long concrete plug is cast on top of the bridge plug creating a floor for the disposal of the next sting. In total 10 canister strings, in all 400 canisters, are assumed to be disposed of at between 3 and 5 kilometres depth in one borehole. An analysis of potential accidents during the disposal operations shows that the potentially worst accident would be that a canister string is stuck above the disposal zone of a borehole and cannot be retrieved. In such a case, the borehole may have to be sealed in the best possible way and abandoned. The consequences of this could be that one or more leaking canisters are stuck in a borehole section with mobile groundwater. In the case of a leaking canister being stuck in a borehole section with mobile groundwater, the potential radiological consequences are likely to be dominated by the release of the so-called Instant Release Fraction (IRF) of the radionuclide inventory, i.e. the fraction of the radionuclides that as a consequence of the in-core conditions are present in the annulus between the fuel pellets and the cladding or on the grain boundaries of the UO{sub 2} matrix

  10. The importance of long range atmospheric transport in probabilistic accident consequence assessment

    International Nuclear Information System (INIS)

    ApSimon, H.M.; Goddard, A.J.H.; Wilson, J.J.N.

    1988-01-01

    The disaster at the Chernobyl-4 reactor has demonstrated that severe nuclear accidents can give rise to significant radiological consequences several thousand kilometres from the source. The subsequent dispersion of the release over much of Western Europe further demonstrated the importance of synoptic scale weather patterns in determining the magnitude of the consequences of such accidents. A version of the MESOS-II European scale trajectory model, which is able to simulate large scale variations in weather conditions through the use of spatially and temporally variable meteorological input data, has been used to simulate the pattern of dispersion from Chernobyl with some success. This paper presents the results of probabilistic consequence assessments for a number of West European sites, made using the MESOS-II model. The results illustrate the effects, on probabilistic assessments, of using a more realistic treatment of long range atmospheric transport than the Gaussian plume model and also the spatial variation in the distributions of consequences arising from the variation in synoptic scale weather conditions across Western Europe

  11. Bounding Radionuclide Inventory and Accident Consequence Calculation for the 1L Target

    International Nuclear Information System (INIS)

    Kelsey, Charles T. IV

    2011-01-01

    A bounding radionuclide inventory for the tungsten of the Los Alamos Neutron Science Center (LANSCE) IL Target is calculated. Based on the bounding inventory, the dose resulting from the maximum credible incident (MCI) is calculated for the maximally exposed offsite individual (MEOl). The design basis accident involves tungsten target oxidation following a loss of cooling accident. Also calculated for the bounding radionuclide inventory is the ratio to the LANSCE inventory threshold for purposes of inventory control as described in the target inventory control policy. A bounding radionuclide inventory calculation for the lL Target was completed using the MCNPX and CINDER'90 codes. Continuous beam delivery at 200 (micro)A to 2500 mA·h was assumed. The total calculated activity following this irradiation period is 205,000 Ci. The dose to the MEOI from the MCI is 213 mrem for the bounding inventory. The LANSCE inventory control threshold ratio is 132.

  12. Fukushima, one year after. First analyses of the accident and of its consequences

    International Nuclear Information System (INIS)

    2012-01-01

    This report proposes assessments and discussions of knowledge gathered by the IRSN during the first twelve months following the Fukushima accident to understand the status of the installations, to assess the releases, and to analyse and assess the consequences of the accident on workers and the impact on the population and environment. After a description of a boiling water reactor (general description, confinement barriers, safeguard systems), and of the earthquake, the authors describe and comment the consequences for several reactors (Fukushima-Dai-ini, Onagawa, Tokai, Higashidoru and Hamaoka). Then, they more precisely describe the Fukushima-Dai-ichi accident by distinguishing different periods (first two weeks, next three weeks, after the 17 of April). They analyse and comment the environmental impact in Japan (atmospheric dispersion of radioactive releases, ground contamination, impact of radioactive fallouts, contamination of the marine environment, and predictable impact on marine and ground ecosystems). They describe the actions undertaken to protect the population and in terms of post-accidental management, comment assessments of the dosimetric and health impact (workers and population exposure). They finally discuss the long range impact

  13. Uncertainty and sensitivity analysis of early exposure results with the MACCS Reactor Accident Consequence Model

    International Nuclear Information System (INIS)

    Helton, J.C.; Johnson, J.D.; McKay, M.D.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the early health effects associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 34 imprecisely known input variables on the following reactor accident consequences are studied: number of early fatalities, number of cases of prodromal vomiting, population dose within 10 mi of the reactor, population dose within 1000 mi of the reactor, individual early fatality probability within 1 mi of the reactor, and maximum early fatality distance. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: scaling factor for horizontal dispersion, dry deposition velocity, inhalation protection factor for nonevacuees, groundshine shielding factor for nonevacuees, early fatality hazard function alpha value for bone marrow exposure, and scaling factor for vertical dispersion

  14. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, S. E-mail: reyessuarezl@llnl.gov; Latkowski, J.F.; Gomez del Rio, J.; Sanz, J

    2001-05-21

    Previous studies of the safety and environmental aspects of the HYLIFE-II inertial fusion energy power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work, computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) have been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here we consider a severe loss of coolant accident (LOCA) in conjunction with simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the confinement) and of the two barriers surrounding the chamber (inner shielding and confinement building itself). Even though confinement failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product transport and release. The results of these calculations show that the estimated off-site dose is less than 5 mSv (0.5 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  15. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    Science.gov (United States)

    Reyes, S.; Latkowski, J. F.; Gomez del Rio, J.; Sanz, J.

    2001-05-01

    Previous studies of the safety and environmental aspects of the HYLIFE-II inertial fusion energy power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work, computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) have been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here we consider a severe loss of coolant accident (LOCA) in conjunction with simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the confinement) and of the two barriers surrounding the chamber (inner shielding and confinement building itself). Even though confinement failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product transport and release. The results of these calculations show that the estimated off-site dose is less than 5 mSv (0.5 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  16. Computer code TRANS-ACE predicting for fire and explosion accidents in nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Abe, Hitoshi; Nishio; Gunji; Naito, Yoshitaka

    1993-11-01

    The accident analysis code TRANS-ACE was developed to evaluate the safety of a ventilation system in a reprocessing plant in the event of fire and explosion accidents. TRANS-ACE can evaluate not only the integrity of a ventilation system containing HEPA filters but also the source term of radioactive materials for release out of a plant. It calculates the temperature, pressure, flow rate, transport of combustion materials and confinement of radioactive materials in the network of a ventilation system that might experience a fire or explosion accident. TRANS-ACE is based on the one-dimensional compressible thermo-fluid analysis code EVENT developed by Los Alamos National Laboratory (LANL). Calculational functions are added for the radioactive source term, heat transfer and radiation to cell and duct walls and HEPA filter integrity. For the second edition in the report, TRANS-ACE has been improved incorporating functions for the initial steady-state calculation to determine the flow rates, pressure drops and temperature in the network before an accident mode analysis. It is also improved to include flow resistance calculations of the filters and blowers in the network and to have an easy to use code by simplifying the input formats. This report is to prepare an explanation of the mathematical model for TRANS-ACE code and to be the user's manual. (author)

  17. Audit calculations of accidents analysis for second unit of Ignalina NPP with ATHLET code

    International Nuclear Information System (INIS)

    Adomavicius, A.; Belousov, A.; Ognerubov, V.

    2004-01-01

    Background of thermo hydraulic processes audit calculations in the frame of RSR-2 project is presented. Assumptions for the design based accident - RBMK-1500 group distributor header break analysis and modeling are presented. Audit calculations by ATHLET code and evaluation of results were provided. (author)

  18. Research on the improvement of nuclear safety -The development of a severe accident analysis code-

    International Nuclear Information System (INIS)

    Kim, Heui Dong; Cho, Sung Won; Park, Jong Hwa; Hong, Sung Wan; Yoo, Dong Han; Hwang, Moon Kyoo; Noh, Kee Man; Song, Yong Man

    1995-07-01

    For prevention and mitigation of the containment failure during severe accident, the study is focused on the severe accident phenomena, especially, the ones occurring inside the cavity and is intended to improve existing models and develop analytical tools for the assessment of severe accidents. A correlation equation of the flame velocity of pre mixture gas of H 2 /air/steam has been suggested and combustion flame characteristic was analyzed using a developed computer code. For the analysis of the expansion phase of vapor explosion, the mechanical model has been developed. The development of a debris entrainment model in a reactor cavity with captured volume has been continued to review and examine the limitation and deficiencies of the existing models. Pre-test calculation was performed to support the severe accident experiment for molten corium concrete interaction study and the crust formation process and heat transfer characteristics of the crust have been carried out. A stress analysis code was developed using finite element method for the reactor vessel lower head failure analysis. Through international program of PHEBUS-FP and participation in the software development, the research on the core degradation process and fission products release and transportation are undergoing. CONTAIN and MELCOR codes were continuously updated under the cooperation with USNRC and French developed computer codes such as ICARE2, ESCADRE, SOPHAEROS were also installed into the SUN workstation. 204 figs, 61 tabs, 87 refs. (Author)

  19. Calculation code used in criticality analyses for the accident of JCO precipitation tank

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori

    2000-01-01

    In order to evaluate nuclear features on criticality accident formed at the nuclear fuel processing facility in Tokai Works of the JCO, Ltd. (JCO), in Tokai-mura, Ibaraki prefecture, dynamic analyses to calculate output change after occurring the accident as well as criticality analyses to calculate reactivity added to precipitation tank, were carried out according to scenario on accident formation. For the criticality analyses, a continuous energy Monte Carlo code MCNP was used to carry out calculation of reactivity fed into the precipitation tank as correctly as possible. And, SRAC code system was used for calculation on temperature and void reactivity coefficients, effective delayed neutron ratio beta eff , and instantaneous neutron generation time required for parameters controlling transition features at criticality accident. In addition, for the dynamic analyses, because of necessity of considering on volume expansion of solution fuels used as exothermic body and radiation decomposition gas forming into solution, output behavior, numbers of nuclear fission, and so forth at initial burst portion were calculated by using TRACE and quasi-regular code, at a center of AGNES-2 promoting on its development in JAERI. Here were reported on outlines and an analysis example on calculation code using for the nuclear features evaluation. (G.K.)

  20. Consequences in Norway after a hypothetical accident at Sellafield - Predicted impacts on the environment

    International Nuclear Information System (INIS)

    Thoerring, H.; Ytre-Eide, M.A.; Liland, A.

    2010-12-01

    This report describes the possible environmental consequences for Norway due to a hypothetical accident at the Sellafield complex in the UK. The scenario considered involves an explosion and fire at the B215 facility resulting in a 1 % release of the total HAL (Highly Active liquor) inventory of radioactive waste with a subsequent air transport and deposition in Norway. Air transport modelling is based on real meteorological data from October 2008 with wind direction towards Norway and heavy precipitation. This weather is considered to be quite representative as typical seasonal weather. Based on this weather scenario, the estimated fallout in Norway will be ∼ 17 P Bq of caesium-137 which is 7 times higher than the fallout from the Chernobyl accident. The modelled radioactive contamination is linked with data on transfer to the food chain and statistics on production and hunting to assess the consequences for foodstuffs. The investigation has been limited to the terrestrial environment, focussing on wild berries, fungi, and animals grazing unimproved pastures (i.e. various types of game, reindeer, sheep and goats). The predicted consequences are severe - especially in connection to sheep and goat production. Up to 80 % of the lambs in Norway could be exceeding the food intervention levels for radiocaesium the first years after the fallout, with 30-40 % likely to be above for many years. There will, consequently, be a need for extensive countermeasures in large areas for years or even decades involving several hundred thousand animals each year. Large consequences are also expected for reindeer husbandry - the first year in particular due to the time of fallout which is just prior to winter slaughter. The consequences will be most sever for reindeer herding in middle and southern parts of Norway, but problems may reach as far north as Finnmark where we find the majority of Norwegian reindeer production. The consequences for game will mostly depend on the regional

  1. Consequences in Norway after a hypothetical accident at Sellafield - Predicted impacts on the environment

    Energy Technology Data Exchange (ETDEWEB)

    Thoerring, H.; Ytre-Eide, M.A.; Liland, A.

