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Sample records for melcor accident consequence

  1. 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

  2. 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

  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)

    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.

  6. 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

  7. 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

  8. MELCOR analysis of the TMI-2 accident

    International Nuclear Information System (INIS)

    Boucheron, E.A.

    1990-01-01

    This paper describes the analysis of the Three Mile Island-2 (TMI-2) standard problem that was performed with MELCOR. The MELCOR computer code is being developed by Sandia National Laboratories for the Nuclear Regulatory Commission for the purpose of analyzing severe accident in nuclear power plants. The primary role of MELCOR is to provide realistic predictions of severe accident phenomena and the radiological source team. The analysis of the TMI-2 standard problem allowed for comparison of the model predictions in MELCOR to plant data and to the results of more mechanistic analyses. This exercise was, therefore valuable for verifying and assessing the models in the code. The major trends in the TMI-2 accident are reasonably well predicted with MELCOR, even with its simplified modeling. Comparison of the calculated and measured results is presented and, based on this comparison, conclusions can be drawn concerning the applicability of MELCOR to severe accident analysis. 5 refs., 10 figs., 3 tabs

  9. 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

  10. 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

  11. 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

  12. MELCOR modeling of Fukushima unit 2 accident

    Energy Technology Data Exchange (ETDEWEB)

    Sevon, Tuomo [VTT Technical Research Centre of Finland, Espoo (Finland)

    2014-12-15

    A MELCOR model of the Fukushima Daiichi unit 2 accident was created in order to get a better understanding of the event and to improve severe accident modeling methods. The measured pressure and water level could be reproduced relatively well with the calculation. This required adjusting the RCIC system flow rates and containment leak area so that a good match to the measurements is achieved. Modeling of gradual flooding of the torus room with water that originated from the tsunami was necessary for a satisfactory reproduction of the measured containment pressure. The reactor lower head did not fail in this calculation, and all the fuel remained in the RPV. 13 % of the fuel was relocated from the core area, and all the fuel rods lost their integrity, releasing at least some volatile radionuclides. According to the calculation, about 90 % of noble gas inventory and about 0.08 % of cesium inventory was released to the environment. The release started 78 h after the earthquake, and a second release peak came at 90 h. Uncertainties in the calculation are very large because there is scarce public data available about the Fukushima power plant and because it is not yet possible to inspect the status of the reactor and the containment. Uncertainty in the calculated cesium release is larger than factor of ten.

  13. MELCOR DB Construction for the Severe Accident Analysis DB

    International Nuclear Information System (INIS)

    Song, Y. M.; Ahn, K. I.

    2011-01-01

    The Korea Atomic Energy Research Institute (KAERI) has been constructing a severe accident analysis database (DB) under a National Nuclear R and D Program. In particular, an MAAP (commercial code being widely used in industries for integrated severe accident analysis) DB for many scenarios including a station blackout (SBO) has been completed. This paper shows the MELCOR DB construction process with examples of SBO scenarios, and the results will be used for a comparison with the MAAP DB

  14. Calculation of spent fuel pool severe accident with MELCOR

    International Nuclear Information System (INIS)

    Deng Jian; Xiang Qing'an; Zhou Kefeng

    2014-01-01

    A calculation model was established for spent fuel pool (SFP) using MELCOR code to study the severe accident phenomena caused by the long term station black-out (SBO), including spent fuel heatup, zirconium cladding oxidation, and the injection into SFP to mitigate the severe accident. The results show that the severe accident progression is slow and relates directly with the initial water level in SFP. It is illustrated that the injection into SFP is one of the best mitigated measures for the SFP severe accident. (authors)

  15. Development of Methodology for Spent Fuel Pool Severe Accident Analysis Using MELCOR Program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won-Tae; Shin, Jae-Uk [RETech. Co. LTD., Yongin (Korea, Republic of); Ahn, Kwang-Il [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The general reason why SFP severe accident analysis has to be considered is that there is a potential great risk due to the huge number of fuel assemblies and no containment in a SFP building. In most cases, the SFP building is vulnerable to external damage or attack. In contrary, low decay heat of fuel assemblies may make the accident processes slow compared to the accident in reactor core because of a great deal of water. In short, its severity of consequence cannot exclude the consideration of SFP risk management. The U.S. Nuclear Regulatory Commission has performed the consequence studies of postulated spent fuel pool accident. The Fukushima-Daiichi accident has accelerated the needs for the consequence studies of postulated spent fuel pool accidents, causing the nuclear industry and regulatory bodies to reexamine several assumptions concerning beyond-design basis events such as a station blackout. The tsunami brought about the loss of coolant accident, leading to the explosion of hydrogen in the SFP building. Analyses of SFP accident processes in the case of a loss of coolant with no heat removal have studied. Few studies however have focused on a long term process of SFP severe accident under no mitigation action such as a water makeup to SFP. USNRC and OECD have co-worked to examine the behavior of PWR fuel assemblies under severe accident conditions in a spent fuel rack. In support of the investigation, several new features of MELCOR model have been added to simulate both BWR fuel assembly and PWR 17 x 17 assembly in a spent fuel pool rack undergoing severe accident conditions. The purpose of the study in this paper is to develop a methodology of the long-term analysis for the plant level SFP severe accident by using the new-featured MELCOR program in the OPR-1000 Nuclear Power Plant. The study is to investigate the ability of MELCOR in predicting an entire process of SFP severe accident phenomena including the molten corium and concrete reaction. The

  16. MELCOR Severe Accident Analysis on the SMART Reactor

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Madni, I.K.; Eltawila, F.

    1994-01-01

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

  18. Analysis of the primary source term for meltdown accidents using MELCOR 1.8.2

    International Nuclear Information System (INIS)

    Schmuck, P.

    1995-01-01

    The MELCOR code describing accident phenomena in the core and primary systems was used for source term calculations and - in the context of the MELCOR Cooperative Assessment Programme - for studying two-phase flows through components such as valves and chokes. Results of the latter studies in comparison to experiments gave hints for an improved calculation of momentum transfer between the phases. (orig.)

  19. Analysis of the primary source term for meltdown accidents using MELCOR 1.8.2; Analyse des primaeren Quellterms bei Kernschmelzunfaellen mit MELCOR 1.8.2

    Energy Technology Data Exchange (ETDEWEB)

    Schmuck, P.

    1995-08-01

    The MELCOR code describing accident phenomena in the core and primary systems was used for source term calculations and - in the context of the MELCOR Cooperative Assessment Programme - for studying two-phase flows through components such as valves and chokes. Results of the latter studies in comparison to experiments gave hints for an improved calculation of momentum transfer between the phases. (orig.)

  20. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject E. Improvement of the lower head model in MELCOR and calculations in connection with the FUKUSHIMA accident. Final report; WASA-BOSS. Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt E. Verbesserung des Lower Head-Modelles fuer MELCOR und MELCOR-Rechnungen zu Fukushima. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Kretzschmar, Frank; Dietrich, Philipp; Gabriel, Stephan; Miassoedov, Alexei

    2016-12-15

    The knowledge of the key phenomena, which govern the chronological sequence of a core melt accident, is a crucial precondition for the development of SAMGs (Severe Management Guides) to avoid and mitigate the radiological consequences for the population and the environment. In the frame of a dissertation a new model has been coupled with MELCOR, which describes the thermal interaction of a core melt with the RPV (reactor pressure vessel) wall in the lower plenum. This model allows a better description of this phenomenon. The method to couple extern programs with MELCOR had been already developed and used in a former dissertation at KIT-IKET. The model has been validated recalculating according experiments in the LIVE facility. Afterwards a defined accident scenario has been calculated for a German generic KONVOI power plant. 12 months after the start of the project a MELCOR input has been developed using data delivered by the Ruhr university of Bochum (subproject ''Simulation des Unfalls in Fukushima-Daichi zur Bewertung des Stoerfall-Analysecodes ATHLET-CD''). The results of this simulation have made a contribution to review the current understanding of the FUKUSHIMA sequence. HZDR and KIT-IKET have agreed in the course of the project, that KIT-IKET will develop a MELCOR input of a german generic KONVOI power plant following an ATHLET-CDinput of HZDR. Using this MELCOR input, a comparative analysis has been performed.

  1. Insights Gained from Forensic Analysis with MELCOR of the Fukushima-Daiichi Accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, Nathan C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-10-01

    Since the accidents at Fukushima-Daiichi, Sandia National Laboratories has been modeling these accident scenarios using the severe accident analysis code, MELCOR. MELCOR is a widely used computer code developed at Sandia National Laboratories since ~1982 for the U.S. Nuclear Regulatory Commission. Insights from the modeling of these accidents is being used to better inform future code development and potentially improved accident management. To date, our necessity to better capture in-vessel thermal-hydraulic and ex-vessel melt coolability and concrete interactions has led to the implementation of new models. The most recent analyses, presented in this paper, have been in support of the of the Organization for Economic Cooperation and Development Nuclear Energy Agency’s (OECD/NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Project. The goal of this project is to accurately capture the source term from all three releases and then model the atmospheric dispersion. In order to do this, a forensic approach is being used in which available plant data and release timings is being used to inform the modeled MELCOR accident scenario. For example, containment failures, core slumping events and lower head failure timings are all enforced parameters in these analyses. This approach is fundamentally different from a blind code assessment analysis often used in standard problem exercises. The timings of these events are informed by representative spikes or decreases in plant data. The combination of improvements to the MELCOR source code resulting from analysis previous accident analysis and this forensic approach has allowed Sandia to generate representative and plausible source terms for all three accidents at Fukushima Daiichi out to three weeks after the accident to capture both early and late releases. In particular, using the source terms developed by MELCOR, the MACCS software code, which models atmospheric dispersion and

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  3. MELCOR assessment of sequential severe accident mitigation actions under SGTR accident

    International Nuclear Information System (INIS)

    Choi, Wonjun; Jeon, Joongoo; Kim, Nam Kyung; Kim, Sung Joong

    2017-01-01

    The representative example of the severe accident studies using the severe accident code is investigation of effectiveness of developed severe accident management (SAM) strategy considering the positive and adverse effects. In Korea, some numerical studies were performed to investigate the SAM strategy using various severe accident codes. Seo et.al performed validation of RCS depressurization strategy and investigated the effect of severe accident management guidance (SAMG) entry condition under small break loss of coolant accident (SBLOCA) without safety injection (SI), station blackout (SBO), and total loss of feed water (TLOFW) scenarios. The SGTR accident with the sequential mitigation actions according to the flow chart of SAMG was simulated by the MELCOR 1.8.6 code. Three scenariospreventing the RPV failure were investigated in terms of fission product release, hydrogen risk, and the containment pressure. Major conclusions can be summarized as follows: (1) According to the flow chart of SAMG, RPV failure can be prevented depending on the method of RCS depressurization. (2) To reduce the release of fission product during the injecting into SGs, a temporary opening of SDS before the injecting into SGs was suggested. These modified sequences of mitigation actions can reduce the release of fission product and the adverse effect of SDS.

  4. Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J., E-mail: brad.merrill@inl.gov; Bragg-Sitton, Shannon M., E-mail: shannon.bragg-sitton@inl.gov; Humrickhouse, Paul W., E-mail: paul.humrickhouse@inl.gov

    2017-04-15

    Highlights: • Accident tolerant fuels (ATF) systems are currently under development for LWRs. • Many performance analysis tools are specifically developed for UO{sub 2}–Zr alloy fuel. • Modifications were made to the MELCOR code for candidate ATF cladding. • Preliminary analysis results for SiC and FeCrAl cladding concepts are presented. - Abstract: A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO{sub 2}–Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This paper describes modifications

  5. Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out

    International Nuclear Information System (INIS)

    Wang, Jun; Mccabe, Mckinleigh; Wu, Lei; Dong, Xiaomeng; Wang, Xianmao; Haskin, Troy Christopher; Corradini, Michael L.