    2010-12-15

    This report describes the possible environmental consequences for Norway due to a hypothetical accident at the Sellafield complex in the UK. The scenario considered involves an explosion and fire at the B215 facility resulting in a 1 % release of the total HAL (Highly Active liquor) inventory of radioactive waste with a subsequent air transport and deposition in Norway. Air transport modelling is based on real meteorological data from October 2008 with wind direction towards Norway and heavy precipitation. This weather is considered to be quite representative as typical seasonal weather. Based on this weather scenario, the estimated fallout in Norway will be approx 17 P Bq of caesium-137 which is 7 times higher than the fallout from the Chernobyl accident. The modelled radioactive contamination is linked with data on transfer to the food chain and statistics on production and hunting to assess the consequences for foodstuffs. The investigation has been limited to the terrestrial environment, focussing on wild berries, fungi, and animals grazing unimproved pastures (i.e. various types of game, reindeer, sheep and goats). The predicted consequences are severe - especially in connection to sheep and goat production. Up to 80 % of the lambs in Norway could be exceeding the food intervention levels for radiocaesium the first years after the fallout, with 30-40 % likely to be above for many years. There will, consequently, be a need for extensive countermeasures in large areas for years or even decades involving several hundred thousand animals each year. Large consequences are also expected for reindeer husbandry - the first year in particular due to the time of fallout which is just prior to winter slaughter. The consequences will be most sever for reindeer herding in middle and southern parts of Norway, but problems may reach as far north as Finnmark where we find the majority of Norwegian reindeer production. The consequences for game will mostly depend on the

  2. Estimation of doses received by operators in the 1958 RB reactor accident using the MCNP5 computer code simulation

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2012-01-01

    Full Text Available A numerical simulation of the radiological consequences of the RB reactor reactivity excursion accident, which occurred on October 15, 1958, and an estimation of the total doses received by the operators were run by the MCNP5 computer code. The simulation was carried out under the same assumptions as those used in the 1960 IAEA-organized experimental simulation of the accident: total fission energy of 80 MJ released in the accident and the frozen positions of the operators. The time interval of exposure to high doses received by the operators has been estimated. Data on the RB1/1958 reactor core relevant to the accident are given. A short summary of the accident scenario has been updated. A 3-D model of the reactor room and the RB reactor tank, with all the details of the core, created. For dose determination, 3-D simplified, homogenised, sexless and faceless phantoms, placed inside the reactor room, have been developed. The code was run for a number of neutron histories which have given a dose rate uncertainty of less than 2%. For the determination of radiation spectra escaping the reactor core and radiation interaction in the tissue of the phantoms, the MCNP5 code was run (in the KCODE option and “mode n p e”, with a 55-group neutron spectra, 35-group gamma ray spectra and a 10-group electron spectra. The doses were determined by using the conversion of flux density (obtained by the F4 tally in the phantoms to doses using factors taken from ICRP-74 and from the deposited energy of neutrons and gamma rays (obtained by the F6 tally in the phantoms’ tissue. A rough estimation of the time moment when the odour of ozone was sensed by the operators is estimated for the first time and given in Appendix A.1. Calculated total absorbed and equivalent doses are compared to the previously reported ones and an attempt to understand and explain the reasons for the obtained differences has been made. A Root Cause Analysis of the accident was done and

  3. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    CERN Document Server

    Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

    1982-01-01

    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

  4. Synthesis of the models used in France for the evaluation of the consequences of accident

    International Nuclear Information System (INIS)

    Crabol, B.

    1992-01-01

    In order to evaluate the consequences of an atmospheric release in case of an accident on a nuclear installation, different predictive models have been developed by the organizations involved in the management of the crisis. These models are of different numerical complexity: precalculated graphs, gaussian puff models or 3D models. The harmonization of these models, the definition of their use, notably in the first phases of the accident (predictive and real-time phases) have been discussed in a working group including representants of the utility, the safety authorities and the Meteorological Office. The reflexions of the group, the models already operational, those still under discussion and their use in the different technical crisis centers are presented

  5. Radioactive contamination of the Dutch soil in consequence of the nuclear reactor accident at Chernobyl

    International Nuclear Information System (INIS)

    Koester, H.W.; Mattern, F.C.M.; Pennders, R.M.J.

    1987-01-01

    As a consequence of the reactor accident in Chernobyl air, contaminated with radioactive materials, spread over the Netherlands. From 2nd May to 6th May, with dry and to a greater extent with wet deposits, important quantities of radionuclides came upon the earth surface. In this period the weather circumstances within the Netherlands differed strongly resulting in distinct variations in deposit. In this document a preliminary picture is given of the contamination of the Dutch bottom on the basis of soil samplings made in the first few months after the accident. No attention is paid to geographic differences in bottom contamination. The contamination of the bottom is expressed in Bq/kg dry soil as well as in Bq/m 2 soil. 5 refs.; 6 tabs. (H.W.)

  6. Consequences of the Chernobyl reactor accident with respect to the feeding of infants

    International Nuclear Information System (INIS)

    Schmidt, E.

    1987-01-01

    In view of the persisting and understandable fear of parents with regard to radioactivity in the food of their babies as a consequence of the Chernobyl reactor accident, the Commission on Nutrition of the Deutsche Gesellschaft fuer Kinderheilkunde (German Society of Pediatrics) and the Strahlenschutzkommission have published a statement. According to this statement, the maximum permissible level of radioactivity in commercial baby food has been fixed by the EC to be 370 Bq/kg. The dietetic food industry itself has fixed a maximum for its products which is only a tenth of the radioactivity level permitted by the EC directive. The milk powders for infants tested since the reactor accident contained no measurable radioactivity or only very low amounts of Cs 134 or Cs 137, correspondung to a maximum of 25 Bq/kg in the product. Late damage to health is not to be expected. (orig./ECB) [de

  7. Accident consequence analysis models applied to licensing process of nuclear installations, radioactive and conventional industries

    International Nuclear Information System (INIS)

    Senne Junior, Murillo; Vasconcelos, Vanderley de; Jordao, Elizabete

    2002-01-01

    The industrial accidents happened in the last years, particularly in the eighty's decade, had contributed in a significant way to call the attention to government authorities, industry and society as a whole, demanding mechanisms for preventing episodes that could affect people's safety and environment quality. Techniques and methods already thoroughly used in the nuclear, aeronautic and war industries were then adapted for performing analysis and evaluation of the risks associated to other industrial activities, especially in the petroleum, chemistry and petrochemical areas. Some models for analyzing the consequences of accidents involving fire and explosion, used in the licensing processes of nuclear and radioactive facilities, are presented in this paper. These models have also application in the licensing of conventional industrial facilities. (author)

  8. Radiological consequences of the reactor accident at Three Mile Island, Pennsylvania, USA

    International Nuclear Information System (INIS)

    Bryant, P.M.

    1979-01-01

    The findings of the Ad Hoc Population Dose Assessment Group are reviewed and summarized (Population Dose and Health Impact of the Accident at the Three Mile Island Nuclear Station. A preliminary assessment for the period March 28 through April 7, 1979; May 10, 1979. Washington DC, US Government Printing Office, 1979). The principal radionuclides released were xenon-133 and xenon-135, with some iodine-131. External exposure to gamma radiation was estimated from TLDs positioned at various on-site and off-site locations. Lung exposure from inhaled xenon-133 was calculated and air and milk monitoring results gave potential dose equivalents to a child's thyroid. These numerical estimates will be further refined, but only minor corrections to the present values are anticipated. The findings of this preliminary assessment have indicated that the radiological consequences to the public of the reactor accident are minimal. (UK)

  9. Nuclear accidents at the Fukushima Dai-ichi power plant. History, events and consequences

    International Nuclear Information System (INIS)

    Berniolles, Jean Marc

    2011-01-01

    Written few weeks after the accident, this article first recalls the circumstances (earthquake and tsunami), and then describes the accidental process within the primary vessels of the Fukushima Dai-ichi number 1, 2 and 3 reactors. The author then describes the interventions which aimed at cooling these three reactors, the problem faced for the storage of used fuels, and then the sequence of accidents: loss of cooling means leading to an explosion, problems faced in the different storage pools. He describes the various steps of recovery (primary cooling, electricity supply), discusses the consequences in terms of radioactivity releases in the plant environment with a comparison with Chernobyl, and also in terms of nature and quantity of radioactive elements. He comments radioactivity controls and measurements, evacuation measures, measurements performed by the IAEA, measurements of sea radioactivity, and the establishment of maps of ground radioactivity around the plant. He discusses the perspectives associated with these measurements for the surroundings of the Fukushima site

  10. Analysis code for large rupture accidents in ATR. SENHOR/FLOOD/HEATUP

    International Nuclear Information System (INIS)

    1997-08-01

    In the evaluation of thermo-hydraulic transient change, the behavior of core reflooding and the transient change of fuel temperature in the events which are classified in large rupture accidents of reactor coolant loss, that is the safety evaluation event of the ATR, the analysis codes for thermo-hydraulic transient change at the time of large rupture SENHOR, for core reflooding characteristics FLOOD and for fuel temperature HEATUP are used, respectively. The analysis code system for loss of coolant accident comprises the analysis code for thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC in addition to the above three codes. Based on the changes with time lapse of reactor thermal output and steam drum pressure obtained by the SENHOR, average reflooding rate is analyzed by the FLOOD, and the time of starting the turnaround of fuel cladding tube temperature and the heat transfer rate after the turnaround are determined. Based on these data, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The SENHOR code, the FLOOD code and the HEATUP code and various models for these codes are explained. The example of evaluation and the sensitivity analysis of the ATR plant are reported in the Appendix. (K.I.)

  11. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-15

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and err000.

  12. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-01

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and error

  13. Uncertainty and sensitivity analysis of chronic exposure results with the MACCS reactor accident consequence model

    International Nuclear Information System (INIS)

    Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the chronic exposure pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 75 imprecisely known input variables on the following reactor accident consequences are studied: crop growing season dose, crop long-term dose, water ingestion dose, milk growing season dose, long-term groundshine dose, long-term inhalation dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, total latent cancer fatalities, area-dependent cost, crop disposal cost, milk disposal cost, population-dependent cost, total economic cost, condemnation area, condemnation population, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: dry deposition velocity, transfer of cesium from animal feed to milk, transfer of cesium from animal feed to meat, ground concentration of Cs-134 at which the disposal of milk products will be initiated, transfer of Sr-90 from soil to legumes, maximum allowable ground concentration of Sr-90 for production of crops, fraction of cesium entering surface water that is consumed in drinking water, groundshine shielding factor, scale factor defining resuspension, dose reduction associated with decontamination, and ground concentration of 1-131 at which disposal of crops will be initiated due to accidents that occur during the growing season

  14. Uncertainty and sensitivity analysis of chronic exposure results with the MACCS reactor accident consequence model

    International Nuclear Information System (INIS)

    Helton, J.C; Johnson, J.D; Rollstin, J.A; Shiver, A.W; Sprung, J.L

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the chronic exposure pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 75 imprecisely known input variables on the following reactor accident consequences are studied: crop growing-season dose, crop long-term dose, water ingestion dose, milk growing-season dose, long-term groundshine dose, long-term inhalation dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, total latent cancer fatalities, area-dependent cost, crop disposal cost, milk disposal cost, population-dependent cost, total economic cost, condemnation area, condemnation population, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: dry deposition velocity, transfer of cesium from animal feed to milk, transfer of cesium from animal feed to meet, ground concentration of Cs-134 at which the disposal of milk products will be initiated, transfer of Sr-90 from soil to legumes, maximum allowable ground concentration of Sr-90 for production of crops, fraction of cesium entering surface water that is consumed in drinking water, groundshine shielding factor, scale factor defining resuspension, dose reduction associated with decontamination, and ground concentration of I-131 at which disposal of crops will be initiated due to accidents that occur during the growing season. Reducing the uncertainty in the preceding variables was found to substantially reduce the uncertainty in the

  15. 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

  16. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    International Nuclear Information System (INIS)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J.

    2001-01-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  17. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J. [Tractebel Energy Engineering, Brussels (Belgium)

    2001-07-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  18. Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

    International Nuclear Information System (INIS)

    Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu

    2015-01-01

    Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance

  19. Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance.

  20. Economic consequences of the Chernobyl accident in Norway in 1986 and 1987

    International Nuclear Information System (INIS)

    Tveten, U.

    1988-01-01

    In the accident consequence assessment (ACA) area there is extensive cooperation between the Nordic countries (Denmark, Finland, Norway, and Sweden), performed within the Nordic Safety Program, and partially funded by the Nordic Council of Ministers, via the Nordic Liaison Committee for Atomic Energy. One of the 17 projects in the ACA-related program area is concerned with the economic consequences of the Chernobyl accident in Finland, Norway, and Sweden. This paper is limited to describing conditions in Norway. There are areas in Norway where the Chernobyl fallout is >100 kBq/m 2 , and the total amount of radiocesium deposited over Norway is estimated by the National Institute for Radiation Hygiene to be 6% of the radiocesium released from the reactor. The areas where ground concentrations are highest are mostly in sparsely populated mountain areas. These areas are, however, important in connection with several nutritional pathways, notably, sheep, goats, reindeer, and freshwater fish. The purpose of this paper is to summarize information on mitigating actions and economic consequences of the deposited radioactive materials to Norwegian agriculture in the 1986-87 and 1987-88 slaughtering periods

  1. Model review and evaluation for application in DOE safety basis documentation of chemical accidents - modeling guidance for atmospheric dispersion and consequence assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Woodarad, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanna, S. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hesse, D. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Huang, J. -C. [Argonne National Lab. (ANL), Argonne, IL (United States); Lewis, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Mazzola, C. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    1997-09-01

    The U.S. Department of Energy (DOE), through its Defense Programs (DP), Office of Engineering and Operations Suppon, established the Accident Phenomenology and Consequence (AP AC) Methodology Evaluation Program to identify and evaluate methodologies and computer codes to support accident phenomenological and consequence calculations for both radiological and nonradiological materials at DOE facilities and to identify development needs. The program is also intended to define and recommend "best or good engineering/safety analysis practices" to be followed in preparing ''design or beyond design basis" assessments to be included in DOE nuclear and nonnuclear facility safety documents. The AP AC effort is intended to provide scientifically sound and more consistent analytical approaches, by identifying model selection procedures and application methodologies, in order to enhance safety analysis activities throughout the DOE complex.