    2017-01-01

    Highlights: • Thermo-physical and oxidation kinetics properties calculation and analysis of FeCrAl. • Properties modelling of FeCrAl in MELCOR. • Benchmark calculation of Surry nuclear power plant. - Abstract: Accident tolerant fuel and cladding materials are being investigated to provide a greater resistance to fuel degradation, oxidation and melting if long-term cooling is lost in a Light Water Reactor (LWR) following an accident such as a Station Blackout (SBO) or Loss of Coolant Accident (LOCA). Researchers at UW-Madison are analyzing an SBO sequence and examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. Our research work considers accident tolerant cladding materials (e.g., FeCrAl alloy) and their effect on the accident behavior. We first gathered the physical properties of this alternative cladding material via literature review and compared it to the usual zirconium alloys used in LWRs. We then developed a model for the Surry reactor for a Short-term SBO sequence and examined the effect of replacing FeCrAl for Zircaloy cladding. The analysis uses MELCOR, Version 1.8.6 YR, which is developed by Idaho National Laboratory in collaboration with MELCOR developers at Sandia National Laboratories. This version allows the user to alter the cladding material considered, and our study examines the behavior of the FeCrAl alloy as a substitute for Zircaloy. Our benchmark comparisons with the Sandia National Laboratory’s analysis of Surry using MELCOR 1.8.6 and the more recent MELCOR 2.1 indicate good overall agreement through the early phases of the accident progression. When FeCrAl is substituted for Zircaloy to examine its performance, we confirmed that FeCrAl slows the accident progression and reduce the amount of hydrogen generated. Our analyses also show that this special version of MELCOR can be used to evaluate other potential ATF cladding materials, e.g., SiC as well as innovative coatings on zirconium cladding

  6. Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jun, E-mail: jwang564@wisc.edu [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Mccabe, Mckinleigh [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Wu, Lei [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Dong, Xiaomeng [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Wang, Xianmao [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Haskin, Troy Christopher [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Corradini, Michael L., E-mail: corradini@engr.wisc.edu [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States)

    2017-03-15

    Highlights: • Thermo-physical and oxidation kinetics properties calculation and analysis of FeCrAl. • Properties modelling of FeCrAl in MELCOR. • Benchmark calculation of Surry nuclear power plant. - Abstract: Accident tolerant fuel and cladding materials are being investigated to provide a greater resistance to fuel degradation, oxidation and melting if long-term cooling is lost in a Light Water Reactor (LWR) following an accident such as a Station Blackout (SBO) or Loss of Coolant Accident (LOCA). Researchers at UW-Madison are analyzing an SBO sequence and examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. Our research work considers accident tolerant cladding materials (e.g., FeCrAl alloy) and their effect on the accident behavior. We first gathered the physical properties of this alternative cladding material via literature review and compared it to the usual zirconium alloys used in LWRs. We then developed a model for the Surry reactor for a Short-term SBO sequence and examined the effect of replacing FeCrAl for Zircaloy cladding. The analysis uses MELCOR, Version 1.8.6 YR, which is developed by Idaho National Laboratory in collaboration with MELCOR developers at Sandia National Laboratories. This version allows the user to alter the cladding material considered, and our study examines the behavior of the FeCrAl alloy as a substitute for Zircaloy. Our benchmark comparisons with the Sandia National Laboratory’s analysis of Surry using MELCOR 1.8.6 and the more recent MELCOR 2.1 indicate good overall agreement through the early phases of the accident progression. When FeCrAl is substituted for Zircaloy to examine its performance, we confirmed that FeCrAl slows the accident progression and reduce the amount of hydrogen generated. Our analyses also show that this special version of MELCOR can be used to evaluate other potential ATF cladding materials, e.g., SiC as well as innovative coatings on zirconium cladding

  7. MELCOR Peer Review

    International Nuclear Information System (INIS)

    Boyack, B.E.; Dhir, V.K.; Gieseke, J.A.; Haste, T.J.; Kenton, M.A.; Khatib-Rahbar, M.; Leonard, M.T.; Viskanta, R.

    1992-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. The newest version of MELCOR is Version 1.8.1, July 1991. MELCOR development has reached the point that the United States Nuclear Regulatory Commission sponsored a broad technical review by recognized experts to determine or confirm the technical adequacy of the code for the serious and complex analyses it is expected to perform. For this purpose, an eight-member MELCOR Peer Review Committee was organized. The Committee has completed its review of the MELCOR code: the review process and findings of the MELCOR Peer Review Committee are documented in this report. The Committee has determined that recommendations in five areas are appropriate: (1) MELCOR numerics, (2) models missing from MELCOR Version 1.8.1, (3) existing MELCOR models needing revision, (4) the need for expanded MELCOR assessment, and (5) documentation

  8. Modular Accident Analysis Program (MAAP) - MELCOR Crosswalk: Phase II Analyzing a Partially Recovered Accident Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, Nathan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Faucett, Christopher [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Haskin, Troy Christopher [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Luxat, Dave [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Geiger, Garrett [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Codella, Brittany [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-10-01

    Following the conclusion of the first phase of the crosswalk analysis, one of the key unanswered questions was whether or not the deviations found would persist during a partially recovered accident scenario, similar to the one that occurred in TMI - 2. In particular this analysis aims to compare the impact of core degradation morphology on quenching models inherent within the two codes and the coolability of debris during partially recovered accidents. A primary motivation for this study is the development of insights into how uncertainties in core damage progression models impact the ability to assess the potential for recovery of a degraded core. These quench and core recovery models are of the most interest when there is a significant amount of core damage, but intact and degraded fuel still remain in the cor e region or the lower plenum. Accordingly this analysis presents a spectrum of partially recovered accident scenarios by varying both water injection timing and rate to highlight the impact of core degradation phenomena on recovered accident scenarios. This analysis uses the newly released MELCOR 2.2 rev. 966 5 and MAAP5, Version 5.04. These code versions, which incorporate a significant number of modifications that have been driven by analyses and forensic evidence obtained from the Fukushima - Daiichi reactor site.

  9. MELCOR based severe accident simulation for WWER-440 type nuclear power plants

    International Nuclear Information System (INIS)

    Vegh, E.; Buerger, L.; Gacs, A.; Gyenes, F.G.; Hozer, Z.; Makovi, P.

    1997-01-01

    SUBA is a MELCOR based severe accident simulator, installed this summer at the Hungarian Nuclear Safety Directorate. In this simulator the thermohydraulics, chemical reactions and material transport in the primary and secondary systems are calculated by the MELCOR code, but the containment, except the cavity, is modelled by the HERMET code, developed in our Institute. The instrumentation and control, the safety systems and the plant logic, are calculated by our models. This paper describes the main features of the used models and presents three different test transients. The presented transients are as follows: a small break LOCA, a cold leg large break LOCA, and the station blackout, without Diesel generators. In each treated transients the most important parameters are presented as time functions and the most significant events are analysed. (author)

  10. 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.)

  11. 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

  12. 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.

  13. Iodine chemistry effect on source term assessments. A MELCOR 186 YT study of a PWR severe accident sequence

    International Nuclear Information System (INIS)

    Herranz, Luis E.; Garcia, Monica; Otero, Bernadette

    2009-01-01

    Level-2 Probabilistic Safety Analysis has demonstrated to be a powerful tool to give insights into multiple aspects concerning severe accidents: phenomena with the greatest potential to lead to containment failure, safety systems performance and, even, to identify any additional accident management that could mitigate the consequences of such an even, etc. A major result of level-2 PSA is iodine content in Source Term since it is the main responsible for the radiological impact during the first few days after a hypothetical severe accident. Iodine chemistry is known to considerably affect iodine behavior and although understanding has improved substantially since the early 90's, a thorough understanding is still missing and most PSA studies do not address it when assessing severe accident scenarios. This paper emphasizes the quantitative and qualitative significance of considering iodine chemistry in level-2 PSA estimates. To do so a cold leg break, low pressure severe accident sequence of an actual pressurized water reactor has been analyzed with the MELCOR 1.8.6 YT code. Two sets of calculations, with and without chemistry, have been carried out and compared. The study shows that iodine chemistry could result in an iodine release to environment about twice higher, most of which would consist of around 60% of iodine in gaseous form. From these results it is concluded that exploratory studies on the potential effect of iodine chemistry on source term estimates should be carried out. (author)

  14. 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 sequence of the Maanshan NPP

  15. Accident analyses on TMLB' and LOCA for KNGR using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Choi, Y.; Ahn, K.I

    2000-11-01

    Plant specific phenomenological analyses for the Korean Next Generation Reactor, using MELCOR program, are described in this report. The most important two accident sequences, a station blackout and a loss of coolant scenario, are selected. Complete coverage of corium behavior both in-vessel and ex-vessel, and the corresponding containment responses, are analyzed. The in-vessel progression includes the thermal hydraulics in the primary system, core heat up, hydrogen generation, and melt progression up to the reactor vessel breach. The ex-vessel progression describes molten corium - concrete interaction phenomena and the pressure behavior in the containment atmosphere.

  16. Application of the integral code MELCOR for German NPPs and use within accident management and PSA projects

    International Nuclear Information System (INIS)

    Sonnenkalb, Martin

    2006-01-01

    The paper summarizes the application of MELCOR to German NPPS with PWR and BWR. A development of different code systems like ATHLET/ATHLET-CD, COCOSYS and ASTEC is done as well at GRS but it is not discussed in this paper. GRS has been using MELCOR since 1990 for real plant calculations. The results of MELCOR analyses are used mainly in PSA level 2 studies and in Accident Management projects for both types of NPPs. MELCOR has been a very useful and robust tool for these analyses. The calculations performed within the PSA level 2 studies for both types of German NPPs have shown that typical severe accident scenarios are characterized by several phases and that the consideration of plant specifics are important not only for realistic source term calculations. An overview of typically severe accident phases together with main accident management measures installed in German NPPs is presented in the paper. Several severe accident sequences have been calculated for both reactor types and some detailed nodalisation studies and code to code comparisons have been prepared in the past, to prove the developed core, reactor circuit and containment/building nodalisation schemes. Together with the compilation of the MELCOR data set, the qualification of the nodalisation schemes has been pursued with comparative calculations with detailed GRS codes for selected phases of severe accidents. The results of these comparative analyses showed in most of the areas a good agreement of essential parameters and of the general description of the plant behaviour during the accident progression. The in general detail of the German plant nodalisation schemes developed for MELCOR contributes significantly to this good agreement between integral and detailed code results. The implementation of MELCOR into the GRS simulator ATLAS was very important for the assessment of the results, not only due to the great detail of the nodalisation schemes used. It is used for training of severe accident

  17. Modifications made to the MELCOR Code for Analyzing Lithium Fires in Fusion Reactors

    International Nuclear Information System (INIS)

    Merrill, B.J.

    2000-01-01

    This report documents initial modifications made to the MELCOR code that allows MELCOR to predict the consequences of lithium spill accidents for evolving fusion reactor designs. These modifications include thermodynamic and transport properties for lithium, and physical models for predicting the rate of reaction of and energy production from the lithium-air reaction. A benchmarking study was performed with this new MELCOR capability. Two lithium-air reaction tests conducted at the Hanford Engineering Development Laboratory (HEDL) were selected for this benchmark study. Excellent agreement was achieved between MELCOR predictions and measured data. Recommendations for modeling lithium fires with MELCOR and for future work in this area are included in this report

  18. MELCOR computer code manuals

    Energy Technology Data Exchange (ETDEWEB)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.