  2. The environmental restoration in the management of radiological accidents with off site consequences

    International Nuclear Information System (INIS)

    Vazquez, C.; Montero, M.; Moraleda, M.; Diaz, J.; Claver, F.; Valles, O.; Rodriguez, N.; Gutierrez, J.

    1998-01-01

    Radiological accidents are among the potential cases of environmental contamination that could have consequences on the health of the population. These accidents, associated with an increase in the level of radiological exposure surpassing the natural background, have been investigated in greater depth than other conventional accidents. This investigation has included the evaluation of their probability, magnitude and consequences in order to establish safety norms. Nevertheless, the social perception of this type of risk appears to be disproportionately high. The development of a comprehensible and adequate standardized system for the evaluation of the radiological risk and the applicability of corrective actions to reduce this type of risk at local level, will undoubtedly contribute to increase the public confidence in the advised options for the restoration of environments contaminated with the long lived radionuclides. This system should consider the local specificity of each contaminated place, and take into account the associated unwanted consequences for each option. This paper presents the first results of a system to help the decision makers in the quantitative evaluation of the radiological risk produced by long lived radionuclides Cs 137, Cs 134 and Sr 90 spread over urban, agricultural and semi-natural environments and the applicable options to reduce it. The evaluation of these applicable options is made considering the reduction of dose that can be reached, the monetary costs and the significant associated secondary effects if there are any. All these factors are integrated for a time period depending on the half-life of the contaminants and on their strength and distribution on the scenario when intervention is being planned. (authors)

  3. The Fukushima Daiichi Accident. Technical Volume 4/5. Radiological Consequences

    International Nuclear Information System (INIS)

    2015-08-01

    This technical volume describes the consequences associated with radioactivity and radiation from the accident at the Fukushima Daiichi nuclear power plant (NPP) for people and the environment. A number of international organizations have already issued reports on the potential health and environmental consequences of the accident, notably the World Health Organization (WHO) and the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR). The intention of the assessments presented in this volume is to build on their work, using more recent data where available. Quantitative information arising from both personal and environmental monitoring has been provided by the Government of Japan. Section 4.1 provides the best estimates of the magnitude and form of radioactive releases during the accident to the atmosphere and directly into the surrounding sea. It also explains the movement of the discharged radionuclides through air and water and the eventual deposition of the atmospheric activity on land in Japan and other countries worldwide, as well as on the open oceans. The goal is to provide a consolidated repository of information on releases to, and levels of radionuclides in, the environment. Some of this information is used in the analyses in subsequent sections of this volume. Section 4.2 gives an overview of exposures to the main groups of emergency workers at the Fukushima Daiichi NPP, to groups of off-site workers and to members of the public. Where sufficient data are available, average effective dose and thyroid equivalent doses derived from personal measurements are compared with the results of previous assessments for specific locations, population groups and time periods. Section 4.3 summarizes relevant aspects of the system of radiation protection in place at the time of the accident. It includes an overview of the legislation and guidance used to implement the radiation protection framework in Japan. This section also provides a

  4. Prevention of the Causes and Consequences of Criticality Accidents: Measures Adopted in France; Prevention des Causes et des Consequences d'un Accident de Criticite: Solutions Adoptees en France

    Energy Technology Data Exchange (ETDEWEB)

    Fruchard, Y.; Lavie, J. -M. [Commissariat a l' Energie Atomique, Paris (France)

    1966-05-15

    It is important to guard against the risk of criticality accidents by seeking to prevent their occurrence through the elimination of their causes and also by taking steps to provide against their consequences. These two aspects are closely linked since the efforts made to elaborate preventive procedures are dictated by the importance of the repercussions which such accidents are liable to have in the human, economic and psychological spheres. The criticality accidents which have occurred in the nuclear industry, though small in number, do reveal the imperfect nature of the techniques adopted to prevent them, and they constitute the only available realistic basis for evaluating their consequences and developing suitable precautionary techniques. The authors give a detailed analysis of the known causes and consequences of past criticality accidents and on this basis make a number of comments in connection with the validity of traditional safety criteria, the probability of accidents for different types of operation, the characteristic accidents capable of serving as models, and the extent of possible radiological consequences. The measures adopted in France to limit the consequences of a possible criticality accident (location, design and lay-out of installations, accident detection dosimetry for exposed personnel) are briefly described after a short account of the criteria used in deciding on them. Finally, the authors discuss the economic implications of adopting particular precautionary measures and of applying them uniformly, taking due account of the question of reliability. (author) [French] II est important de se proteger contre les risques d'accidents de criticite en tentant, d'une part, de prevenir les accidents eux-memes par l'elimination de leurs causes, d'autre part, de parer a leurs consequences. Ces deux aspects sont tres lies: l'effort portant sur la prevention des accidents decoule de l'importance de leurs consequences sur les plans humain, economique

  5. Depressurization as an accident management strategy to minimize the consequences of direct containment heating

    International Nuclear Information System (INIS)

    Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A.

    1990-10-01

    Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of ∼922 K (1200 degree F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs

  6. Depressurization as an accident management strategy to minimize the consequences of direct containment heating

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-10-01

    Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of {approximately}922 K (1200{degree}F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs.

  7. Hypertension and left ventricular hypertrophy in liquidators of consequences of the Chernobyl nuclear accident

    International Nuclear Information System (INIS)

    Shal'nova, S.A.; Smolenskij, A.V.; Shamarin, V.M.; Ehktova, T.V.; Berzak, N.V.; Zemtsova, N.A.; Timofeeva, S.G.; Zhavoronkova, E.A.; Muromtseva, G.A.; Arkad'eva, M.A.; Deev, A.D.

    1998-01-01

    Echocardiography was used for the study of prevalence of left ventricular hypertrophy in 839 liquidators of consequences of the Chernobyl accident. Prevalence of left ventricular hypertrophy (left ventricular myocardial mass 134 g/m 2 ) was 10.3, 13.4 and 22.5 % in liquidators with normal blood pressure, borderline hypertension and hypertension, respectively. Liquidators with normal blood pressure had significantly greater left ventricular myocardial mass than normotensive men from general population while liquidators and non liquidators with hypertension had equal values of this parameter [ru

  8. Medico-demographic criteria in estimating the consequences of the Chernobyl accident

    Energy Technology Data Exchange (ETDEWEB)

    Linge, I I; Melikhova, I A; Pavlovski, O [Nuclear Safety Inst., Academy of Sciences, Moscow (Russian Federation)

    1997-09-01

    Correct comparison of population statistics in affected and unaffected areas prior to and after the accident allows to detect any noticeable deviations in basic medico-demographic parameters in contaminated territories from common trends. In view of that when in 1990 in Nuclear Safety Institute a start has been made on construction of an information support system for government and regional executives to overcome the consequences of the Chernobyl disaster a specialized data bank on demography and medical statistics (MDBD) was created. 12 refs, 7 figs, 8 tabs.

  9. Medico-demographic criteria in estimating the consequences of the Chernobyl accident

    International Nuclear Information System (INIS)

    Linge, I.I.; Melikhova, I.A.; Pavlovski, O.

    1997-01-01

    Correct comparison of population statistics in affected and unaffected areas prior to and after the accident allows to detect any noticeable deviations in basic medico-demographic parameters in contaminated territories from common trends. In view of that when in 1990 in Nuclear Safety Institute a start has been made on construction of an information support system for government and regional executives to overcome the consequences of the Chernobyl disaster a specialized data bank on demography and medical statistics (MDBD) was created. 12 refs, 7 figs, 8 tabs

  10. Functional status of thyroid of Chernobyl accident consequences liquidators after 10 years after disaster

    International Nuclear Information System (INIS)

    Ilieva, A.A.

    1997-01-01

    Analysis of Chernobyl accident consequences liquidators' complaints is carried out and their clinical surveillance is conducted as well. Pronounced disorders of neuro-immune-endocrine system of the liquidators majority and ahill reflex latency half-period prolongation have been observed. By data of ultrasonic study the majority of examined ones have thyroid hyperplasia without features of chronic autoimmune inflammation and formation of adenomatous knots. Thyroid levels of hormone concentration are reduced. There is direct dependence between hormones levels and irradiation dose. The is concluded, that in delayed period after irradiation by low doses the hypo-function status of thyroid is observing

  11. Analysis of unmitigated large break loss of coolant accidents using MELCOR code

    Science.gov (United States)

    Pescarini, M.; Mascari, F.; Mostacci, D.; De Rosa, F.; Lombardo, C.; Giannetti, F.

    2017-11-01

    In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation.

  12. 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

  13. Analysis of ASTEC code adaptability to severe accident simulation for CANDU type reactors

    International Nuclear Information System (INIS)

    Constantin, Marin; Rizoiu, Andrei

    2008-01-01

    In order to prepare the adaptation of the ASTEC code to CANDU NPP severe accident analysis two kinds of activities were performed: - analyses of the ASTEC modules from the point of view of models and options, followed by CANDU exploratory calculation for the appropriate modules/models; - preparing the specifications for ASTEC adaptation for CANDU NPP. The paper is structured in three parts: - a comparison of PWR and CANDU concepts (from the point of view of severe accident phenomena); - exploratory calculations with some ASTEC modules- SOPHAEROS, CPA, IODE, CESAR, DIVA - for CANDU type reactors specific problems; - development needs analysis - algorithms, methods, modules. (authors)

  14. Knowledge resources on the Chernobyl accident and its consequences in the INIS Database

    International Nuclear Information System (INIS)

    Negeri, B.

    2005-07-01

    Literature on the Chernobyl accident and its consequences is an important subject covered by the International Nuclear Information System (INIS) Database. The INIS Database contains 19872 bibliographic records and 8400 full text documents on this subject from 1986 up to 04/2005. A bibliometric study of these records was made to generate statistical summaries that characterise, in general terms, the intellectual content of the records and the nature of the records in terms of its major bibliographic attributes. Environmental aspects and human health constitute the two dominant subjects with a respective contribution of 49% and 38%. The rest is evenly divided among legal aspects, reactor safety and socio-economic impacts of the accident. The three countries that are most affected by the accident, namely Ukraine, Russian Federation and Belarus contributed 44% of the total input. 57% of the literature analysed are conference papers and reports while 25% are journal articles. Most of the documents were written in English (47%) and in Russian (36%). Seven percent of the publications were written in German. (author)

  15. Assessment of the potential consequences of a large primary to secondary leakage accident. Final report

    International Nuclear Information System (INIS)

    D'Auria, F.S.; Sartmadjiev, A.; Spalj, S.; Macek, J.; Kantee, H.; Elter, J.; Kostka, P.; Bukin, N.; Alexandrov, A.G.; Kristof, M.; Kvizda, B.; Matejovic, P.; Makihara, Y.

    2006-01-01

    The present paper discusses one of the IAEA's Coordinated Research Projects (CRPs). The CRP was started in 2003 to evaluate complex phenomena of primary to secondary leakage (PRISE) accidents for WWER-440 reactors. The first Research Coordination Meeting (RCM), held in March 2003, identified the possible consequences of PRISE accidents (radioactive release to the atmosphere, pressurized thermal shock, boron dilution, loss of integrity of secondary systems and severe accidents) and designated six task groups to evaluate these, as well as uncertainties associated with PRISE analyses. The second RCM, held in March 2004, discussed the preliminary results of each task group and addressed the main safety concerns related to PRISE phenomena as well as providing recommendations on modelling for PRISE analyses and on operator actions. The third RCM, held in March 2005, discussed the results of the work performed in 2004. The CRP was concluded in 2005. Publication of the final results of the CRP is planned as an IAEA TECDOC. The paper provides a review of the final results of the project. (author)

  16. Research on consequence analysis method for probabilistic safety assessment of nuclear fuel facilities (5). Evaluation method and trial evaluation of criticality accident

    International Nuclear Information System (INIS)

    Yamane, Yuichi; Abe, Hitoshi; Nakajima, Ken; Hayashi, Yoshiaki; Arisawa, Jun; Hayami, Satoru

    2010-01-01

    A special committee of 'Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)' was organized by the Atomic Energy Society of Japan (AESJ) under the entrustment of Japan Atomic Energy Agency (JAEA). The committee aims to research on the state-of-the-art consequence analysis method for the Probabilistic Safety Assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objectives of this research are to obtain information useful for establishing quantitative performance objectives and to demonstrate risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on the consequence analysis method for postulated accidents with potentially large consequences in NFFs, e.g., events of criticality, spill of molten glass, hydrogen explosion, boiling of radioactive solution and fire (including the rapid decomposition of TBP complexes), resulting in the release of radioactive materials to the environment. The results of the research were summarized in a series of six reports, which consist of a review report and five technical ones. In this report, the evaluation methods of criticality accident, such as simplified methods, one-point reactor kinetics codes and quasi-static method, were investigated and their features were summarized to provide information useful for the safety evaluation of NFFs. In addition, several trial evaluations were performed for a hypothetical scenario of criticality accident using the investigated methods, and their results were compared. The release fraction of volatile fission products in a criticality accident was also investigated. (author)

  17. Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104)

    International Nuclear Information System (INIS)

    Kress, T.S.

    1985-04-01

    The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time

  18. Development of a dose assessment computer code for the NPP severe accident

    International Nuclear Information System (INIS)

    Cheong, Jae Hak

    1993-02-01

    A real-time emergency dose assessment computer code called KEDA (KAIST NPP Emergency Dose Assessment) has been developed for the NPP severe accident. A new mathematical model which can calculate cloud shine has been developed and implemented in the code. KEDA considers the specific Korean situations(complex topography, orientals' thyroid metabolism, continuous washout, etc.), and provides functions of dose-monitoring and automatic decision-making. To verify the code results, KEDA has been compared with an NRC officially certified code, RASCAL, for eight hypertical accident scenarios. Through the comparison, KEDA has been proved to provide reasonable results. Qualitative sensitivity analysis also the been performed for potentially important six input parameters, and the trends of the dose v.s. down-wind distance curve have been analyzed comparing with the physical phenomena occurred in the real atmosphere. The source term and meteorological conditions are turned out to be the most important input parameters. KEDA also has been applied to simulate Kori site and a hyperthetical accident with semi-real meteorological data has been simulated and analyzed

  19. Radioecological consequences of potential accident in Norwegian coastal waters. Uncertainties and knowledges gaps in methodology

    International Nuclear Information System (INIS)

    Iosjpe, M.; Reistad, O.; Brown, J.; Jaworska, A.; Amundsen, I.

    2007-01-01

    Complete text of publication follows. A potential accident involving the transport of spent nuclear fuel along the Norwegian coastline has been choosen for evaluation of a dose assessment methodology. The accident scenario assumes that the release the radioactivity takes place under water and that there is free exchange of water between the spent fuel and the sea. The inventory has been calculated using the ORIGEN programme. Radioecological consequences are provided by the NRPA compartment model which includes the processes of advection of radioactivity between compartments and water-sediment interactions. The contamination of biota is further calculated from the radionuclide concentrations in filtered seawater in the different water regions. Doses to man are calculated on the basis of radionuclide concentrations in marine organisms, water and sediment and dose conversion factors. Collective dose-rates to man, doses to the critical groups, concentration of radionuclides in biota/sea-foods and doses to marine organisms were calculated through the evaluation of radioecological consequences after accidents. Results of calculations indicate that concentrations of radionuclides for some marine organisms can exceed guideline levels. At the same time collective dose rates to man as well as doses to a critical group are not higher than guideline levels. Comparison of results from calculations with provisional benchmark values suggests that doses to biota are in most cases unlikely to be of concern. However, to some marine organisms can be much higher than the screening dose of 10 μGyh over long periods. It is apparent that water-sediment distribution coefficients and concentration factors constitute the main sources of uncertainties in the present case. It is important to note that knowledge gaps concerning the influence of relatively low dose to populations of marine organisms over long time periods (many generations) substantially constrain the accessor's ability to

  20. Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code; Verbundprojekt WASA-BOSS: Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt B: Druckwasserreaktor-Stoerfallanalysen unter Verwendung des Severe-Accident-Codes ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

    2017-02-15

    Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

  1. Dispersion of radioactive materials from JRTR following a postulated accident using HOTSPOT code

    International Nuclear Information System (INIS)

    Mistarihi, Qusai M.; Lee, Kwan Hee

    2013-01-01

    Jordan Research and Training Reactor (JRTR) is the first nuclear facility in Jordan. The JRTR is 5 MW, light water moderated and open type pool reactor. In case of an accident, the radioactive materials will be released to the surrounding environment and endanger the people living in the vicinity of the reactor. However, up to now, no study has been published about the dispersion of radioactive materials from JRTR in case of an accident. As preliminary stage for the construction of the JRTR, the dispersion of the radioactive materials from JRTR in case of an accident was studied using HOTSOT code. The result of the report indicates that for ground level release with an average speed of 3.6 m/s of hourly averaged meteorological data for one year with a dominant direction from the west a person located at distance .062 km from the reactor site will receive .25 Sv

  2. Dispersion of radioactive materials from JRTR following a postulated accident using HOTSPOT code

    Energy Technology Data Exchange (ETDEWEB)

    Mistarihi, Qusai M.; Lee, Kwan Hee [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    Jordan Research and Training Reactor (JRTR) is the first nuclear facility in Jordan. The JRTR is 5 MW, light water moderated and open type pool reactor. In case of an accident, the radioactive materials will be released to the surrounding environment and endanger the people living in the vicinity of the reactor. However, up to now, no study has been published about the dispersion of radioactive materials from JRTR in case of an accident. As preliminary stage for the construction of the JRTR, the dispersion of the radioactive materials from JRTR in case of an accident was studied using HOTSOT code. The result of the report indicates that for ground level release with an average speed of 3.6 m/s of hourly averaged meteorological data for one year with a dominant direction from the west a person located at distance .062 km from the reactor site will receive .25 Sv.

  3. Fuel relocation modeling in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Tentner, A.M.; Miles, K.J.

    1985-01-01

    SAS4A is a new code system which has been designed for analyzing the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modeling the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel motion experiment analyses are also presented

  4. Fuel relocation modeling in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Tentner, A.M.; Miles, K.J.; Kalimullah; Hill, D.J.

    1986-01-01

    The SAS4A code system has been designed for the analysis of the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modelong the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel relocation are emphasized. 12 refs

  5. Research activities about the radiological consequences of the Chernobyl NPS accident and social activities to assist the sufferers by the accident

    International Nuclear Information System (INIS)

    Imanaka, T.

    1998-03-01

    The 12th anniversary is coming soon of the accident at the Chernobyl nuclear power station in the former USSR on April 26, 1986. Many issues are, however, still unresolved about the radiological impacts on the environment and people due to the Chernobyl accident. This report contains the results of an international collaborative project about the radiological consequences of the Chernobyl accident, carried out from November 1995 to October 1997 under the research grant of the Toyota foundation. Collaborative works were promoted along with the following 5 sub-themes: 1) General description of research activities in Russia, Belarus and Ukraine concerning the radiological consequences of the accident. 2) Investigation of the current situation of epidemiological studies about Chernobyl in each affected country. 3) Investigation of acute radiation syndrome among inhabitants evacuated soon after the accident from the 30 km zone around the Chernobyl NPS. 4) Overview of social activities to assist the sufferers by the accident in each affected country. 5) Preparation of special reports of interesting studies being carried out in each affected country. The 27 papers are indexed individually. (J.P.N.)

  6. 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.

  7. 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

  8. Techniques and decision making in the assessment of off-site consequences of an accident in a nuclear facility

    International Nuclear Information System (INIS)

    1987-01-01

    This Guide is intended to complement the IAEA's existing technical guidance on emergency planning and preparedness by providing information and practical guidance related to the assessment of off-site consequences of an accident in a nuclear or radioactive materials installation and to the decision making process in implementing protective measures. This Guide contains information on emergency response philosophy, fundamental factors affecting accident consequences, principles of accident assessment, data acquisition and handling, systems, techniques and decision making principles. Many of the accident assessment concepts presented are considerably more advanced than some of those that now pertain in most countries. They could, if properly interpreted, developed and applied, significantly improve emergency response in the early and intermediate phases of an accident. Furthermore, they are considered to be applicable to a broad range of serious nuclear accidents and radiological emergencies. The extent of their application is governed by both the scale of the accident and by the availability of preplanned resources for accident assessment and emergency response. 68 refs, 28 figs, 14 tabs

  9. Development of the assessment of nuclear accident consequences and decision support system in China: status, requirement and recommendations

    International Nuclear Information System (INIS)

    Shi Zhongqi; Wang Xingyu

    2003-01-01

    This paper introduces the status of nuclear accident consequence assessment/development of decision-making support system in China. The basic functions and roles of the consequence assessment/decision-making support system for three levels of nuclear emergency response organization (i.e. national, local offsite and nuclear power plant operator) in China are presented in the paper

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  11. Analysis of fuel behaviour after loss-of-coolant accident with the TESPA-code

    International Nuclear Information System (INIS)

    Keusenhoff, J.

    1981-01-01

    After a loss-of-coolant accident fuel rods go through a phase of high temperature and differential pressure before quenching and initiation of long term cooling. For licensing purpose the highest cladding temperature and the coolability of the core is of interest. The highest temperature is evaluated by a hot channel calculation with conservative assumptions. It gives little information about the status of the entire core. Therefore more detailed information is necessary. TESPA is a fast running code, which uses best-estimate assumptions, considers statistical uncertainties in the input parameters and calculates clad ballooning and rupture. The code is a usefull tool for calculation of channel blockage and cladding rupture

  12. Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

    International Nuclear Information System (INIS)

    Lee, A.G.; Wilkin, G.B.

    1995-01-01

    This report is a compilation of the information submitted by AECL, CIAE, JAERI, ORNL and Siemens in response to a need identified at the 'Workshop on R and D Needs' at the IGORR-3 meeting. The survey compiled information on the national standards applied to the Safety Quality Assurance (SQA) programs undertaken by the participants. Information was assembled for the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods used to verify and validate the codes and libraries. Although the survey was not comprehensive, it provides a basis for exchanging information of common interest to the research reactor community

  13. Modeling of severe accident sequences with the new modules CESAR and DIVA of ASTEC system code

    International Nuclear Information System (INIS)

    Pignet, Sophie; Guillard, Gaetan; Barre, Francois; Repetto, Georges

    2003-01-01

    Systems of computer codes, so-called 'integral' codes, are being developed to simulate the scenario of a hypothetical severe accident in a light water reactor, from the initial event until the possible radiological release of fission products out of the containment. They couple the predominant physical phenomena that occur in the different reactor zones and simulate the actuation of safety systems by procedures and by operators. In order to allow to study a great number of scenarios, a compromise must be found between precision of results and calculation time: one day of accident time should take less than one day of real time to simulate on a PC computer. This search of compromise is a real challenge for such integral codes. The development of the ASTEC integral code was initiated jointly by IRSN and GRS as an international reference code. The latest version 1.0 of ASTEC, including the new modules CESAR and DIVA which model the behaviour of the reactor cooling system and the core degradation, is presented here. Validation of the modules and one plant application are described

  14. Verification of fire and explosion accident analysis codes (facility design and preliminary results)

    International Nuclear Information System (INIS)

    Gregory, W.S.; Nichols, B.D.; Talbott, D.V.; Smith, P.R.; Fenton, D.L.

    1985-01-01

    For several years, the US Nuclear Regulatory Commission has sponsored the development of methods for improving capabilities to analyze the effects of postulated accidents in nuclear facilities; the accidents of interest are those that could occur during nuclear materials handling. At the Los Alamos National Laboratory, this program has resulted in three computer codes: FIRAC, EXPAC, and TORAC. These codes are designed to predict the effects of fires, explosions, and tornadoes in nuclear facilities. Particular emphasis is placed on the movement of airborne radioactive material through the gaseous effluent treatment system of a nuclear installation. The design, construction, and calibration of an experimental ventilation system to verify the fire and explosion accident analysis codes are described. The facility features a large industrial heater and several aerosol smoke generators that are used to simulate fires. Both injected thermal energy and aerosol mass can be controlled using this equipment. Explosions are simulated with H 2 /O 2 balloons and small explosive charges. Experimental measurements of temperature, energy, aerosol release rates, smoke concentration, and mass accumulation on HEPA filters can be made. Volumetric flow rate and differential pressures also are monitored. The initial experiments involve varying parameters such as thermal and aerosol rate and ventilation flow rate. FIRAC prediction results are presented. 10 figs

  15. The coupling algorithm between fuel pin and coolant channel in the European Accident Code EAC-2

    International Nuclear Information System (INIS)

    Goethem, G. van; Lassmann, K.

    1989-01-01

    In the field of fast breeder reactors the Commission of the European Communities (CEC) is conducting coordination and harmonisation activities as well as its own research at the CEC's Joint Research Centre (JRC). The development of the modular European Accident Code (EAC) is a typical example of concerted action between EC Member States performed under the leadership of the JRC. This computer code analyzes the initiation phase of low-probability whole-core accidents in LMFBRs with the aim of predicting the rapidity of sodium voiding, the mode of pin failure, the subsequent fuel redistribution and the associated energy release. This paper gives a short overview on the development of the EAC-2 code with emphasis on the coupling mechanism between the fuel behaviour module TRANSURANUS and the thermohydraulics modules which can be either CFEM or BLOW3A. These modules are also briefly described. In conclusion some numerical results of EAC-2 are given: they are recalculations of an unprotected LOF accident for the fictitious EUROPE fast breeder reactor which was earlier analysed in the frame of a comparative exercise performed in the early 80s and organised by the CEC. (orig.)