  19. Improved hydrogen distribution calculation in the containment using the coupled codes MELCOR and GASFLOW for the analysis of severe accidents in nuclear power plants; Verbesserte Berechnung der Wasserstoffverteilung im Sicherheitsbehaelter bei der Analyse schwerer Stoerfaelle in Kernkraftwerken durch Kopplung von MELCOR und GASFLOW

    Energy Technology Data Exchange (ETDEWEB)

    Szabo, Tobias

    2014-09-01

    The risk of a hydrogen combustion within a containment of a pressurized water reactor during a severe loss of coolant accident (LOCA) is evaluated using numerical simulations. The code MELCOR provides integral analysis capabilities for severe accidents. Yet, its Lumped Parameter (LP) model provides less accurate information about on thermal hydraulics within the containment during a LOCA. GASFLOW is a CFD code that simulates both the local hydrogen distribution and the pressure inside the containment more realistically. Currently, to perform these GASFLOW simulations, the common procedure is to use a source term from a previous MELCOR calculation. However, with this approach, the influence of the more realistic GASFLOW pressure on the mass flow through the leak cannot be taken into account. This inconsistency is overcome by coupling both codes in this thesis. Here, the MELCOR instance is responsible for the primary and secondary systems. At the same time, the GASFLOW instance predicts the thermal hydraulics of the containment. The more accurate containment pressure from the GASFLOW instance is used in the MELCOR instance to calculate consistent outflow rates through the leak. In order to couple both codes, the existing interface in MELCOR is modified and a new interface for GASFLOW is developed and implemented. To begin with, the hydrogen distribution inside a generic containment is calculated by MELCOR using a typical coarse LP nodalization and a refined one. The results obtained are compared to a GASFLOW simulation. It is shown that the refinement only leads to a better agreement with the GASFLOW result if the correct flow directions are predefined by the nodalization. The safety relevant, local peak concentrations of hydrogen cannot be resolved by MELCOR. Consequently, the use of the CFD code is indispensable. The correct functioning of the coupling is proven within four steps. At first, the modified MELCOR interface is checked by computing a test case using two

  20. MELCOR computer code manuals

    International Nuclear Information System (INIS)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package

  1. Analysis of steam generator tube rupture as a severe accident using MELCOR 1.8.4

    International Nuclear Information System (INIS)

    Yang Hongrun; Hidaka, Akihide; Sugimoto, Jun

    1999-03-01

    This report presents the results from the MELCOR 1.8.4 calculations for Steam Generator Tube Rupture (SGTR) with stuck open of all the safety valves in faulted SG as a severe accident. The calculations are based on Surry nuclear power plant. After performed using the once-through primary system model alone by 1.0x10 5 s, the calculations were conducted with both of the once-through and the hot leg countercurrent natural circulation models. The results, including event sequences, processes and progressions of core degradation, radionuclides release from core and reactor cavity, and source terms to the environment are described in detail. It is concluded that the availability of High Pressure Safety Injection (HPSI) can significantly delay the progression of core heat-up and approximately 7% of cesium iodide (CsI) can be released to the environment directly through the stuck open safety valve. Comparisons between the results from the two models are also given in this report. The present analyses also showed that during SGTR accident, the hot leg countercurrent natural circulation flow cannot be established well and therefore it has little effect on the mitigation of the core degradation. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  3. A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR ampersand BWR accident sequences

    International Nuclear Information System (INIS)

    Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

    1996-01-01

    This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences

  4. Comparative accident analyses for a KONVOI-type PWR using the integral codes ASTEC V1.33 and MELCOR 1.8.6

    International Nuclear Information System (INIS)

    Reinke, Nils; Erdmann, Walter; Nowack, Holger; Sonnenkalb, Martin

    2010-08-01

    In the frame of the project RS1180 funded by the German Federal Ministry for Economics and Technology (BMWi) calculations have been carried out with the integral code ASTEC V1.33 p3 developed by GRS for two postulated accidents in a nuclear power plant with KONVOI type a pressurized water reactor and compared to calculations with MELCOR 1.8.6 YU. Major objective was to assess the capability of ASTEC for application in level 2 probabilistic safety analyses (PSA). In particular, it was investigated to which extent ASTEC is able to perform such integral calculations meeting criteria with regard to both reasonable calculation time and specific boundary conditions necessary for PSA analyses. Two exemplary accidents were selected: - A transient with loss of steam generator feed water, - A small break loss of coolant accident (50 cm 2 ) in the cold leg of the coolant line connected to the pressurizer. In principle, the results demonstrate the capability of ASTEC V1.33 to carry out such PSA level 2 calculations. In addition, it has to be noted that for both ASTEC and MELCOR the requirements in view of the quality of the results leads to prolonged calculation times due to more detailed nodalisations of the whole plant. This is valid for the core region as well as for the primary circuit and for the containment. Consequently, calculation times in the order of one day to two weeks are accomplished, thereby excluding extensive parameter analyses in order to assess the sensitivity of the calculation results. Concerning the quality of the results a good agreement can be stated between ASTEC and MELCOR results in terms of global data. In detail some results are sensitive to user effects. Here, the nodalisation seems to be of major influence besides differences in modeling specific phenomena. The comparison suggests that in particular the influence of the nodalisation defined by the user and depending on the user's experience should be carefully evaluated. Since some

  5. 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)

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

    International Nuclear Information System (INIS)

    De Rosa, Felice

    2006-01-01

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

  7. A restructuring proposal based on MELCOR for severe accident analysis code development

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sun Hee; Song, Y. M.; Kim, D. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    In order to develop a template based on existing MELCOR code, current data saving and transferring methods used in MELCOR are addressed first. Then a naming convention for the constructed module is suggested and an automatic program to convert old variables into new derived type variables has been developed. Finally, a restructured module for the SPR package has been developed to be applied to MELCOR. The current MELCOR code ensures a fixed-size storage for four different data types, and manages the variable-sized data within the storage limit by storing the data on the stacked packages. It uses pointer to identify the variables between the packages. This technique causes a difficult grasping of the meaning of the variables as well as memory waste. New features of FORTRAN90, however, make it possible to allocate the storage dynamically, and to use the user-defined data type which lead to a restructured module development for the SPR package. An efficient memory treatment and as easy understanding of the code are allowed in this developed module. The validation of the template has been done by comparing the results of the modified code with those from the existing code, and it is confirmed that the results are the same. The template for the SPR package suggested in this report hints the extension of the template to the entire code. It is expected that the template will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models. 3 refs., 15 figs., 16 tabs. (Author)

  8. Development of a severe accident module of a nuclear power plant based in the MELCOR nuclear code and its incorporation to the room simulator

    International Nuclear Information System (INIS)

    Cortes M, F.S.; Ramos P, J.C.; Nelson E, P.; Chavez M, C.

    2004-01-01

    This work describes the development of the Severe Accidents Module (MAS) based on the Code MELCOR and its incorporation to the Simulator of Classroom of the Group of Nuclear Engineering of the Engineering Faculty (GrINFI) of the National Autonomous University of Mexico (UNAM). The module of Severe Accidents has the purpose of counting with installed and operational capacity for the simulation of accident sequences with capacitation purposes, training, analysis and design. A shallow description of SimAula is presented, and the philosophy used to obtain the interactive version of MELCOR are discussed, as well as its implementation in the atmosphere of SimAula. Finally, after confirming the correct operation of the development of the tool, some possible topics are discussed for specific applications of the MAS. (Author)

  9. Prediction of corium debris characteristics in lower plenum of a nordic BWR in different accident scenarios using MELCOR code - 15367

    International Nuclear Information System (INIS)

    Phung, V.A.; Galushin, S.; Raub, S.; Goronovski, A.; Villanueva, W.; Koeoep, K; Grishchenko, D.; Kudinov, P.

    2015-01-01

    Severe accident management strategy in Nordic boiling water reactors (BWRs) relies on ex-vessel core debris coolability. The mode of corium melt release from the vessel determines conditions for ex-vessel accident progression and threats to containment integrity, e.g., formation of a non-coolable debris bed and possibility of energetic steam explosion. In-vessel core degradation and relocation is an important stage which determines characteristics of corium debris in the vessel lower plenum, such as mass, composition, thermal properties, timing of relocation, and decay heat. These properties affect debris reheating and remelting, melt interactions with the vessel structures, and possibly vessel failure and melt ejection mode. Core degradation and relocation is contingent upon the accident scenario parameters such as recovery time and capacity of safety systems. The goal of this work is to obtain a better understanding of the impact of the accident scenarios and timing of the events on core relocation phenomena and resulting properties of the debris bed in the vessel lower plenum of Nordic BWRs. In this study, severe accidents in a Nordic BWR reference plant are initiated by a station black out event, which is the main contributor to core damage frequency of the reactor. The work focuses on identifying ranges of debris bed characteristics in the lower plenum as functions of the accident scenario with different recovery timing and capacity of safety systems. The severe accident analysis code MELCOR coupled with GA-IDPSA is used in this work. GA-IDPSA is a Genetic Algorithm-based Integrated Deterministic Probabilistic Safety Analysis tool, which has been developed to search uncertain input parameter space. The search is guided by different target functions. Scenario grouping and clustering approach is applied in order to estimate the ranges of debris characteristics and identify scenario regions of core relocation that can lead to significantly different debris bed

  10. MELCOR 1.8.3 application to NUPEC M-7-1 test (ISP-35) and two hydrogen severe accident scenarios in a typical PWR plant

    International Nuclear Information System (INIS)

    Jimenez Garcia, M.A.; Martin-Fuertes, F.; Martin-Valdepenas, J.M.

    1997-01-01

    Combustion of the hydrogen released to the containment during a severe accident is one of the issues to establish the real threats to the third barrier integrity in nuclear power facilities. Computational efforts on management procedures, such as the containment spray operation, are being addressed at the CTN-UPM to cope with the problem. On top of this, studies about in-containment hydrogen distribution and combustion are currently carried out with the codes MELCOR 1.8.3 and ESTER 1.0-RALOC 2.2. In this study, MELCOR 1.8.3 has been validated against the NUPEC M-7-1 Test, which already showed in 1993 that a good agreement was reached out when the previous MELCOR 1.8.2 calculations were performed regarding to the helium distribution throughout the facility. Nevertheless, some discrepancies were detected when analysing wall and atmosphere temperatures. Generally, well-mixed atmosphere scenarios, in which the role played by the containment water spraying is of the major importance, appear when such a mechanism promotes the onset of convection driven flow patterns that rapidly homogenize the gas properties. The purpose of the new MELCOR 1.8.3 assessment is to take advantage of the newest implemented models to obtain a more realistic thermalhydraulics simulation. A variation case was also performed to highlight the influence of water spray operation. In a second part of the study, insights coming from the previous work were used to apply MELCOR 1.8.3 models to a SBO severe accident scenario management in a commercial 2700 MWt 3-loop W PWR containment

  11. 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.

  12. 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

  13. 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)

  14. Accident Source Terms for Pressurized Water Reactors with High-Burnup Cores Calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Goldmann, Andrew; Kalinich, Donald A.; Powers, Dana A.

    2016-12-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs 2 MoO 4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  15. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  16. The analysis of thermal-hydraulic models in MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M H; Hur, C; Kim, D K; Cho, H J [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)

    1996-07-15

    The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed.

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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

  2. 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

  3. 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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  5. 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.

  6. Quench/reflood modeling in MELCOR

    International Nuclear Information System (INIS)

    Gauntt, R.O.

    2001-01-01

    The authors describe the reactor accident simulation model MELCOR. It comprises hydrodynamic investigations on reactor core quenching, hydrogen generation in the reactor core vessel, quench front advances. Preliminary comparisons to data are reasonable but need further validation. (uke)

  7. 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

  8. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2

    International Nuclear Information System (INIS)

    Kmetyk, L.N.; Brown, T.D.

    1995-03-01

    To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP ampersand S) modes of operation at commercial nuclear power plants. To investigate the likelihood of severe core damage accidents during off power conditions, probabilistic risk assessments (PRAs) were performed for two nuclear plants: Unit 1 of the Grand Gulf Nuclear Station, which is a BWR-6 Mark III boiling water reactor (BWR), and Unit 1 of the Surry Power Station, which is a three-loop, subatmospheric, pressurized water reactor (PWR). The analysis of the BWR was conducted at Sandia National Laboratories while the analysis of the PWR was performed at Brookhaven National Laboratory. This multi-volume report presents and discusses the results of the BWR analysis. The subject of this part presents the deterministic code calculations, performed with the MELCOR code, that were used to support the development and quantification of the PRA models. The background for the work documented in this report is summarized, including how deterministic codes are used in PRAS, why the MELCOR code is used, what the capabilities and features of MELCOR are, and how the code has been used by others in the past. Brief descriptions of the Grand Gulf plant and its configuration during LP ampersand S operation and of the MELCOR input model developed for the Grand Gulf plant in its LP ampersand S configuration are given

  9. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.; Brown, T.D. [Sandia National Labs., Albuquerque, NM (United States)

    1995-03-01

    To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP&S) modes of operation at commercial nuclear power plants. To investigate the likelihood of severe core damage accidents during off power conditions, probabilistic risk assessments (PRAs) were performed for two nuclear plants: Unit 1 of the Grand Gulf Nuclear Station, which is a BWR-6 Mark III boiling water reactor (BWR), and Unit 1 of the Surry Power Station, which is a three-loop, subatmospheric, pressurized water reactor (PWR). The analysis of the BWR was conducted at Sandia National Laboratories while the analysis of the PWR was performed at Brookhaven National Laboratory. This multi-volume report presents and discusses the results of the BWR analysis. The subject of this part presents the deterministic code calculations, performed with the MELCOR code, that were used to support the development and quantification of the PRA models. The background for the work documented in this report is summarized, including how deterministic codes are used in PRAS, why the MELCOR code is used, what the capabilities and features of MELCOR are, and how the code has been used by others in the past. Brief descriptions of the Grand Gulf plant and its configuration during LP&S operation and of the MELCOR input model developed for the Grand Gulf plant in its LP&S configuration are given.

  10. 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)

  11. Development of a severe accident module of a nuclear power plant based in the MELCOR nuclear code and its incorporation to the room simulator; Desarrollo del modulo de accidentes severos de una central nucleoelectrica basado en el codigo nuclear MELCOR y su incorporacion al simulador de aula

    Energy Technology Data Exchange (ETDEWEB)

    Cortes M, F.S.; Ramos P, J.C.; Nelson E, P.; Chavez M, C. [Facultad de Ingenieria, Division de Ingenieria Electrica, Grupo de Ingenieria Nuclear, UNAM, Ciudad Universitaria, Distrito Federal (Mexico)]. E-mail: samuelcortes@correo.unam.mx

    2004-07-01

    This work describes the development of the Severe Accidents Module (MAS) based on the Code MELCOR and its incorporation to the Simulator of Classroom of the Group of Nuclear Engineering of the Engineering Faculty (GrINFI) of the National Autonomous University of Mexico (UNAM). The module of Severe Accidents has the purpose of counting with installed and operational capacity for the simulation of accident sequences with capacitation purposes, training, analysis and design. A shallow description of SimAula is presented, and the philosophy used to obtain the interactive version of MELCOR are discussed, as well as its implementation in the atmosphere of SimAula. Finally, after confirming the correct operation of the development of the tool, some possible topics are discussed for specific applications of the MAS. (Author)

  12. 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

  13. 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

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. 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

  19. Comparison of CORA and MELCOR core degradation simulation and the MELCOR oxidation model

    International Nuclear Information System (INIS)

    Wang, Jun; Corradini, Michael L.; Fu, Wen; Haskin, Troy; Tian, Wenxi; Zhang, Yapei; Su, Guanghui; Qiu, Suizheng

    2014-01-01

    Highlights: • Oxidation model of MELCOR is analyzed and the improving suggestion is provided. • MELCOR core degradation calculating results are compared with CORA experiment. • Flow rate of argon and steam, the generating rate of hydrogen is calculated and compared. • Temperature spatial variation and temperature history is calculated and presented. - Abstract: MELCOR is widely used and sufficiently trusted for severe accident analysis. However, the occurrence of Fukushima has increased the focus on severe accident codes and their use. A MELCOR core degradation calculation was conducted at the University of Wisconsin–Madison under the help of Sandia. The calculation results were checked by comparing with a past CORA experiment. MELCOR calculation results included the flow rate of argon and steam, the generation rate of hydrogen. Through this work, the performance of MELCOR COR package was reviewed in detail. This paper compares the hydrogen generation rates predicted by MELCOR to the CORA test data. While agreement is reasonable it could be improved. Additionally, the MELCOR zirconium oxidation model was analyzed

  20. Comparison of CORA and MELCOR core degradation simulation and the MELCOR oxidation model

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jun [College of Engineering, The University of Wisconsin-Madison, Madison, WI 53706 (United States); State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Corradini, Michael L., E-mail: corradini@engr.wisc.edu [College of Engineering, The University of Wisconsin-Madison, Madison, WI 53706 (United States); Fu, Wen [College of Engineering, The University of Wisconsin-Madison, Madison, WI 53706 (United States); Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Haskin, Troy [College of Engineering, The University of Wisconsin-Madison, Madison, WI 53706 (United States); Tian, Wenxi; Zhang, Yapei; Su, Guanghui; Qiu, Suizheng [State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China)

    2014-09-15

    Highlights: • Oxidation model of MELCOR is analyzed and the improving suggestion is provided. • MELCOR core degradation calculating results are compared with CORA experiment. • Flow rate of argon and steam, the generating rate of hydrogen is calculated and compared. • Temperature spatial variation and temperature history is calculated and presented. - Abstract: MELCOR is widely used and sufficiently trusted for severe accident analysis. However, the occurrence of Fukushima has increased the focus on severe accident codes and their use. A MELCOR core degradation calculation was conducted at the University of Wisconsin–Madison under the help of Sandia. The calculation results were checked by comparing with a past CORA experiment. MELCOR calculation results included the flow rate of argon and steam, the generation rate of hydrogen. Through this work, the performance of MELCOR COR package was reviewed in detail. This paper compares the hydrogen generation rates predicted by MELCOR to the CORA test data. While agreement is reasonable it could be improved. Additionally, the MELCOR zirconium oxidation model was analyzed.

  1. 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)

  2. An uncertainty analysis of the hydrogen source term for a station blackout accident in Sequoyah using MELCOR 1.8.5

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Bixler, Nathan E.; Wagner, Kenneth Charles.

    2014-03-01

    A methodology for using the MELCOR code with the Latin Hypercube Sampling method was developed to estimate uncertainty in various predicted quantities such as hydrogen generation or release of fission products under severe accident conditions. In this case, the emphasis was on estimating the range of hydrogen sources in station blackout conditions in the Sequoyah Ice Condenser plant, taking into account uncertainties in the modeled physics known to affect hydrogen generation. The method uses user-specified likelihood distributions for uncertain model parameters, which may include uncertainties of a stochastic nature, to produce a collection of code calculations, or realizations, characterizing the range of possible outcomes. Forty MELCOR code realizations of Sequoyah were conducted that included 10 uncertain parameters, producing a range of in-vessel hydrogen quantities. The range of total hydrogen produced was approximately 583kg 131kg. Sensitivity analyses revealed expected trends with respected to the parameters of greatest importance, however, considerable scatter in results when plotted against any of the uncertain parameters was observed, with no parameter manifesting dominant effects on hydrogen generation. It is concluded that, with respect to the physics parameters investigated, in order to further reduce predicted hydrogen uncertainty, it would be necessary to reduce all physics parameter uncertainties similarly, bearing in mind that some parameters are inherently uncertain within a range. It is suspected that some residual uncertainty associated with modeling complex, coupled and synergistic phenomena, is an inherent aspect of complex systems and cannot be reduced to point value estimates. The probabilistic analyses such as the one demonstrated in this work are important to properly characterize response of complex systems such as severe accident progression in nuclear power plants.

  3. 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.)

  4. Development of the stationary state and simulation of an accident severe stage type station blackout with the MELCOR code version 1.8.6 for the nuclear power plant of Laguna Verde

    International Nuclear Information System (INIS)

    Mugica R, C. A.; Godinez S, V.

    2011-11-01

    Considering the events happened since the 11 March of 2011, in Japan, where an earthquake of 9.1 grades Ritcher of intensity and a later tsunami impacted in an important way the operation of a nuclear power plant located in the Fukushima, Japan; damaging and disabling their cooling systems and injection of emergency water due to the total loss of electric power (commonly denominated Station Blackout), is eminent the analysis of this stage type that took to the nuclear power plant to conditions of damage to the core and explosions generation by hydrogen concentrations in the reactor building. In this work an analysis of a stage type station blackout is presented, using the model of the nuclear power plant of Laguna Verde starting of the stationary state. The analysis is carried out using the MELCOR code (Methods for Estimation of Leakages and Consequences of Releases) version 1.8.6 whose purpose is to model the accidents progression for light water reactors. The obtained results are qualitatively similar to the events observed in the Fukushima nuclear power plant even though limitations exist to achieve a precise simulation of the events happened in Japan, such as the information flow of the chronology of the operator actions, as well as of the characteristic design of the power plant, volumes in cavities and rooms, water/cooling inventories, interconnected systems and their own emergency procedures or guides for the administration of severe accidents among others. (Author)

  5. MELCOR code modeling for APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young; Park, S. Y.; Kim, D. H.; Ahn, K. I.; Song, Y. M.; Kim, S. D.; Park, J. H

    2001-11-01

    The severe accident phenomena of nuclear power plant have large uncertainties. For the retention of the containment integrity and improvement of nuclear reactor safety against severe accident, it is essential to understand severe accident phenomena and be able to access the accident progression accurately using computer code. Furthermore, it is important to attain a capability for developing technique and assessment tools for an advanced nuclear reactor design as well as for the severe accident prevention and mitigation. The objective of this report is to establish technical bases for an application of the MELCOR code to the Korean Next Generation Reactor (APR1400) by modeling the plant and analyzing plant steady state. This report shows the data and the input preparation for MELCOR code as well as state-state assessment results using MELCOR code.