  16. Evaluating the consequences of loss of flow accident for a typical VVER-1000 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mirvakili, S.M.; Safaei, S. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering, School of Mechanical Engineering; Faghihi, F. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Safety Research Center

    2010-07-01

    The loss of coolant flow in a nuclear reactor can result from a mechanical or electrical failure of the coolant pump. If the reactor is not tripped promptly, the immediate effect is a rapid increase in coolant temperature, decrease in minimum departure from nucleate boiling ratio (DNBR) and fuel damage. This study evaluated the shaft seizure of a reactor coolant pump in a VVER-1000 nuclear reactor. The locked rotor results in rapid reduction of flow through the affected reactor coolant loop and in turn leads to an increase in the primary coolant temperature and pressure. The analysis was conducted with regard for superimposing loss of power to the power plant at the initial accident moment. The required transient functions of flow, pressure and power were obtained using system transient calculations applied in COBRA-EN computer code in order to calculate the overall core thermal-hydraulic parameters such as temperature, critical heat flux and DNBR. The study showed that the critical period for the locked rotor accident is the first few seconds during which the maximum values of pressure and temperature are reached. 10 refs., 1 tab., 3 figs.

  17. Application of the accident consequences model of the German risk study to assessments of accident risks in different types of nuclear power plants

    International Nuclear Information System (INIS)

    Ehrhardt, J.; Bayer, A.

    1982-01-01

    Within the scope of the 'German Risk Study for Nuclear Power Plants' (Phase A) the accident consequence model UFOMOD was developed in the Karlsruhe Nuclear Research Center. This model originally developed for pressurized water reactors has now been extended in order to obtain results about accidental releases of activity from fast breeder and high-temperature reactors, too. (RW) [de

  18. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix VI. Calculation of reactor accident consequences. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the radioactive releases from the containment following accidents; radioactive inventory of the reactor core; atmospheric dispersion; reactor sites and meteorological data; radioactive decay and deposition from plumes; finite distance of plume travel; dosimetric models; health effects; demographic data; mitigation of radiation exposure; economic model; and calculated results with consequence model.

  19. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix VI. Calculation of reactor accident consequences

    International Nuclear Information System (INIS)

    1975-10-01

    Information is presented concerning the radioactive releases from the containment following accidents; radioactive inventory of the reactor core; atmospheric dispersion; reactor sites and meteorological data; radioactive decay and deposition from plumes; finite distance of plume travel; dosimetric models; health effects; demographic data; mitigation of radiation exposure; economic model; and calculated results with consequence model

  20. The radiological consequences of degraded core accidents for the Sizewell PWR The impact of adopting revised frequencies of occurrence

    CERN Document Server

    Kelly, G N

    1983-01-01

    The radiological consequences of degraded core accidents postulated for the Sizewell PWR were assessed in an earlier study and the results published in NRPB-R137. Further analyses have since been made by the Central Electricity Generating Board (CEGB) of degraded core accidents which have led to a revision of their predicted frequencies of occurrence. The implications of these revised frequencies, in terms of the risk to the public from degraded core accidents, are evaluated in this report. Increases, by factors typically within the range of about 1.5 to 7, are predicted in the consequences, compared with those estimated in the earlier study. However, the predicted risk from degraded core accidents, despite these increases, remains exceedingly small.

  1. Coupling Computer Codes for The Analysis of Severe Accident Using A Pseudo Shared Memory Based on MPI

    International Nuclear Information System (INIS)

    Cho, Young Chul; Park, Chang-Hwan; Kim, Dong-Min

    2016-01-01

    As there are four codes in-vessel analysis code (CSPACE), ex-vessel analysis code (SACAP), corium behavior analysis code (COMPASS), and fission product behavior analysis code, for the analysis of severe accident, it is complex to implement the coupling of codes with the similar methodologies for RELAP and CONTEMPT or SPACE and CAP. Because of that, an efficient coupling so called Pseudo shared memory architecture was introduced. In this paper, coupling methodologies will be compared and the methodology used for the analysis of severe accident will be discussed in detail. The barrier between in-vessel and ex-vessel has been removed for the analysis of severe accidents with the implementation of coupling computer codes with pseudo shared memory architecture based on MPI. The remaining are proper choice and checking of variables and values for the selected severe accident scenarios, e.g., TMI accident. Even though it is possible to couple more than two computer codes with pseudo shared memory architecture, the methodology should be revised to couple parallel codes especially when they are programmed using MPI

  2. Coupling Computer Codes for The Analysis of Severe Accident Using A Pseudo Shared Memory Based on MPI

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young Chul; Park, Chang-Hwan; Kim, Dong-Min [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    As there are four codes in-vessel analysis code (CSPACE), ex-vessel analysis code (SACAP), corium behavior analysis code (COMPASS), and fission product behavior analysis code, for the analysis of severe accident, it is complex to implement the coupling of codes with the similar methodologies for RELAP and CONTEMPT or SPACE and CAP. Because of that, an efficient coupling so called Pseudo shared memory architecture was introduced. In this paper, coupling methodologies will be compared and the methodology used for the analysis of severe accident will be discussed in detail. The barrier between in-vessel and ex-vessel has been removed for the analysis of severe accidents with the implementation of coupling computer codes with pseudo shared memory architecture based on MPI. The remaining are proper choice and checking of variables and values for the selected severe accident scenarios, e.g., TMI accident. Even though it is possible to couple more than two computer codes with pseudo shared memory architecture, the methodology should be revised to couple parallel codes especially when they are programmed using MPI.

  3. Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Chellapandi, P.; Velusamy, K.; Chetal, S.C.; Bhoje, S.B.; Lal, H.; Sethi, V.S.

    2003-01-01

    The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

  4. Analysis of uncertainties caused by the atmospheric dispersion model in accident consequence assessments with UFOMOD

    International Nuclear Information System (INIS)

    Fischer, F.; Ehrhardt, J.

    1988-06-01

    Various techniques available for uncertainty analysis of large computer models are applied, described and selected as most appropriate for analyzing the uncertainty in the predictions of accident consequence assessments. The investigation refers to the atmospheric dispersion and deposition submodel (straight-line Gaussian plume model) of UFOMOD, whose most important input variables and parameters are linked with probability distributions derived from expert judgement. Uncertainty bands show how much variability exists, sensitivity measures determine what causes this variability in consequences. Results are presented as confidence bounds of complementary cumulative frequency distributions (CCFDs) of activity concentrations, organ doses and health effects, partially as a function of distance from the site. In addition the ranked influence of the uncertain parameters on the different consequence types is shown. For the estimation of confidence bounds it was sufficient to choose a model parameter sample size of n (n=59) equal to 1.5 times the number of uncertain model parameters. Different samples or an increase of sample size did not change the 5%-95% - confidence bands. To get statistically stable results of the sensitivity analysis, larger sample sizes are needed (n=100, 200). Random or Latin-hypercube sampling schemes as tools for uncertainty and sensitivity analyses led to comparable results. (orig.) [de

  5. Fukushima, one year later. Initial analyses of the accident and its consequences

    International Nuclear Information System (INIS)

    2012-01-01

    The earthquake of magnitude 9 of March 11, 2011 with an epicenter 80 km east of the Japanese island of Honshu, and the subsequent tsunami, severely affected the region of Tohoku, with major consequences for its population and infrastructure. Devastating the site of the Fukushima Dai-ichi nuclear power plant, these natural events were the cause of the core meltdowns of three nuclear reactors and the loss of cooling of several spent fuel pools. Explosions also occurred in reactor buildings 1 through 4 due to hydrogen produced during fuel degradation. Very significant radioactive releases into the environment took place. The accident was classified level 7 on the International Nuclear Event Scale (INES). This report provides an assessment and perspective on the information gathered by IRSN during the first twelve months following the disaster in an effort to understand the condition of the installations, evaluate the releases and analyze and evaluate the consequences of the accident on workers and the impact on the population and the environment. On the basis of available information, the report provides an initial analysis of the chain of events. It should be noted that a year after the accident, the full sequence of events is still not understood. Operating experience feedback from the 1979 Three Mile Island accident in the United States, in which reactor core damage was not confirmed until 1986, suggests that it may be several years before a detailed scenario can be constructed of the accident that led to radioactive releases. It will require access to the damaged installations. The situation at the site remains dangerous (reactor pressure vessels and containments are not leak-tight, diffuse releases, etc.). If it has significantly improved as a result of the significant resources deployed by the Tokyo Electro Power Company (TEPCO) to regain control of the installations, this effort must continue over the long term to begin evacuation of fuel from pools (in two

  6. Chernobyl NPP accident consequences cleaning up participants in Ukraine -health status epidemiologic study main results

    International Nuclear Information System (INIS)

    Buzunov, V.; Omelyanetz, N.; Strapko, N.; Ledoschuck, B.; Krasnikova, L.; Kartushin, G.

    1996-01-01

    The Epidemiologic Studies System for Chernobyl NPP Accident consequences cleaning up participants (CNPP ACCP) health status was worked out and than improving in Ukraine after the CNPP Accident. The State Register of Ukraine both with several other Registers are the organizational, methodological and informational basis here. The ACCP health status worsening ,-was registered in dynamics through the post-accidental period i.e. the nervous system, digestive system, blood circulation system, respiratory system, bone-muscular system, endocrine and genitourinary systems chronic non-tumoral pathology both with mental disorders amount increase. In cohort study the differences of morbidity formation were fixed among emergency workers with different radiation exposure doses. The dependence of leukemia morbidity on presence in 30-km zone duration was noticed, it's access manifested 5 years after the participance in ACC. The ACCP disablement increase with main reason of general somatic diseases, and annual mortality growth are registered. But that doesn't exceed the mortality rate among population of working age in Ukraine

  7. A model for the calculation of the off-site economic consequences of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Gallego, E.; Alonso, A.

    1988-01-01

    The off-site economic cost of nuclear reactor accidents will depend on the countermeasures adopted to reduce its radiological impact. The assessment of the direct costs of emergency countermeasures (evacuation, early relocation and food disposal) as well as those of long-term protective actions (food disposal, decontamination or interdiction) is the objective of a model under development, with the sponsorship of the CEC Radiation Protection Programme, called MECA (Model for assessing the Economic Consequences of Accidents). The meteorological and socio-economical peculiarities of each site studied will be taken into account, by means of a flexible meteorological sampling scheme, which considers the geographical distribution of population and economic centers, and a data-base, compatible with the existing European grid, that contains the population distribution and the economic characteristics of the environs of the site to be studied with more detail near the reactor. The paper summarizes the particular models which will be included in MECA and shows the importance of site-specific adaptable modelling for economic risk evaluation

  8. The radiological consequences in the USSR from the Chernobyl accident: Description of the scheme of implementation

    International Nuclear Information System (INIS)

    Dreicer, Mona

    2004-01-01

    After October 1989 the Government of the USSR requested the IAEA to organize an international assessment of the 'concept which the USSR has evolved to enable the population to live safely in areas affected by radioactive contamination following the Chernobyl accident, and an evaluation of the effectiveness of the steps taken in these areas to safeguard the health of the population'. The IAEA responded positively to this request for special assistance. The IAEA is carried out an extensive international project involving over 100 experts who assessed the human health and environmental consequences in the affected areas of Byelorussia, the Ukraine, and the Russian Federation, and evaluate measures taken by Soviet authorities to protect the population and the environment

  9. Health effects models for off-site radiological consequence analysis on nuclear reactor accidents (II)

    Energy Technology Data Exchange (ETDEWEB)

    Homma, Toshimitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takahashi, Tomoyuki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Yonehara, Hidenori [National Inst. of Radiological Sciences, Chiba (Japan)] [eds.

    2000-12-01

    This report is a revision of JAERI-M 91-005, 'Health Effects Models for Off-Site Radiological Consequence Analysis of Nuclear Reactor Accidents'. This revision provides a review of two revisions of NUREG/CR-4214 reports by the U.S. Nuclear Regulatory Commission which is the basis of the JAERI health effects models and other several recent reports that may impact the health effects models by international organizations. The major changes to the first version of the JAERI health effects models and the recommended parameters in this report are for late somatic effects. These changes reflect recent changes in cancer risk factors that have come from longer followup and revised dosimetry in major studies on the Japanese A-bomb survivors. This report also provides suggestions about future revisions of computational aspects on health effects models. (author)

  10. Health effects models for off-site radiological consequence analysis on nuclear reactor accidents (II)

    International Nuclear Information System (INIS)

    Homma, Toshimitsu

    2000-12-01

    This report is a revision of JAERI-M 91-005, 'Health Effects Models for Off-Site Radiological Consequence Analysis of Nuclear Reactor Accidents'. This revision provides a review of two revisions of NUREG/CR-4214 reports by the U.S. Nuclear Regulatory Commission which is the basis of the JAERI health effects models and other several recent reports that may impact the health effects models by international organizations. The major changes to the first version of the JAERI health effects models and the recommended parameters in this report are for late somatic effects. These changes reflect recent changes in cancer risk factors that have come from longer followup and revised dosimetry in major studies on the Japanese A-bomb survivors. This report also provides suggestions about future revisions of computational aspects on health effects models. (author)

  11. Introductory remarks by the Chairman. [Session 1: Environmental and health consequences of the Chernobyl accident

    International Nuclear Information System (INIS)

    Bennett, B.G.