  6. 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

  7. Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Min; Sabundjian, Gaianê, E-mail: smlee@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

  8. Simulation with the MELCOR code of two severe accident sequences, Station Blackout and Small Break LOCA, for the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Valle Cepero, Reinaldo

    2004-01-01

    The results of the PSA-I applied to the Atucha I nuclear power plant (CNA I) determine the accidental sequences with the most influence related to the probability of the core reactor damage. Among those sequences are include, the Station Blackout and lost of primary coolant, combine with the failure of the emergency injection systems by pipe breaks of diameters between DN100 - DN25 or equivalent areas, Small LOCA. This paper has the objective to model and analyze the behavior of the primary circuit and the pressure vessel during the evolution of those two accidental sequences. It presented a detailed analysis of the main phenomena that occur from the initial moment of the accident to the failure moment of the pressure vessel and the melt material fall to the reactor cavity. Two sequences were taken into account, considering the main phenomena (core uncover, heating, fuel element oxidation, hydrogen generation, degradation and relocation of the melt material, failure of the support structures, etc.) and the time of occurrence, of those events will be different, if it is considered that both sequences will be developed in different scenarios. One case is an accident with the primary circuit to a high pressure (Station Blackout scenario) and the other with a early primary circuit depressurization due to the lost of primary coolant. For this work the MELCOR 1.8.5 code was used and it allows within a unified framework to modeling an extensive spectrum of phenomenology associated with the severe accidents. (author)

  9. Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

    International Nuclear Information System (INIS)

    Lee, Seung Min; Sabundjian, Gaianê

    2017-01-01

    The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

  10. 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

  11. 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

  12. MELCOR 1.8.2 Analyses in Support of ITER's RPrS

    International Nuclear Information System (INIS)

    Brad J Merrill

    2008-01-01

    The International Thermonuclear Experimental Reactor (ITER) Program is performing accident analyses for ITER's 'Rapport Preliminaire de Surete' (Report Preliminary on Safety - RPrS) with a modified version of the MELCOR 1.8.2 code. The RPrS is an ITER safety document required in the ITER licensing process to obtain a 'Decret Autorisation de Construction' (a Decree Authorizing Construction - DAC) for the ITER device. This report documents the accident analyses performed by the US with the MELCOR 1.8.2 code in support of the ITER RPrS effort. This work was funded through an ITER Task Agreement for MELCOR Quality Assurance and Safety Analyses. Under this agreement, the US was tasked with performing analyses for three accident scenarios in the ITER facility. Contained within the text of this report are discussions that identify the cause of these accidents, descriptions of how these accidents are likely to proceed, the method used to analyze the consequences of these accidents, and discussions of the transient thermal hydraulic and radiological release results for these accidents

  13. 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.

  14. Containment Sodium Chemistry Models in MELCOR.

    Energy Technology Data Exchange (ETDEWEB)

    Louie, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Humphries, Larry L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Denman, Matthew R [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-04-01

    To meet regulatory needs for sodium fast reactors’ future development, including licensing requirements, Sandia National Laboratories is modernizing MELCOR, a severe accident analysis computer code developed for the U.S. Nuclear Regulatory Commission (NRC). Specifically, Sandia is modernizing MELCOR to include the capability to model sodium reactors. However, Sandia’s modernization effort primarily focuses on the containment response aspects of the sodium reactor accidents. Sandia began modernizing MELCOR in 2013 to allow a sodium coolant, rather than water, for conventional light water reactors. In the past three years, Sandia has been implementing the sodium chemistry containment models in CONTAIN-LMR, a legacy NRC code, into MELCOR. These chemistry models include spray fire, pool fire and atmosphere chemistry models. Only the first two chemistry models have been implemented though it is intended to implement all these models into MELCOR. A new package called “NAC” has been created to manage the sodium chemistry model more efficiently. In 2017 Sandia began validating the implemented models in MELCOR by simulating available experiments. The CONTAIN-LMR sodium models include sodium atmosphere chemistry and sodium-concrete interaction models. This paper presents sodium property models, the implemented models, implementation issues, and a path towards validation against existing experimental data.

  15. 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

  16. A step forward towards understanding the severe accident evolution in Fukushima: Parametric analyses of Unit-I with MELCOR 2.I

    International Nuclear Information System (INIS)

    Herranz, L. E.; Garcia Martin, M.; Lopez del Pra, C.

    2013-01-01

    The Fukushima accident on March 11 2011, largely affected nuclear community all over the world. Immediately, many R and D organizations and companies drew their attention to the facts that were being released by Japanese authorities, so that the most through understanding of the situation could be gained at the moment. This paper synthesizes the major progress achieved by CIEMAT in this international environment. The specific target pursued is the comprehension of the severe accident evolution, particularly the progression followed in the unit 1 of Fukushima Daiichi. To do so, the MELCOR 2.1 code has been used to model 6 Fukushima-like SBO sequences in a BWR-Mark I reactor. Particular emphasis has been given to understand the potential impact of systems like the isolation condenser (IC) and the over-pressure protection system. As expected, availability of AC power from a diesel would have drastically changed the scenario, particularly in the containment, due to actuation of DW (Dry-Well) sprays. IC actuation might have delayed the reactor pressure vessel (RPV) failure for hours, if they had performed according to their design. Contrarily, a stuck-open safety relief valve would have resulted in an earlier PRV failure. A definitive picture of the scenario is still far away; however, these results may be seen as a step forward. (Author)

  17. A step forward towards understanding the severe accident evolution in Fukushima: Parametric analyses of Unit-I with MELCOR 2.I

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Garcia Martin, M.; Lopez del Pra, C.

    2013-03-01

    The Fukushima accident on March 11 2011, largely affected nuclear community all over the world. Immediately, many R and D organizations and companies drew their attention to the facts that were being released by Japanese authorities, so that the most through understanding of the situation could be gained at the moment. This paper synthesizes the major progress achieved by CIEMAT in this international environment. The specific target pursued is the comprehension of the severe accident evolution, particularly the progression followed in the unit 1 of Fukushima Daiichi. To do so, the MELCOR 2.1 code has been used to model 6 Fukushima-like SBO sequences in a BWR-Mark I reactor. Particular emphasis has been given to understand the potential impact of systems like the isolation condenser (IC) and the over-pressure protection system. As expected, availability of AC power from a diesel would have drastically changed the scenario, particularly in the containment, due to actuation of DW (Dry-Well) sprays. IC actuation might have delayed the reactor pressure vessel (RPV) failure for hours, if they had performed according to their design. Contrarily, a stuck-open safety relief valve would have resulted in an earlier PRV failure. A definitive picture of the scenario is still far away; however, these results may be seen as a step forward. (Author)

  18. Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde

    International Nuclear Information System (INIS)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V.

    2013-10-01

    In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft 2 (4.6 cm 2 ), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

  19. 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

  20. 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

  1. 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

  2. The MELCOR peer review process and findings

    International Nuclear Information System (INIS)

    Boyack, B.E.; Dhir, V.K.; Haste, T.J.; Gieseke, J.A.; Viskanta, R.; Kenton, M.A.; Khatib-Rahbar, M.; Leonard, M.T.

    1991-01-01

    MELCOR is a fully integrated, engineering-level computer code the models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission (NRC) as a second-generation plant risk assessment tool and as the successor to the Source Term Code Package. MELCOR has been under development since 1982. The code has now reached sufficient maturity that a number of organizations inside and outside the NRC are using or are planning to use the code. Although the quality control and validation efforts are in progress, the NRC identified the need to have a broad technical review of recognized experts to determine or confirm the technical adequacy of the code for the serious and complex analyses it is expected to perform. A peer review committee was organized using recognized experts from the national laboratories, universities, MELCOR user community, and independent contractors to perform this assessment. The objective of this paper is to summarize the peer review process and to summarize the findings of the MELCOR Peer Review Committee formed to conduct the MELCOR peer review

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. Melcor benchmarking against integral severe fuel damage tests

    Energy Technology Data Exchange (ETDEWEB)

    Madni, I.K. [Brookhaven National Lab., Upton, NY (United States)

    1995-09-01

    MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the U.S. Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC to provide independent assessment of MELCOR, and a very important part of this program is to benchmark MELCOR against experimental data from integral severe fuel damage tests and predictions of that data from more mechanistic codes such as SCDAP or SCDAP/RELAP5. Benchmarking analyses with MELCOR have been carried out at BNL for five integral severe fuel damage tests, namely, PBF SFD 1-1, SFD 14, and NRU FLHT-2, analyses, and their role in identifying areas of modeling strengths and weaknesses in MELCOR.

  11. 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

  12. IPLOT, interactive MELCOR data plotting system

    International Nuclear Information System (INIS)

    2008-01-01

    1 - Description of program or function: IPLOT is an interactive MELCOR data plotting system. It provides several kinds of GUI interfaces for a flexible data plotting. IPLOT capabilities include creation, saving and loading of user specified MELCOR variables trend graphs. IPLOT can use one or several plot files for a graph generation while the graphs can be either in one window or in several windows. Besides IPLOT provides several graph convenient functions such as zooming, re-sizing, printing for a detail analysis of severe accidents. 2 - Methods: Trend values seeking in a plot file is performed by a binary search method for fast performance. 3 - Restrictions on the complexity of the problem: MELCOR plot files are required for plotting

  13. Progress in MELCOR development and assessment

    International Nuclear Information System (INIS)

    Summers, R.M.; Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Elsbernd, A.E.; Stuart, D.S.; Thompson, S.L.

    1995-01-01

    MELCOR models the progression of severe accidents in light water reactor nuclear power plants. Recent efforts in MELCOR development to incorporate CORCON-Mod3 models for core-concrete interactions, new models for advanced reactors, and improvements to several other existing models have resulted in release of MELCOR 1.8.3. In addition, continuing efforts to expand the code assessment database have filled in many of the gaps in phenomenological coverage. Efforts are now under way to develop models for chemical interactions of fission products with structural surfaces and for reactions of iodine in the presence of water, and work is also in progress to improve models for the scrubbing of fission products by water pools, the chemical reactions of boron carbide with steam, and the coupling of flow blockages with the hydrodynamics. Several code assessment analyses are in progress, and more are planned

  14. 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.

  15. 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.)

  16. Experience and results of MELCOR application for German PWRs

    International Nuclear Information System (INIS)

    Sonnenkalb, M.

    1999-01-01

    An introduction into severe accident research work performed at GRS with regard to the use of the MELCOR code is given in Chapter One of the paper. Experience in applying MELCOR 1.8.3 for German PWRs and results of MELCOR calculations done within the project 'Accident management - Mitigation' for German LWRs are presented in Chapter Two. This 3-year project was finished February 1998. It was funded by the German Ministry for Environment, Nature Conservation and Nuclear Safety - BMU. In Chapter Three, a short overview of a training course on 'Phenomenology of Severe Accidents in PWR-Plants' is given. Mainly due to the interest from German NPPs GRS developed this special training session in 1996. Since 1996 it has been held several times for operators, shift personnel and the management board of two different German NPPs and for lecture of the German NPP training centre in Essen. In Chapter Four, results of the application of MELCOR 1.8.4 for German PWRs are presented. This work is done within a new project on 'Accident Management - Mitigation' for German LWRs. It was started in March 1998 and is again funded by the German Federal Ministry BMU. An objective of this project is to perform further MELCOR calculations, to be used within a PSA level 2 study for a German PWR, which is done at GRS in parallel. The experience of using MELCOR for German PWRs are summarised in Chapter Five. (author)

  17. 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

  18. 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.)

  19. Liftoff Model for MELCOR.

    Energy Technology Data Exchange (ETDEWEB)

    Young, Michael F. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-07-01

    Aerosol particles that deposit on surfaces may be subsequently resuspended by air flowing over the surface. A review of models for this liftoff process is presented and compared to available data. Based on this review, a model that agrees with existing data and is readily computed is presented for incorporation into a system level code such as MELCOR. Liftoff Model for MELCOR July 2015 4 This page is intentionally blank

  20. 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

  1. 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

  2. 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)

  3. 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)

  4. 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)

  5. 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)

  6. 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)

  7. 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

  8. 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)

  9. 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

  10. 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)

  11. 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)

  12. 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)

  13. Ex-plant consequence assessment for NUREG-1150: models, typical results, uncertainties

    International Nuclear Information System (INIS)

    Sprung, J.L.

    1988-01-01

    The assessment of ex-plant consequences for NUREG-1150 source terms was performed using the MELCOR Accident Consequence Code System (MACCS). This paper briefly discusses the following elements of MACCS consequence calculations: input data, phenomena modeled, computational framework, typical results, controlling phenomena, and uncertainties. Wherever possible, NUREG-1150 results will be used to illustrate the discussion. 28 references

  14. 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)

  15. 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

  16. 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.)

  17. 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)

  18. 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

  19. 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.

  20. 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)

  1. Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

    International Nuclear Information System (INIS)

    Lindholm, I.; Pekkarinen, E.; Sjoevall, H.

    1995-01-01

    Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H 2 produced, though the trends in H 2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

  2. 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)

  3. Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde; Simulacion de escenarios de accidente severo tipo perdida de refrigerante (Loca) pequeno, con el codigo MELCOR version 2.1 para la central nucleo-electrica de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft{sup 2} (4.6 cm{sup 2}), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

  4. NSRD-10: Leak Path Factor Guidance Using MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Louie, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Humphries, Larry L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-03-01

    Estimates of the source term from a U.S. Department of Energy (DOE) nuclear facility requires that the analysts know how to apply the simulation tools used, such as the MELCOR code, particularly for a complicated facility that may include an air ventilation system and other active systems that can influence the environmental pathway of the materials released. DOE has designated MELCOR 1.8.5, an unsupported version, as a DOE ToolBox code in its Central Registry, which includes a leak-path-factor guidance report written in 2004 that did not include experimental validation data. To continue to use this MELCOR version requires additional verification and validations, which may not be feasible from a project cost standpoint. Instead, the recent MELCOR should be used. Without any developer support and lack of experimental data validation, it is difficult to convince regulators that the calculated source term from the DOE facility is accurate and defensible. This research replaces the obsolete version in the 2004 DOE leak path factor guidance report by using MELCOR 2.1 (the latest version of MELCOR with continuing modeling development and user support) and by including applicable experimental data from the reactor safety arena and from applicable experimental data used in the DOE-HDBK-3010. This research provides best practice values used in MELCOR 2.1 specifically for the leak path determination. With these enhancements, the revised leak-path-guidance report should provide confidence to the DOE safety analyst who would be using MELCOR as a source-term determination tool for mitigated accident evaluations.

  5. 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.

  6. 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

  7. 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.)

  8. 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

  9. Insight from Fukushima Daiichi Unit 3 Investigations using MELCOR

    International Nuclear Information System (INIS)

    Robb, Kevin R.; Francis, Matthew W.; Ott, Larry J.

    2014-01-01

    During the emergency response period of the accidents that took place at Fukushima Daiichi in March of 2011, researchers at Oak Ridge National Laboratory (ORNL) conducted a number of studies using the MELCOR code to help understand what was occurring and what had occurred. During the post-accident period, the Department of Energy (DOE) and the US Nuclear Regulatory Commission (NRC) jointly sponsored a study of the Fukushima Daiichi accident with collaboration among Oak Ridge, Sandia, and Idaho national laboratories. The purpose of the study was to compile relevant data, reconstruct the accident progression using computer codes, assess the codes predictive capabilities, and identify future data needs. The current paper summarizes some of the early MELCOR simulations and analyses conducted at ORNL of the Fukushima Daiichi Unit 3 accident. Extended analysis and discussion of the Unit 3 accident is also presented taking into account new knowledge and modeling refinements made since the joint DOE/NRC study

  10. 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

  11. Improvement of a combustion model in MELCOR code

    International Nuclear Information System (INIS)

    Ogino, Masao; Hashimoto, Takashi

    1999-01-01

    NUPEC has been improving a hydrogen combustion model in MELCOR code for severe accident analysis. In the proposed combustion model, the flame velocity in a node was predicted using five different flame front shapes of fireball, prism, bubble, spherical jet, and plane jet. For validation of the proposed model, the results of the Battelle multi-compartment hydrogen combustion test were used. The selected test cases for the study were Hx-6, 13, 14, 20 and Ix-2 which had two, three or four compartments under homogeneous hydrogen concentration of 5 to 10 vol%. The proposed model could predict well the combustion behavior in multi-compartment containment geometry on the whole. MELCOR code, incorporating the present combustion model, can simulate combustion behavior during severe accident with acceptable computing time and some degree of accuracy. The applicability study of the improved MELCOR code to the actual reactor plants will be further continued. (author)

  12. 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

  13. 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

  14. 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)

  15. MELCOR computer code manuals: Primer and user's guides, Version 1.8.3 September 1994. Volume 1

    International Nuclear Information System (INIS)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users' Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package

  16. Simulation of the PHEBUS FPT-1 experiment using MELCOR and exploration of the primary core degradation mechanism

    International Nuclear Information System (INIS)

    Wang, Jun; Corradini, Michael L.; Fu, Wen; Haskin, Troy; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-01-01

    Highlights: • Core degradation evaluation is an important process in risk analysis. • PHEBUS experiment was simulated using MELCOR. • The results confirm the validity of MELCOR’s simulation of the PHEBUS experiment. • These results are used to analyze the mode and behavior of core degradation. - Abstract: Core degradation evaluation of probability, progression and consequences of a core degradation accident is critical for evaluation of risk as well as its mitigation. However, research and modeling of severe accidents to date are limited, and their accuracy in predicting severe accident consequences is still insufficient. It is therefore important to explore the mechanisms of core degradation and to develop mitigation measures for severe accidents. PHEBUS FPT1 is a typical and classic core degradation experiment. MELCOR is a world famous severe accident analysis code developed by Sandia National Lab that has seen wide application, a broad user base, and a number of supporting experiments. The PHEBUS experiment was simulated using MELCOR in this paper. Experimental data on, thermal power and steam mass flow rates are used to determine average pressure, energy distribution, molten mass, temperature of the fuel, and hydrogen generation. Data from the PHEBUS experiment and Cho’s calculations are used to compare the average pressure, several fuel temperatures and the hydrogen generation rate. The results confirm the validity of MELCOR’s simulation of the PHEBUS experiment. The temperature distribution of the core is provided. These results are used to determine the mode and behavior of core degradation with the intent of building a foundation for further research

  17. Analyses of SBO sequence of VVER1000 reactor using TRACE and MELCOR codes

    International Nuclear Information System (INIS)

    Mazzini, Guido; Kyncl, Milos; Miglierini, Bruno; Kopecek, Vit

    2015-01-01

    In response to the Fukushima accident, the European Commission ordered to perform stress tests to all European Nuclear Power Plants (NPPs). Due to shortage of time a number of conclusions in national stress tests reports were based on engineering judgment only. In the Czech Republic, as a follow up, a consortium of Research Organizations and Universities has decided to simulate selected stress tests scenarios, in particular station Black-Out (SBO) and Loss of Ultimate Sink (LoUS), with the aim to verify conclusions made in the national stress report and to analyse time response of respective source term releases. These activities are carried out in the frame of the project 'Prevention, preparedness and mitigation of consequences of Severe Accident (SA) at Czech NPPs in relation to lessons learned from stress tests after Fukushima' financed by the Ministry of Interior. The Research Centre Rez has been working on the preparation of a MELCOR model for VVER1000 NPP starting with a plant systems nodalization. The basic idea of this paper is to benchmark the MELCOR model with the validated TRACE model, first comparing the steady state and continuing in a long term SBO plus another event until the beginning of the severe accident. The presented work focuses mainly on the preliminary comparison of the thermo-hydraulics of the two models created in MELCOR and TRACE codes. After that, preliminary general results of the SA progression showing the hydrogen production and the relocation phenomena will be shortly discussed. This scenario is considered closed after some seconds to the break of the lower head. (author)

  18. 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

  19. 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)

  20. 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

  1. 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.

  2. 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

  3. 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

  4. 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)

  5. 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.)

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. 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.

  11. Air ingression calculations for selected plant transients using MELCOR

    International Nuclear Information System (INIS)

    Kmetyk, L.N.

    1994-01-01

    Two sets of MELCOR calculations have been completed studying the effects of air ingression on the consequences of various severe accident scenarios. One set of calculations analyzed a station blackout with surge line failure prior to vessel breach, starting from nominal operating conditions; the other set of calculations analyzed a station blackout occurring during shutdown (refueling) conditions. Both sets of analyses were for the Surry plant, a three-loop Westinghouse PWR. For both accident scenarios, a basecase calculation was done, and then repeated with air ingression from containment into the core region following core degradation and vessel failure. In addition to the two sets of analyses done for this program, a similar air-ingression sensitivity study was done as part of a low-power/shutdown PRA, with results summarized here; that PRA study also analyzed a station blackout occurring during shutdown (refueling) conditions, but for the Grand Gulf plant, a BWR/6 with Mark III containment. These studies help quantify the amount of air that would have to enter the core region to have a significant impact on the severe accident scenario, and demonstrate that one effect, of air ingression is substantial enhancement of ruthenium release. These calculations also show that, while the core clad temperatures rise more quickly due to oxidation with air rather than steam, the core also degrades and relocates more quickly, so that no sustained, enhanced core heatup is predicted to occur with air ingression

  12. 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

  13. 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.)

  14. 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)

  15. 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)

  16. 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.)

  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. 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.

  19. 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

  20. 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)

  1. 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

  2. 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

  3. 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

  4. 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

  5. MELCOR/CONTAIN LMR Implementation Report - FY16 Progress.

    Energy Technology Data Exchange (ETDEWEB)

    Louie, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Humphries, Larry L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-11-01

    This report describes the progress of the CONTAIN - LMR sodium physics and chemistry models to be implemented in MELCOR 2.1. In the past three years , the implementation included the addition of sodium equations of state and sodium properties from two different sources. The first source is based on the previous work done by Idaho National Laboratory by modifying MELCOR to include liquid lithium equation of state as a working fluid to model the nuclear fusion safety research. The second source uses properties generated for the SIMMER code. The implemented modeling has been tested and results are reported in this document. In addition, the CONTAIN - LMR code was derived from an early version of the CONTAIN code, and many physical models that were developed since this early version of CONTAIN are not available in this early code version. Therefore, CONTAIN 2 has been updated with the sodium models in CONTAIN - LMR as CONTAIN2 - LMR, which may be used to provide code-to-code comparison with CONTAIN - LMR and MELCOR when the sodium chemistry models from CONTAIN - LMR have been completed. Both the spray fire and pool fire chemistry routines from CONTAIN - LMR have been integrated into MELCOR 2.1, and debugging and testing are in progress. Because MELCOR only models the equation of state for liquid and gas phases of the coolant, a modeling gap still exists when dealing with experiments or accident conditions that take place when the ambient temperature is below the freezing point of sodium. An alternative method is under investigation to overcome this gap . We are no longer working on the separate branch from the main branch of MELCOR 2.1 since the major modeling of MELCOR 2.1 has been completed. At the current stage, the newly implemented sodium chemistry models will be a part of the main MELCOR release version (MELCOR 2.2). This report will discuss the accomplishments and issues relating to the implementation. Also, we will report on the planned completion of all

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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)

  17. MELCOR 1.8.2 assessment: Surry PWR TMLB' (with a DCH study)

    International Nuclear Information System (INIS)

    Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.