    2005-01-01

    Many scientists as well as representatives from UN organizations and governments of affected regions participated in the work of the Chernobyl Forum. Several meetings of the Forum were necessary to initiate the work and monitor the progress of the expert groups. Two expert groups formulated comprehensive reports - one on environmental issues, organized by the IAEA, and one on health issues, organized by the WHO. Experts from throughout the world were invited to contribute to these evaluations. The representatives of governments and the staff of international organizations then reviewed the results of these groups to be sure that the reviews were complete and the evaluations reasonable, so that they could serve as the basis for consensus agreements and effective recommendations for further dealing with the consequences of the accident

  12. The consequences of radioactive contamination of forest ecosystems due to Chernobyl accident

    International Nuclear Information System (INIS)

    Tikhomirov, F.A.; Shcheglova, A.I.

    1997-01-01

    The effect of forests on the radionuclide primary distribution in different components of the contaminated ecosystems is considered by the example of Chernobyl accident. A basic mathematical model is developed describing 137 Cs biogeochemical cycling under conditions of quasi-steady state radionuclide redistribution in the ecosystem. Forest ecosystems are proved to diminish radionuclide migration in the environment, and forest should be regarded as an important sanitary factor. The contribution of contaminated forests and forest products to the total irradiation dose to local population is estimated. Special countermeasures are elaborated in order to diminish unfavorable consequences of forest radioactive contamination. A long-term dynamics of radioactive situation in the forest ecosystems in forecasted and further studies on the subject are drafted

  13. The application of the health effects models to the severe accident consequence analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Yang Ling; Yeung, M.R.

    1998-01-01

    Health Effect Model (HEM) is an important model used in the analysis of severe accidents consequence of the Nuclear Power Plants (NPP). The accuracy of HEM affects the reliability of the assessment for the accidents consequences, and furthermore, the effectiveness of the emergency countermeasures taken for the health protection of the public around the NPPs. Based on the NUREG/CR4214 series reports, the paper sets appropriate parameters for HEM by studying both early and late HEMs used for domestic NPP accident consequence analysis. In the study, the Guangdong Daya Bay NPP is chosen as an example study to calculate the health risk of the Hong Kong population caused by Daya Bay NPP

  14. The Development of Severe Accident Codes at IRSN and Their Application to Support the Safety Assessment of EPR

    International Nuclear Information System (INIS)

    Caroli, Cataldo; Bleyer, Alexandre; Bentaib, Ahmed; Chatelard, Patrick; Cranga, Michel; Van Dorsselaere, Jean-Pierre

    2006-01-01

    IRSN uses a two-tier approach for development of codes analysing the course of a hypothetical severe accident (SA) in a Pressurized Water Reactor (PWR): on one hand, the integral code ASTEC, jointly developed by IRSN and GRS, for fast-running and complete analysis of a sequence; on the other hand, detailed codes for best-estimate analysis of some phenomena such as ICARE/CATHARE, MC3D (for steam explosion), CROCO and TONUS. They have been extensively used to support the level 2 Probabilistic Safety Assessment of the 900 MWe PWR and, in general, for the safety analysis of the French PWR. In particular the codes ICARE/CATHARE, CROCO, MEDICIS (module of ASTEC) and TONUS are used to support the safety assessment of the European Pressurized Reactor (EPR). The ICARE/CATHARE code system has been developed for the detailed evaluation of SA consequences in a PWR primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermal-hydraulics French code CATHARE2. The CFD code CROCO describes the corium flow in the spreading compartment. Heat transfer to the surrounding atmosphere and to the basemat, leading to the possible formation of an upper and lower crust, basemat ablation and gas sparging through the flow are modelled. CROCO has been validated against a wide experimental basis, including the CORINE, KATS and VULCANO programs. MEDICIS simulates MCCI (Molten-Corium-Concrete-Interaction) using a lumped-parameter approach. Its models are being continuously improved through the interpretation of most MCCI experiments (OECD-CCI, ACE...). The TONUS code has been developed by IRSN in collaboration with CEA for the analysis of the hydrogen risk (both distribution and combustion) in the reactor containment. The analyses carried out to support the EPR safety assessment are based on a CFD formulation. At this purpose a low-Mach number multi-component Navier-Stokes solver is used to analyse the hydrogen distribution. Presence of air, steam and

  15. MABEL-1. A code to analyse cladding deformation in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Bowring, R.W.; Cooper, C.A.

    1978-06-01

    The MABEL-1 code has been written to investigate the deformation, of fuel pin cladding and its effects on fuel pin temperature transients during a loss-of-coolant accident. The code considers a single fuel pin with heated fuel concentric within the cladding. The fuel pin temperature distribution is evaluated using a one-dimensional conduction model with heat transfer to the coolant represented by an input set of heat transfer coefficients. The cladding deformation is calculated using the code CANSWEL, which assumes all strain to be elastic or creep and models the creep under a multi-axial stress system by a spring/dashpot combination undergoing alternate relaxation and elastic strain. (author)

  16. NAUAHYGROS - A code for calculating aerosol behavior in nuclear power plant containments following a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Sher, R. [Rudolph Sher Associates, Stanford, CA (United States); Li, J. [Polestar Applied Technology, Inc., Los Altos, CA (United States)

    1995-02-01

    NAUAHYGROS is a computer code to calculate the behavior of fission product and other aerosol particles in the containment of a nuclear reactor following a severe accident. It is an extension of the German code NAUA, which has been in widespread use for many years. Early versions of NAUA treated various aerosol phenomena in dry atmospheres, including aerosol agglomeration, diffusion (plateout), and settling processes. Later versions added treatments of steam condensation on particles in saturated or supersaturated containment atmospheres. The importance of these condensation effects on aerosol removal rates was demonstrated in large scale simulated containment tests. The additional features incorporated in NAUAHYGROS include principally a treatment of steam condensation on hygroscopic aerosols, which can grow as a result of steam condensation even in superheated atmospheres, and improved modelling of steam condensation on the walls of the containment. The code has been validated against the LACE experiments.

  17. Development of Evaluation Technology for Hydrogen Combustion in containment and Accident Management Code for CANDU

    International Nuclear Information System (INIS)

    Kim, S. B.; Kim, D. H.; Song, Y. M.

    2011-08-01

    For a licensing of nuclear power plant(NPP) construction and operation, the hydrogen combustion and hydrogen mitigation system in the containment is one of the important safety issues. Hydrogen safety and its control for the new NPPs(Shin-Wolsong 1 and 2, Shin-Ulchin 1 and 2) have been evaluated in detail by using the 3-dimensional analysis code GASFLOW. The experimental and computational studies on the hydrogen combustion, and participations of the OEDE/NEA programs such as THAI and ISP-49 secures the resolving capabilities of the hydrogen safety and its control for the domestic nuclear power plants. ISAAC4.0, which has been developed for the assessment of severe accident management at CANDU plants, was already delivered to the regulatory body (KINS) for the assessment of the severe accident management guidelines (SAMG) for Wolsong units 1 to 4, which are scheduled to be submitted to KINS. The models for severe accident management strategy were newly added and the graphic simulator, CAVIAR, was coupled to addition, the ISAAC computer code is anticipated as a platform for the development and maintenance of Wolsong plant risk monitor and Wolsong-specific SAMG

  18. A2 Code - Internal Accident Report. Does it ring a bell?

    CERN Document Server

    HSE Unit

    2015-01-01

    A2 Code* - It is under this designation (used by the CERN community) that the form for internal accident reports is hidden. More specifically it refers to the CERN Safety Code A2 “Reporting of Accidents and Near Misses” (EDMS: 335502 or here via the official Safety Rules website).   Which events should be declared? All accidental events, which cause or could have caused injuries or damage to property or the environment, must be reported especially if they involve: a) a member of the personnel, visitor, temporary labourer or contractor if it occurred on the CERN site or between sites. b) a member of the personnel if it occurred while commuting or during duty travel. Who can fill in the report? The reporting of occurred accidents or near misses should be made by the person involved or by any direct or indirect witness of the event as soon as possible after the event. Contribute to the improvement of Safety within the Organizatio...

  19. Social aspects in evaluation of health status of subjects who participated in liquidation of radiation accident consequences

    International Nuclear Information System (INIS)

    Tukov, A.R.; Kleev, N.A.; Shafranskij, I.L.

    2000-01-01

    The morbidity rate of the Russian atomic industry workers, the liquidators of ChNPP accident consequences and their future life span shorting with an account of their social status are evaluated. Tentative and standard morbidity values were calculated with an account of various social groups of the liquidators. Intensive values of the man-year losses were used in the methodology for evaluating the vital potential losses. The study results indicated considerable morbidity difference in certain diseases by the persons of various social groups, who took part in liquidation of the ChNPP accident consequences [ru

  20. Development of severe accident analysis code - A study on the molten core-concrete interaction under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Chang Hyun; Lee, Byung Chul; Huh, Chang Wook; Kim, Doh Young; Kim, Ju Yeul [Seoul National University, Seoul (Korea, Republic of)

    1996-07-01

    The purpose of this study is to understand the phenomena of the molten core/concrete interaction during the hypothetical severe accident, and to develop the model for heat transfer and physical phenomena in MCCIs. The contents of this study are analysis of mechanism in MCCIs and assessment of heat transfer models, evaluation of model in CORCON code and verification in CORCON using SWISS and SURC Experiments, and 1000 MWe PWR reactor cavity coolability, and establishment a model for prediction of the crust formation and temperature of melt-pool. The properties and flow condition of melt pool covering with the conditions of severe accident are used to evaluate the heat transfer coefficients in each reviewed model. Also, the scope and limitation of each model for application is assessed. A phenomenological analysis is performed with MELCOR 1.8.2 and MELCOR 1.8.3 And its results is compared with corresponding experimental reports of SWISS and SURC experiments. And the calculation is performed to assess the 1000 MWe PWR reactor cavity coolability. To improve the heat transfer model between melt-pool and overlying coolant and analyze the phase change of melt-pool, 2 dimensional governing equations are established using the enthalpy method and computational program is accomplished in this study. The benchmarking calculation is performed and its results are compared to the experiment which has not considered effects of the coolant boiling and the gas injection. Ultimately, the model shall be developed for considering the gas injection effect and coolant boiling effect. 66 refs., 10 tabs., 29 refs. (author)

  1. JERICHO computer code: PWR containment response during severe accidents description and sensitivity analysis

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.

    1983-12-01

    The JERICHO code has been developed in order to study the thermodynamic behaviour inside the reactor containment building for the complete spectrum of accident sequences likely to occur in such a reactor, including models for the various mass and energy transfer phenomena, for water spray, for hydrogen and carbon monoxide flammability limits and combustion, as well as for containment venting. Sensitivity analyses have been performed on a severe accident sequence, (namely, small LOCA with failure of the emergency core cooling and containment spray systems), involving core melting and subsequent concrete containment basemat erosion. The effect of various models, such as mass and energy transfer to the structures, has been studied. The influence of the concrete composition, of the fission product deposition and of the thermal degradation of the reactor cavity concrete walls on long term thermodynamic behaviour has also been investigated

  2. Assessment of SPACE code for multiple failure accident: 1% Cold Leg Break LOCA with HPSI failure at ATLAS Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Hyuk; Lee, Seung Wook; Kim, Kyung-Doo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Design extension conditions (DECs) is a popular key issue after the Fukushima accident. In a viewpoint of the reinforcement of the defense in depth concept, a high-risk multiple failure accident should be reconsidered. The target scenario of ATLAS A5.1 test was LSTF (Large Scale Test Facility) SB-CL-32 test, a 1% SBLOCA with total failure of high pressure safety injection (HPSI) system of emergency core cooling system (ECCS) and secondary side depressurization as the accident management (AM) action, as a counterpart test. As the needs to prepare the DEC accident because of a multiple failure of the present NPPs are emphasized, the capability of SPACE code, just like other system analysis code, is required to expand the DEC area. The objectives of this study is to validate the capability of SPACE code for a DEC scenario, which represents multiple failure accident like as a SBLOCA with HPSI fail. Therefore, the ATLAS A5.1 test scenario was chosen. As the needs to prepare the DEC accident because of a multiple failure of operating NPPs are emphasized, the capability of SPACE code is needed to expand the DEC area. So the capability of SPACE code was validated for one of a DEC scenario. The target scenario was selected as the ATLAS A5.1 test, which is a 1% SBLOCA with total failure of HPSI system of ECCS and secondary side depressurization. Through the sensitivity study on discharge coefficient of break flow, the best fit of integrated mass was found. Using the coefficient, the ATLAS A5.1 test was analyzed using the SPACE code. The major thermal hydraulic parameters such as the system pressure, temperatures were compared with the test and have a good agreement. Through the simulation, it was concluded that the SPACE code can effectively simulate one of multiple failure accidents like as SBLOCA with HPSI failure accident.

  3. 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.

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

    International Nuclear Information System (INIS)

    Roewekamp, M.