    1994-02-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified

  18. 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

  19. 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)

  20. 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.

  1. 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

  2. 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

  3. PACTEL ISP-33. MELCOR assessment

    International Nuclear Information System (INIS)

    Siccama, N.B.

    1995-09-01

    The OECD/CSNI International Standard Problem (ISP-33) experiment was a natural-circulation experiment with a stepwide reduced primary coolant inventory in the PACTEL facility. The MELCOR code has been used to simulate this experiment. The main goal of these post-test calculations was to assess MELCOR on one- and two-phase natural-circulation phenomena which occur in Eastern European VVER plants in case of LOCA conditions. A base case and several senstivity calculations have been performed. In addition, the MELCOR results have been compared to results obtained by the RELAP5 code. Different natural-circulation modes have been identified during the experiment and simulated with MELCOR in the analyses of the ISP-33 experiment. These are successively: The single-phase liquid flow, the transient two-phase flow, the steady two-phase flow, and the boiler-condenser heat removal. These regimes, except the transient two-phase flow, are calculated in good agreement with the experiment. Special attention has been paid to the modeling of the two-phase flow in the hot legs of the PACTEL facility. Sensitivity calculation have shown that the results to a large extent are influenced by the nodalization of the hot legs and the opening heights of the hot-leg flow paths. Other senstivity calculations have shown that the time step and the core model do not influence the results, and accurate values for form loss coefficients and properties of the insulation are not necessary. The integrated MELCOR code is not inferior to the mechanistic RELAP5 code for the PACTEL ISP-33 post-test calculations. Some phenomena are modeled even better by MELCOR, because of the ability fit MELCOR parameters. (orig.)

  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. 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

  7. MELCOR 1.8.1 Assessment: LOFT integral experiment LP-FP-2

    International Nuclear Information System (INIS)

    Kmetyk, L.N.

    1992-12-01

    The MELCOR code has been used to model experiment LP-FP-2, an important source of integral data for qualifying severe accident code predictive capabilities. This assessment analysis clearly demonstrates MELCOR's ability to fulfill a large part of its primary, intended use, the calculation of severe accidents from full-power steady-state initiation through primary-system thermal/hydraulic response and core damage to fission product release, transport and deposition. After a number of code errors were identified and corrected, few nonstandard inputs and no code problem-specific modifications were needed to provide reasonable agreement with test data in all areas considered. Code-to-code comparisons show that MELCOR does at least as well as other ''best-estimate'' (i.e., SCDAP/RELAP5) or integral (i.e., MAAP) codes in predicting the thermal/hydraulic and core responses in this large-scale, integral experiment; in fact, MELCOR and MAAP appear to give the best agreement with data, especially for clad temperature histories. Further, our code-to-code comparisons indicate that MELCOR does at least as well as ''best-estimate'' fission product codes in predicting the source term, with a number of such codes having to be run in tandem and driven by test data or other ''best-estimate'' thermal/hydraulic and core damage codes to provide results equivalent to a single, integrated MELCOR calculation

  8. Development of a MELCOR self-initialization algorithm for boiling water reactors

    International Nuclear Information System (INIS)

    Chien, C.S.; Wang, S.J.; Cheng, S.K.

    1996-01-01

    The MELCOR code, developed by Sandia National Laboratories, is suitable for calculating source terms and simulating severe accident phenomena of nuclear power plants. Prior to simulating a severe accident transient with MELCOR, the initial steady-state conditions must be generated in advance. The current MELCOR users' manuals do not provide a self-initialization procedure; this is the reason users have to adjust the initial conditions by themselves through a trial-and-error approach. A MELCOR self-initialization algorithm for boiling water reactor plants has been developed, which eliminates the tedious trial-and-error procedures and improves the simulation accuracy. This algorithm adjusts the important plant variable such as the dome pressure, downcomer level, and core flow rate to the desired conditions automatically. It is implemented through input with control functions provided in MELCOR. The reactor power and feedwater temperature are fed as input data. The initialization work of full-power conditions of the Kuosheng nuclear power station is cited as an example. These initial conditions are generated successfully with the developed algorithm. The generated initial conditions can be stored in a restart file and used for transient analysis. The methodology in this study improves the accuracy and consistency of transient calculations. Meanwhile, the algorithm provides all MELCOR users an easy and correct method for establishing the initial conditions

  9. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez, C.

    2013-07-01

    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

  10. Study of steam condensation at sub-atmospheric pressure: setting a basic research using MELCOR code

    Science.gov (United States)

    Manfredini, A.; Mazzini, M.

    2017-11-01

    One of the most serious accidents that can occur in the experimental nuclear fusion reactor ITER is the break of one of the headers of the refrigeration system of the first wall of the Tokamak. This results in water-steam mixture discharge in vacuum vessel (VV), with consequent pressurization of this container. To prevent the pressure in the VV exceeds 150 KPa absolute, a system discharges the steam inside a suppression pool, at an absolute pressure of 4.2 kPa. The computer codes used to analyze such incident (eg. RELAP 5 or MELCOR) are not validated experimentally for such conditions. Therefore, we planned a basic research, in order to have experimental data useful to validate the heat transfer correlations used in these codes. After a thorough literature search on this topic, ACTA, in collaboration with the staff of ITER, defined the experimental matrix and performed the design of the experimental apparatus. For the thermal-hydraulic design of the experiments, we executed a series of calculations by MELCOR. This code, however, was used in an unconventional mode, with the development of models suited respectively to low and high steam flow-rate tests. The article concludes with a discussion of the placement of experimental data within the map featuring the phenomenon characteristics, showing the importance of the new knowledge acquired, particularly in the case of chugging.

  11. 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

  12. 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.

  13. 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)

  14. MELCOR/VISOR PWR desktop simulator

    International Nuclear Information System (INIS)

    With, Anka de; Wakker, Pieter

    2010-01-01

    Increasingly, there is a need for a learning support and training tool for nuclear engineers, utilities and students in order to broaden their understanding of advanced nuclear plant characteristics, dynamics, transients and safety features. Nuclear system analysis codes like ASTEC, RELAP5, RETRAN and MELCOR provide calculation results of and visualization tools can be used to graphically represent these results. However, for an efficient education and training a more interactive tool such as a simulator is needed. The simulator connects the graphical tool with the calculation tool in an interactive manner. A small number of desktop simulators exist [1-3]. The existing simulators are capable of representing different types of power plants and various accident conditions. However, they were found to be too general to be used as a reliable plant-specific accident analysis or training tool. A desktop simulator of the Pressurized Water Reactor (PWR) has been created under contract of the Dutch nuclear regulatory body (KFD). The desktop simulator is a software package that provides a close to real simulation of the Dutch nuclear power plant Borssele (KCB) and is used for training of the accident response. The simulator includes the majority of the power plant systems, necessary for the successful simulation of the KCB plant during normal operation, malfunctions and accident situations, and it has been successfully validated against the results of the safety evaluations from the KCB safety report. (orig.)

  15. 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

  16. 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

  17. 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)

  18. MELCOR 1.8.2 assessment: The DF-4 BWR Damaged Fuel experiment

    International Nuclear Information System (INIS)

    Tautges, T.J.

    1993-10-01

    MELCOR is a fully integrated, engineering-level computer code being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As a part of an ongoing assessment, program, MELCOR has been used to model the ACRR in-pile DF-4 Damaged Fuel experiment. DF-4 provided data for early phase melt progression in BWR fuel assemblies, particularly for phenomena associated with eutectic interactions in the BWR control blade and zircaloy oxidation in the canister and cladding. MELCOR provided good agreement with experimental data in the key areas of eutectic material behavior and canister and cladding oxidation. Several shortcomings associated with the MELCOR modeling of BWR geometries were found and corrected. Twenty-five sensitivity studies were performed on COR, HS and CVH parameters. These studies showed that the new MELCOR eutectics model played an important role in predicting control blade behavior. These studies revealed slight time step dependence and no machine dependencies. Comparisons made with the results from four best-estimate codes showed that MELCOR did as well as these codes in matching DF-4 experimental data

  19. MELCOR 1.8.2 Assessment: IET direct containment heating tests

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.

    1993-10-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRS. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze several of the IET direct containment heating experiments done at 1:10 linear scale in the Surtsey test facility at Sandia and at 1:40 linear scale in the corium-water thermal interactions (CWTI) COREXIT test facility at Argonne National Laboratory. These MELCOR calculations were done as an open post-test study, with both the experimental data and CONTAIN results available to guide the selection of code input. Basecase MELCOR results are compared to test data in order to evaluate the new HPME DCH model recently added in MELCOR version 1.8.2. The effect of various user-input parameters in the HPME model, which define both the initial debris source and the subsequent debris interaction, were investigated in sensitivity studies. In addition, several other non-default input modelling changes involving other MELCOR code packages were required in our IET assessment analyses in order to reproduce the observed experiment behavior. Several calculations were done to identify whether any numeric effects exist in our DCH IET assessment analyses.

  20. 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

  1. 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

  2. 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

  3. 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.

  4. 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)

  5. Assessment of CONTAIN and MELCOR for performing LOCA and LOVA analyses in ITER

    International Nuclear Information System (INIS)

    Merrill, B.J.; Hagrman, D.L.; Gaeta, M.J.; Petti, D.A.

    1994-09-01

    This report describes the results of an assessment of the CONTAIN and MELCOR computer codes for ITER LOCA and LOVA applications. As part of the assessment, the results of running a test problem that describes an ITER LOCA are presented. It is concluded that the MELCOR code should be the preferred code for ITER severe accident thermal hydraulic analyses. This code will require the least modification to be appropriate for calculating thermal hydraulic behavior in ITER relevant conditions that include vacuum, cryogenics, ITER temperatures, and the presence of a liquid metal test module. The assessment of the aerosol transport models in these codes concludes that several modifications would have to be made to CONTAIN and/or MELCOR to make them applicable to the aerosol transport part of severe accident analysis in ITER

  6. 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.)

  7. 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.)

  8. 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)

  9. 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.)

  10. 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.)

  11. 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.

  12. Application of the MELCOR code to design basis PWR large dry containment analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, Jesse; Notafrancesco, Allen (USNRC, Office of Nuclear Regulatory Research, Rockville, MD); Tills, Jack Lee (Jack Tills & Associates, Inc., Sandia Park, NM)

    2009-05-01

    The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

  13. 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

  14. 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.)

  15. 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

  16. 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.)

  17. 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...

  18. 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

  19. 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.

  20. 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

  1. 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

  2. MELCOR 1.8.3 assessment: CSE containment spray experiments

    International Nuclear Information System (INIS)

    Kmetyk, L.N.