    1991-01-01

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

  5. Research activity about the radiological consequences of the Chernobyl NPS accident and social activity to assist its sufferers

    International Nuclear Information System (INIS)

    Imanaka, Tetsuji; Koide, Hiroaki; Kobayashi, Keiji

    1998-01-01

    Due to the Chernobyl Accident in April 1986, a series of serious radiological consequences were brought in Ukraine, Belarus and Russia. The former Soviet Union and the authorities in the world such as IAEA, however, have been denying serious health consequences among the people around Chernobyl since the beginning of the accident. On the other hand, a lot of works indicating serious health effects of the accident have been reported by scientists in these affected countries although they are not well known in the western countries. Since 1993, under the research grant of the Toyota foundation, we have continued a cooperative program to investigate research activities in these countries about the Chernobyl accident and to look into data and information that were not known so far. The information concerning the social system and activity to assist the sufferers from the accident has been also overviewed, including legal aspects of the Chernobyl problem. Here we are presenting an outline of our cooperation activity and our work concerning dose estimation for the inhabitants around the Chernobyl NPS at the first stage after the accident. The results of our estimation suggest that at least several hundreds of inhabitants received radiation dose exceeding 1 Sv before their evacuation. The whole reports of our cooperation program will be published in English and in Japanese in the next year. (author)

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

    International Nuclear Information System (INIS)

    Arffman, A.; Cazalis, B.

    2010-01-01

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

  7. SPACE code simulation of ATLAS DVI line break accident test (SB DVI 08 Test)

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Sang Gyu [KHNP, Daejeon (Korea, Republic of)

    2012-10-15

    APR1400 has adopted new safety design features which are 4 mechanically independent DVI (Direct Vessel Injection) systems and fluidic device in the safety injection tanks (SITs). Hence, DVI line break accident has to be evaluated as one of the small break LOCA (SBLOCA) to ensure the safety of APR1400. KAERI has been performed for DVI line break test (SB DVI 08) using ATLAS (Advanced Thermal Hydraulic Test Loop for Accident Simulation) facility which is an integral effect test facility for APR1400. The test result shows that the core collapsed water level decreased before a loop seal clearance, so that a core uncover occurred. At this time, the peak cladding temperature (PCT) is rapidly increased even though the emergency core cooling (ECC) water is injected from safety injection pump (SIP). This test result is useful for supporting safety analysis using thermal hydraulic safety analysis code and increases the understanding of SBLOCA phenomena in APR1400. The SBLOCA evaluation methodology for APR1400 is now being developed using SPACE code. The object of the development of this methodology is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. ATLAS SB DVI 08 test is selected for the evaluation of SBLOCA methodology using SPACE code. Before applying the conservative models and correlations, benchmark calculation of the test is performed with the best estimate models and correlations to verify SPACE code capability. This paper deals with benchmark calculations results of ATLAS SB DVI 08 test. Calculation results of the major hydraulics variables are compared with measured data. Finally, this paper carries out the SPACE code performances for simulating the integral effect test of SBLOCA.

  8. Research and managing institutions in Ukraine concerning the radiological consequences of the Chernobyl accident

    International Nuclear Information System (INIS)

    Nasvit, O.

    1998-01-01

    The paper presents temporal changes of the national organizations in managing the Chernobyl accident and its activities. The National Academy of Sciences of Ukraine started its activity from the first days after the accident. In 1990 a special executive body, the State committee of Chernobyl Affairs was established in Ukraine to manage the whole activity to overcome the Chernobyl problems. In 1991 it was rearranged into the Ministry of Chernobyl Affairs. In 1996 a new Ministry of Ukraine on Emergences and Affairs of Population Protection from the Consequences of Chernobyl Catastrophe(MEA) was founded on the basis of the Min. Chernobyl and Headquarters Staff of Civil Defence. The National Commission on Radiological Protection of Ukraine (NCRPU) belongs to the Parliament structure. NCRPU is responsible for approval of radiological safety standards and derived regulations. Very often the regulation approved are stricter than the international recommendations. There is an essential lack of attention within the Parliament to the activity of NCRPU. Ministry of Health is responsible for all kinds of medical care for the people suffering from the Chernobyl Catastrophe. In order to provide permanent medical service, a nation-wide scheme has been worked out. Scientific Center for Radiation Medicine is the leading scientific institute of the Academy of Medical Sciences. The State scientific Center of Environmental Radio geochemistry was created in 1996 on the basis of the two departments of the Institute of Geochemistry. The Center was created in order to improve coordination and managing of scientific researches on the behavior of artificial and natural radionuclides and chemical substances in the environment etc.. The Chernobyl Scientific-Technical center for International Research was created in March,1996. The Ukrainian Scientific Hygienic Center of Ministry of Health was created in 1989 and included two institutions. The subjects, the direction of research works

  9. The consequences of the Chernobyl accident - the radioecological database Redac of the French-German initiative

    Energy Technology Data Exchange (ETDEWEB)

    Deville-Cavelin, G.; Biesold, H.; Chabanyuk, V. [Radioprotection and Nuclear Safety Institute (IRSN), Dir. of Environment and Intervention (DEI) - CEA Cadarache, 13 - Saint-Paul-lez-Durance (France)

    2004-07-01

    The French-German Initiative for Chernobyl (FGI), implemented by IRSN and GRS from 1997 until the end of 2003, included the 'Project on the Radioecological Consequences of the Accident'. The most relevant fields of radioecology and post-accidental aspects have been studied, such as radionuclides transfers to plants, to animals, by surface runoff, in the aquatic environment and in the urban environment, wastes management and countermeasures. The main goal was to collect and harmonise, from Belarus, Russia and Ukraine, the highest possible amount of data and results on these different topics. These data have been verified, validated and organized in a common geo-referenced database REDAC (Radioecological Database After Chernobyl). For linking the different data, maps of initial and present contamination by {sup 137}Cs and {sup 90}Sr have been drawn up and relevant environmental non-radioactive data have been included. The operational database built will also allow the management of the wastes disposal sites. Countermeasures used after the accident for urban areas, natural and agricultural environment, have been described and classified. A methodology for evaluating their effectiveness has been developed. This database constitutes a tool for the development and validation of operational, assessment and explicative models. This allows the quantification and assessment of radionuclide transfer in the different compartments of ecosystems. So the main parameters influencing the transfers can be identified. REDAC should be completed by further investigations, for example on transuranic elements and extended to larger geographical zones. The database should also be combined with others provided by different organisations (IAEA, IRSN, UIR, ). (author)

  10. The consequences of the Chernobyl accident - the radioecological database Redac of the French-German initiative

    International Nuclear Information System (INIS)

    Deville-Cavelin, G.; Biesold, H.; Chabanyuk, V.

    2004-01-01

    The French-German Initiative for Chernobyl (FGI), implemented by IRSN and GRS from 1997 until the end of 2003, included the 'Project on the Radioecological Consequences of the Accident'. The most relevant fields of radioecology and post-accidental aspects have been studied, such as radionuclides transfers to plants, to animals, by surface runoff, in the aquatic environment and in the urban environment, wastes management and countermeasures. The main goal was to collect and harmonise, from Belarus, Russia and Ukraine, the highest possible amount of data and results on these different topics. These data have been verified, validated and organized in a common geo-referenced database REDAC (Radioecological Database After Chernobyl). For linking the different data, maps of initial and present contamination by 137 Cs and 90 Sr have been drawn up and relevant environmental non-radioactive data have been included. The operational database built will also allow the management of the wastes disposal sites. Countermeasures used after the accident for urban areas, natural and agricultural environment, have been described and classified. A methodology for evaluating their effectiveness has been developed. This database constitutes a tool for the development and validation of operational, assessment and explicative models. This allows the quantification and assessment of radionuclide transfer in the different compartments of ecosystems. So the main parameters influencing the transfers can be identified. REDAC should be completed by further investigations, for example on transuranic elements and extended to larger geographical zones. The database should also be combined with others provided by different organisations (IAEA, IRSN, UIR, ). (author)

  11. Dosimetric reconstruction of radiological accident by numerical simulations by means associating an anthropomorphic model and a Monte Carlo computation code

    International Nuclear Information System (INIS)

    Courageot, Estelle

    2010-01-01

    After a description of the context of radiological accidents (definition, history, context, exposure types, associated clinic symptoms of irradiation and contamination, medical treatment, return on experience) and a presentation of dose assessment in the case of external exposure (clinic, biological and physical dosimetry), this research thesis describes the principles of numerical reconstruction of a radiological accident, presents some computation codes (Monte Carlo code, MCNPX code) and the SESAME tool, and reports an application to an actual case (an accident which occurred in Equator in April 2009). The next part reports the developments performed to modify the posture of voxelized phantoms and the experimental and numerical validations. The last part reports a feasibility study for the reconstruction of radiological accidents occurring in external radiotherapy. This work is based on a Monte Carlo simulation of a linear accelerator, with the aim of identifying the most relevant parameters to be implemented in SESAME in the case of external radiotherapy

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

  13. RASCAL [Radiological Assessment System for Consequence AnaLysis]: A screening model for estimating doses from radiological accidents

    International Nuclear Information System (INIS)

    Sjoreen, A.L.; Athey, G.F.; Sakenas, C.A.; McKenna, T.J.

    1988-01-01

    The Radiological Assessment System for Consequence AnaLysis (RASCAL) is a new MS-DOS-based dose assessment model which has been written for the US Nuclear Regulatory Commission for use during response to radiological emergencies. RASCAL is designed to provide crude estimates of the effects of an accident while the accident is in progress and only limited information is available. It has been designed to be very simple to use and to run quickly. RASCAL is unique in that it estimates the source term based on fundamental plant conditions and does not rely solely on release rate estimation (e.g., Ci/sec of I-131). Therefore, it can estimate consequences of accidents involving unmonitored pathways or projected failures. RASCAL will replace the older model, IRDAM. 6 refs

  14. Have the consequences of reactor accidents for the population been well assessed? Six questions to the experts in the field

    Energy Technology Data Exchange (ETDEWEB)

    Pohl, Peter

    2016-07-15

    Six questions to the experts in the field are posed: (1) Why is the assessment of accident consequences not separated in long-term and peak exposure? (2) Why is the exposure due to I-131 seen critical mainly in regard to the thyroid? (3) Do you have any reliable relations of health risk versus peak exposure? (4) Why do you not abolish the LNT assumption and replace it with a threshold model? (5) Why do you include indirect, psycho-somatic effects in assessing the consequences of reactor accidents when this is not customary with accidents with often more casualties? (6) How can the number of Chernobyl-assigned thyroid cancers have risen from some 600 about to some 4,000 today, when the latency period is in the range of 4 to 5 years?.

  15. Advantages of Westinghouse BWR control rod drop accidents methodology utilizing integrated POLCA-T code

    International Nuclear Information System (INIS)

    Panayotov, Dobromir

    2008-01-01

    The paper focuses on the activities pursued by Westinghouse in the development and licensing of POLCA-T code Control Rod Drop Accident (CRDA) Methodology. The comprehensive CRDA methodology that utilizes PHOENIX4/POLCA7/POLCA-T calculation chain foresees complete cycle-specific analysis. The methodology consists of determination of candidates of control rods (CR) that could cause a significant reactivity excursion if dropped throughout the entire fuel cycle, selection of limiting initial conditions for CRDA transient simulation and transient simulation itself. The Westinghouse methodology utilizes state-of-the-art methods. Unnecessary conservatisms in the methodology have been avoided to allow the accurate prediction of margin to design bases. This is mainly achieved by using the POLCA-T code for dynamic CRDA evaluations. The code belongs to the same calculation chain that is used for core design. Thus the very same reactor, core, cycle and fuel data base is used. This allows also reducing the uncertainties of input data and parameters that determine the energy deposition in the fuel. Uncertainty treatment, very selective use of conservatisms, selection of the initial conditions for limiting case analyses, incorporation into POLCA-T code models of the licensed fuel performance code are also among the means of performing realistic CRDA transient analyses. (author)

  16. Study on the possible consequences of a severe accident in a Swiss nuclear power plant on the drinking water supply

    International Nuclear Information System (INIS)

    Ustohalova, Veronika; Kueppers, Christian; Claus, Manuel

    2014-01-01

    The study on the possible consequences of a severe accident in a Swiss nuclear power plant on the drinking water supply covers the following issues: estimation of possible source terms and radioactive materials release rates, airborne water contamination, water contamination by direct pollution, consequences for the drinking water supply, emergency measures in case of a drinking water contamination, routine surveillance of surface and ground water and improvement possibilities in nuclear power plants.

  17. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject F. Contributions to code validation using BWR data and to evaluation and optimization of accident management measures. Final report; WASA-BOSS. Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt F. Beitraege zur Codevalidierung anhand von SWR-Daten und zur Bewertung und Optimierung von Stoerfallmassnahmen. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino; Imke, Uwe; Sanchez Espinoza, Victor

    2016-09-15

    The exact knowledge of the transient course of events and of the dominating processes during a severe accident in a nuclear power station is a mandatory requirement to elaborate strategies and measures to minimize the radiological consequences of core melt. Two typical experiments using boiling water reactor assemblies were modelled and simulated with the severe accident simulation code ATHLET-CD. The experiments are related to the early phase of core degradation in a boiling water reactor. The results reproduce the thermal behavior and the hydrogen production due to oxidation inside the bundle until relocation of material by melting. During flooding of the overheated assembly temperatures and hydrogen oxidation are under estimated. The deviations from the experimental results can be explained by the missing model to simulate bore carbide oxidation of the control rods. On basis of a hypothetical loss of coolant accident in a typical German boiling water reactor the effectivity of flooding the partial degraded core is investigated. This measure of mitigation is efficient and prevents failure of the reactor pressure vessel if it starts before molten material is relocated into the lower plenum. Considerable amount of hydrogen is produced by oxidation of the metallic components.

  18. 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)

  19. A thermo-mechanical benchmark calculation of an hexagonal can in the BTI accident with ABAQUS code

    International Nuclear Information System (INIS)

    Zucchini, A.

    1988-07-01

    The thermo-mechanical behaviour of an hexagonal can in a benchmark problem (simulating the conditions of a BTI accident in a fuel assembly) is examined by means of the ABAQUS code: the effects of the geometric nonlinearity are shown and the results are compared with those of a previous analysis performed with the INCA code. (author)

  20. Application of natural adsorbents as decontamination agents for the elimination of the consequences of the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Tarasevich, Yu.I.

    1996-01-01

    The scientific foundations of using natural adsorbents as ion exchangers,filtering media and adagulants for water purification ase presented. The results showing the efficiency of practical application of natural adsorbents for the decontamination of water, clothes, machinery, construction materials, etc. during the elimination of the consequences of the Chernobyl reactor accident in 1986-1987 are presented

  1. Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

    International Nuclear Information System (INIS)

    Lee, A.G.; Wilkin, G.B.

    1996-03-01

    During the 'Workshop on R and D needs' at the 3rd Meeting of the International Group on Research Reactors (IGORR-III), the participants agreed that it would be useful to compile a survey of the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods various organizations use to verify and validate their codes and libraries. Five organizations, Atomic Energy of Canada Limited (AECL, Canada), China Institute of Atomic Energy (CIAE, People's Republic of China), Japan Atomic Energy Research Institute (JAERI, Japan), Oak Ridge National Laboratories (ORNL, USA), and Siemens (Germany) responded to the survey. The results of the survey are compiled in this report. (author) 36 refs., 3 tabs

  2. ASTEC code development, validation and applications for severe accident management within the CESAM European project - 15392

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Chatelard, P.; Chevalier-Jabet, K.; Nowack, H.; Herranz, L.E.; Pascal, G.; Sanchez-Espinoza, V.H.

    2015-01-01

    ASTEC, jointly developed by IRSN and GRS, is considered as the European reference code since it capitalizes knowledge from the European research on the domain. The CESAM project aims at its enhancement and extension for use in severe accident management (SAM) analysis of the nuclear power plants (NPP) of Generation II-III presently under operation or foreseen in near future in Europe, spent fuel pools included. Within the CESAM project 3 main types of research activities are performed: -) further validation of ASTEC models important for SAM, in particular for the phenomena being of importance in the Fukushima-Daichi accidents, such as reflooding of degraded cores, pool scrubbing, hydrogen combustion, or spent fuel pools behaviour; -) modelling improvements, especially for BWR or based on the feedback of validation tasks; and -) ASTEC applications to severe accident scenarios in European NPPs in order to assess prevention and mitigation measures. An important step will be reached with the next major ASTEC V2.1 version planned to be delivered in the first part of 2015. Its main improvements will concern the possibility to simulate in details the core degradation of BWR and PHWR and a model of reflooding of severely degraded cores. A new user-friendly Graphic User Interface will be available for plant analyses

  3. Computer codes for the study of the loss of coolant accident of PWR reactors

    International Nuclear Information System (INIS)

    Gomolinski, M.; Menessier, D.; Tellier, N.

    1975-01-01

    The CEA has undertaken a large programme to study the consequence on the core of the LOCA of a PWR. In the programme, simultaneously carried out experiments and the development of the calculations means are described. Several experiments such as OMEGA, ERSEC and PHEBUS tests, which provide data to check the computer codes are outlined briefly in the paper. For analysis of the LOCA of a PWR, a series of computer codes, which are at present in use or under development, are linked with each other. The codes are DANAIDES for blowdown, CERES for refill and reflood, THETA-1B and FLIRA for heat up calculation during the blow-down and the reflooding period respectively. FLIRA-PASTEL, a combination of FLIRA and PASTEL which calculate the stress and deformations of material using the finite element method, will be used in place of FLIRA. The basic models and flowcharts of the above codes are described in the paper

  4. Consequences of the Chernobyl accident for the natural and human environments

    Energy Technology Data Exchange (ETDEWEB)

    Dreicer, M. [Lawrence Livermore National Lab., CA (United States); Aarkog, A. [Risoe National Lab., Roskilde (Denmark); Alexakhin, R. [Russian Inst. of Agricultural Radiology and Agroecology (Russian Federation); Anspaugh, L. [Lawrence Livermore National Lab., CA (United States); Arkhipov, N.P. [Scientific and Technical Centre of the RIA `Pripyat` (Ukraine); Johansson, K.-J. [Swedish Univ. of Agricultural Sciences, Uppsala (Sweden)

    1996-07-01

    In the ten years since the Chernobyl accident, an enormous amount of work has been done to assess the consequences to the natural and human environment. Although it is difficult to summarize such a large and varied field, some general conclusions can be drawn. This background paper includes the main findings concerning the direct impacts of radiation on the flora and fauna; the general advances of knowledge in the cycling of radionuclides in natural, seminatural and agricultural environments; some evaluation of countermeasures that were used; and a summary of the human radiation doses resulting from the environmental contamination. although open questions still remain, it can be concluded that: (1) at high radiation levels, the natural environment has shown short term impacts but any significant long term impacts remain to be seen; (2) effective countermeasures can be taken to reduce the transfer of contamination from the environment to humans but these are highly site specific and must be evaluated in terms of practicality as well as population does reduction; (3) the majority of the doses have already been received by the human population. If agricultural countermeasures are appropriately taken, the main source of future doses will be the gathering of food and recreational activities in natural and seminatural ecosystems.

  5. General situation of the radiological consequences of the Chernobyl accident in Ukraine

    International Nuclear Information System (INIS)

    Grodzinsky, D.M.

    1998-01-01

    Following the Chernobyl Nuclear Accident on April 26, 1986, epidemiological analyses of data point to impressive deterioration of the health of the people affected by radionuclide contamination in the environment. This deterioration of population health embraces a broad spectrum of diseases. Epidemiological prediction of the rate of thyroid cancer in children near Chernobyl seems strikingly compatible with a real increase. But there is a tendency to consider the morbidity augmentation as a result having been associated with the factors of non-radioactive origin (chemical compounds, heavy metals and mainly social-psychological syndrome development). The Chernobyl catastrophe has implied a heavy burden for Ukraine: pollution of air, water, soils and vegetation in all ecosystems, late radiological effects in the health of people, losses of arable land and forest, necessity of mass-evacuation from thousands of settlements in the contaminated regions, severe psychological shock for millions of people, and painful suffering of unexpected life tragedies. Eleven years after, this tragic event with its causes and consequence brings one to very important conclusions concerning moral aspects of human relations within the nuclear society, as well as interactions between the society and the environment. (J.P.N.)

  6. Consequences of the Chernobyl accident for the natural and human environments

    International Nuclear Information System (INIS)

    Dreicer, M.; Aarkog, A.; Alexakhin, R.; Anspaugh, L.; Arkhipov, N.P.; Johansson, K.-J.

    1996-07-01

    In the ten years since the Chernobyl accident, an enormous amount of work has been done to assess the consequences to the natural and human environment. Although it is difficult to summarize such a large and varied field, some general conclusions can be drawn. This background paper includes the main findings concerning the direct impacts of radiation on the flora and fauna; the general advances of knowledge in the cycling of radionuclides in natural, seminatural and agricultural environments; some evaluation of countermeasures that were used; and a summary of the human radiation doses resulting from the environmental contamination. although open questions still remain, it can be concluded that: (1) at high radiation levels, the natural environment has shown short term impacts but any significant long term impacts remain to be seen; (2) effective countermeasures can be taken to reduce the transfer of contamination from the environment to humans but these are highly site specific and must be evaluated in terms of practicality as well as population does reduction; (3) the majority of the doses have already been received by the human population. If agricultural countermeasures are appropriately taken, the main source of future doses will be the gathering of food and recreational activities in natural and seminatural ecosystems

  7. Code Development on Fission Product Behavior under Severe Accident-Validation of Aerosol Sedimentation

    International Nuclear Information System (INIS)

    Ha, Kwang Soon; Kim, Sung Il; Jang, Jin Sung; Kim, Dong Ha

    2016-01-01

    The gas and aerosol phases of the radioactive materials move through the reactor coolant systems and containments as loaded on the carrier gas or liquid, such as steam or water. Most radioactive materials might escape in the form of aerosols from a nuclear power plant during a severe reactor accident, and it is very important to predict the behavior of these radioactive aerosols in the reactor cooling system and in the containment building under severe accident conditions. Aerosols are designated as very small solid particles or liquid droplets suspended in a gas phase. The suspended solid or liquid particles typically have a range of sizes of 0.01 m to 20 m. Aerosol concentrations in reactor accident analyses are typically less than 100 g/m3 and usually less than 1 g/m3. When there are continuing sources of aerosol to the gas phase or when there are complicated processes involving engineered safety features, much more complicated size distributions develop. It is not uncommon for aerosols in reactor containments to have bimodal size distributions for at least some significant periods of time early during an accident. Salient features of aerosol physics under reactor accident conditions that will affect the nature of the aerosols are (1) the formation of aerosol particles, (2) growth of aerosol particles, (3) shape of aerosol particles. At KAERI, a fission product module has been developed to predict the behaviors of the radioactive materials in the reactor coolant system under severe accident conditions. The fission product module consists of an estimation of the initial inventories, species release from the core, aerosol generation, gas transport, and aerosol transport. The final outcomes of the fission product module designate the radioactive gas and aerosol distribution in the reactor coolant system. The aerosol sedimentation models in the fission product module were validated using ABCOVE and LACE experiments. There were some discrepancies on the predicted

  8. Code Development on Fission Product Behavior under Severe Accident-Validation of Aerosol Sedimentation

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang Soon; Kim, Sung Il; Jang, Jin Sung; Kim, Dong Ha [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The gas and aerosol phases of the radioactive materials move through the reactor coolant systems and containments as loaded on the carrier gas or liquid, such as steam or water. Most radioactive materials might escape in the form of aerosols from a nuclear power plant during a severe reactor accident, and it is very important to predict the behavior of these radioactive aerosols in the reactor cooling system and in the containment building under severe accident conditions. Aerosols are designated as very small solid particles or liquid droplets suspended in a gas phase. The suspended solid or liquid particles typically have a range of sizes of 0.01 m to 20 m. Aerosol concentrations in reactor accident analyses are typically less than 100 g/m3 and usually less than 1 g/m3. When there are continuing sources of aerosol to the gas phase or when there are complicated processes involving engineered safety features, much more complicated size distributions develop. It is not uncommon for aerosols in reactor containments to have bimodal size distributions for at least some significant periods of time early during an accident. Salient features of aerosol physics under reactor accident conditions that will affect the nature of the aerosols are (1) the formation of aerosol particles, (2) growth of aerosol particles, (3) shape of aerosol particles. At KAERI, a fission product module has been developed to predict the behaviors of the radioactive materials in the reactor coolant system under severe accident conditions. The fission product module consists of an estimation of the initial inventories, species release from the core, aerosol generation, gas transport, and aerosol transport. The final outcomes of the fission product module designate the radioactive gas and aerosol distribution in the reactor coolant system. The aerosol sedimentation models in the fission product module were validated using ABCOVE and LACE experiments. There were some discrepancies on the predicted

  9. Comparison of Severe Accident Results Among SCDAP/RELAP5, MAAP, and MELCOR Codes

    International Nuclear Information System (INIS)

    Wang, T.-C.; Wang, S.-J.; Teng, J.-T.

    2005-01-01

    This paper demonstrates a large-break loss-of-coolant accident (LOCA) sequence of the Kuosheng nuclear power plant (NPP) and station blackout sequence of the Maanshan NPP with the SCDAP/RELAP5 (SR5), Modular Accident Analysis Program (MAAP), and MELCOR codes. The large-break sequence initiated with double-ended rupture of a recirculation loop. The main steam isolation valves (MSIVs) closed, the feedwater pump tripped, the reactor scrammed, and the assumed high-pressure and low-pressure spray systems of the emergency core cooling system (ECCS) were not functional. Therefore, all coolant systems to quench the core were lost. MAAP predicts a longer vessel failure time, and MELCOR predicts a shorter vessel failure time for the large-break LOCA sequence. The station blackout sequence initiated with a loss of all alternating-current (ac) power. The MSIVs closed, the feedwater pump tripped, and the reactor scrammed. The motor-driven auxiliary feedwater system and the high-pressure and low-pressure injection systems of the ECCS were lost because of the loss of all ac power. It was also assumed that the turbine-driven auxiliary feedwater pump was not functional. Therefore, the coolant system to quench the core was also lost. MAAP predicts a longer time of steam generator dryout, time interval between top of active fuel and bottom of active fuel, and vessel failure time than those of the SR5 and MELCOR predictions for the station blackout sequence. The three codes give similar results for important phenomena during the accidents, including SG dryout, core uncovery, cladding oxidation, cladding failure, molten pool formulation, debris relocation to the lower plenum, and vessel head failure. This paper successfully demonstrates the large-break LOCA sequence of the Kuosheng NPP and the station blackout sequen