    1994-12-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRS. As part, of an ongoing assessment program, the MELCOR computer code has been used to analyze a series of containment spray tests performed in the Containment Systems Experiment (CSE) vessel to evaluate the performance of aqueous sprays as a means of decontaminating containment atmospheres. Basecase MELCOR results are compared with test data, and a number of sensitivity studies on input modelling parameters and options in both the spray package and the associated aerosol washout and atmosphere decontamination by sprays modelled in the radionuclide package have been done. Time-step and machine-dependency calculations were done to identify whether any numeric effects exist in these CSE assessment analyses. A significant time-step dependency due to an error in the spray package coding was identified and eliminated. A number of other code deficiencies and inconveniences also are noted

  3. 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)

  4. 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

  5. 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

  6. PHEBUS FPT-1 simulation by using MELCOR and primary blockage model exploration

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jun [Institite of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Wang, Chen [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Department of Engineering Physics, Tsinghua University, Beijing 100084 (China); Corradini, Michael L.; Haskin, Troy [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Tian, Wenxi; Su, Guanghui [Institite of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [Institite of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China)

    2016-10-15

    Highlights: • Flow channel blockage model is expected to be the key parameter for hydrogen generation calculation. • Flow channel blockage situation is studied in this work. • MELCOR is used as the tool, and PHEBUS FPT1 is used as benchmark. • Model sensitivity analysis on hydrogen generation will be done in next step. - Abstract: Recently, MAAP and MELCOR research teams completed a set of accident simulations to reconstruct the Fukushima-Daiichi accident in order to better understand severe accident progression. One result from this work is that the predicted hydrogen generation in MELCOR is notably more than that in MAAP. The fuel rod degradation process (i.e., debris formation and blockage models) may be responsible for this difference and opportunity exists to understand the key reasons for the difference. To examine this hypothesis, in this paper, the PHEBUS FPT1 experiment is selected as a benchmark test and MELCOR is used as the analysis tool. MELCOR calculation results are compared with PHEBUS FPT1 data to verify our model. Based on the validation of a nominal MELCOR simulation of the FPT1 test, we use the volume fractions of each component to visualize the debris-blockage geometric arrangement for PHEBUS FPT1 as the fuel degradation event proceeds. Cloud figures for the volume fractions of each component such as flow volume fraction, cladding volume fraction, fuel rod volume fraction, supporting material volume fraction, non-supporting material volume fraction and debris bed porosity fraction are shown in this paper. The results provide us with a visualized approach for improving our understanding of core degradation.

  7. PHEBUS FPT-1 simulation by using MELCOR and primary blockage model exploration

    International Nuclear Information System (INIS)

    Wang, Jun; Wang, Chen; Corradini, Michael L.; Haskin, Troy; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2016-01-01

    Highlights: • Flow channel blockage model is expected to be the key parameter for hydrogen generation calculation. • Flow channel blockage situation is studied in this work. • MELCOR is used as the tool, and PHEBUS FPT1 is used as benchmark. • Model sensitivity analysis on hydrogen generation will be done in next step. - Abstract: Recently, MAAP and MELCOR research teams completed a set of accident simulations to reconstruct the Fukushima-Daiichi accident in order to better understand severe accident progression. One result from this work is that the predicted hydrogen generation in MELCOR is notably more than that in MAAP. The fuel rod degradation process (i.e., debris formation and blockage models) may be responsible for this difference and opportunity exists to understand the key reasons for the difference. To examine this hypothesis, in this paper, the PHEBUS FPT1 experiment is selected as a benchmark test and MELCOR is used as the analysis tool. MELCOR calculation results are compared with PHEBUS FPT1 data to verify our model. Based on the validation of a nominal MELCOR simulation of the FPT1 test, we use the volume fractions of each component to visualize the debris-blockage geometric arrangement for PHEBUS FPT1 as the fuel degradation event proceeds. Cloud figures for the volume fractions of each component such as flow volume fraction, cladding volume fraction, fuel rod volume fraction, supporting material volume fraction, non-supporting material volume fraction and debris bed porosity fraction are shown in this paper. The results provide us with a visualized approach for improving our understanding of core degradation.

  8. 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

  9. 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

  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. 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

  15. 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

  16. 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)

  17. 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.

  18. 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

  19. 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

  20. 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

  1. ASTEC and MELCOR comparison for a VVER-1000 60 mm small break LOCA

    International Nuclear Information System (INIS)

    Georgieva, J.; Stefanova, A.; Groudev, P.; Tusheva, P.; Mladenov, I.; Dimov, D.; Passalacqua, R.

    2005-01-01

    In this paper a comparison between severe accident calculations performed for a WWER 1000 with the ASTEC1.1v0 and MELCOR 1.8.5 computer codes for a small break LOCA (ID 60 mm) without intervention of hydro accumulators is presented. This investigation has been performed in the framework of the SARNET project under the EURATOM 6th framework program. Once the accident sequence scenario is specified, both codes (MELCORE and ASTEC) are able to determine the core and containment damaged states, to estimate the release of radionuclides from the fuel as well as from the primary circuit and containment. Theses results are used to estimate the maximum period of the time during which the personnel could still take particular decisions in order to mitigate such an accident. The aim of the performed analysis is to estimate the discrepancy between ASTEC and MELCORE 1.8.5 calculations. Such discrepancies will be studied, if the case, proposal for ASTEC improvements will be made. Also the ASTEC capability to simulate specific reactor accident scenarios and/or particular safety systems will be tested. The final target is to propose severe accident management procedure for WWER 1000 reactors. In conclusions, the analysis for a small break LOCA (ID 60 mm without hydroelectricities) has shown some discrepancies between ASTEC and MELCORE especially during the degradation of the core. Further analyses are planed in which the MELCORE temperature 'set point' for core degradation (2520 K) will be progressively increased to approach the ASTEC one (which has been estimated to be about 3200 K). The comparison of the new results will allow a better evaluation of the in-vessel models implemented in ASTEC

  2. The establishment of MELCOR/SNAP model of Chinshan nuclear power plant for Ultimate Response Guideline

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Wen-Sheng, E-mail: wshsu@ess.nthu.edu.tw [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); Chiang, Yu, E-mail: s101013702@m101.nthu.edu.tw [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); Wang, Jong-Rong, E-mail: jongrongwang@gmail.com [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); Wang, Ting-Yi, E-mail: minired1119@gmail.com [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); Wang, Te-Chuan, E-mail: tcwang@iner.gov.tw [Institute of Nuclear Energy Research Atomic Energy Council, R.O.C., 1000, Wenhua Road Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Teng, Jyh-Tong, E-mail: jyhtong@cycu.edu.tw [Department of Mechanical Engineering, Chung Yuan Christian University, 200, Chung Pei Rd, Chung Li 32023, Taiwan, ROC (China); Chen, Shao-Wen, E-mail: chensw@mx.nthu.edu.tw [Nuclear Science and Technology Development Center, Institute of Nuclear Engineering and Science, National Tsing Hua University, Nuclear and New Energy Education and Research Foundation, No. 101, Section 2, Kuang Fu Rd., HsinChu 30013, Taiwan, ROC (China); and others

    2017-01-15

    Highlights: • The establishment of a MELCOR/SNAP model of Chinshan (BWR/4). • MELCOR/SNAP model was used to estimate the effectiveness of URG for Chinshan. • The MELCOR results were compared to MAAP, TRACE and PCTRAN. • URG is a new method to prevent a Fukushima-like accident. • The low raw water (150 GPM) can make the cladding temperature below 1088.7 K. - Abstract: After Fukushima Daiichi disaster, the safety analysis of severe accidents became one of the safety concerns in Taiwan. The Emergency Operating Procedure (EOP) cannot cope with a multiple system failure situation under a severe accident since it is a “Symptom-basis” procedure. To deal with that, Taiwan Power Company built up a new strategy for Fukushima-like accident called Ultimate Response Guideline (URG). It is a simple strategy with three main conditions: loss of regular motor driven injection system, loss of all AC power and tsunami/earthquake warning. If two of three happen, the operating procedure will change from EOP to URG and start the main works by following the strategy. There are three main works in URG: controlled-depressurization, line up low pressure injection water and prepare containment venting. In this study, MELCOR2.1 was used to calculate the cases of URG and checked the goal of the strategy that prevents the accident or not. There were three steps in this research. First, a model of Chinshan nuclear power plant (NPP) was built. Second, one was the case with URG and the other was not by using the above MELCOR model. The results were compared to MAAP5.0, TRACE and PCTRAN. Finally, some sensitivity studies of depressurization and water injection rate were done.

  3. 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.)

  4. MELCOR ex-vessel LOCA simulations for ITER+

    International Nuclear Information System (INIS)

    Gaeta, M.J.; Merrill, B.J.; Bartels, H.W.

    1995-01-01

    Ex-vessel Loss-of-Coolant-Accident (LOCA) simulations for the International Thermonuclear Experimental Reactor (ITER) were performed using the MELCOR code. The main goals of this work were to estimate the ultimate pressurization of the heat transport system (HTS) vault in order to gauge the potential for stack releases and to estimate the total amount of hydrogen generated during a design basis ex-vessel LOCA. Simulation results indicated that the amount of hydrogen produced in each transient was below the flammability limit for the plasma chamber. In addition, only moderate pressurization of the HTS vault indicated a very small potential for releases through the stack

  5. 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.

  6. 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.

  7. Calculations of core concrete interaction using MELCOR 1.8.5

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hwan Yeol; Song, Jin Ho; Kim, Hee Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    OECD/MCCI project is scheduled for 4 years from 2002. 1 to 2005. 12 to perform a series of tests through which the data for cooling the molten core spread out at the reactor cavity and for the long-term CCI (Core Concrete Interaction) are secured. This paper deals with the transient calculations of the 2-D CCI tests performed under the OECD/MCCI project by using a well-known severe accident analysis code, MELCOR 1.8.5. The CCI test was performed at the rectangular geometry with one ablative bottom wall and two ablative and two non-ablative side walls. Since the MELCOR 1.8.5 can only accommodate a cylindrical geometry, an appropriate scaling methodology was applied to adjust the geometrical difference between the CCI test and the MELCOR calculations. The default heat transfer models contained in the CORCON-Mod3 module of MELCOR 1.8.5 were used for the base case calculation. The key parameters of the CCI phenomena such as the melt temperature, concrete ablation, cavity shape, gas generation, heat transfer rate, etc. were calculated and compared with the test results. In addition, sensitivity studies with the change of the inputs and character variables of MELCOR were also included.

  8. MELCOR simulation of steam condensation effect on hydrogen behavior in THAI HM-2 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seongnyeon; Lee, Jung-Jae; Cho, Yong-Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

    2015-10-15

    In this study, MELCOR simulation was carried out for THAI HM-2 experiment of OECD. As a results, stratification of hydrogen cloud was reasonably captured in MELCOR simulation. Furthermore, the pressure from simulation results in cases where mass transfer coefficient of MELCOR condensation model was modified was good agreement with the experimental results. Containment Filtered Ventilation System (CFVS) has been introduced as facility to prevent containment failure during severe accident. However, possibility of hydrogen risk has been issued due to inflow of hydrogen, condensation and removal of steam and complicated inner structure in CFVS. Preferentially benchmark work for THAI HM-2 experiment of OECD was decided to validate the methodology before detailed assessment of hydrogen risk in CFVS. The objectives of THAI HM-2 experiment were evaluation of hydrogen behavior, verification of numerical analysis tools and so on. In this paper, therefore, MELCOR simulation was carried out in comparison with the experiment results. Additionally, steam condensation effect was considered for detailed simulation. Hydrogen concentration from MELCOR results was underestimated in comparison to the experimental results.

  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. Development of an Input Model to MELCOR 1.8.5 for the Ringhals 3 PWR

    International Nuclear Information System (INIS)

    Nilsson, Lars

    2004-12-01

    An input file to the severe accident code MELCOR 1.8.5 has been developed for the Swedish pressurized water reactor Ringhals 3. The aim was to produce a file that can be used for calculations of various postulated severe accident scenarios, although the first application is specifically on cases involving large hydrogen production. The input file is rather detailed with individual modelling of all three cooling loops. The report describes the basis for the Ringhals 3 model and the input preparation step by step and is illustrated by nodalization schemes of the different plant systems. Present version of the report is restricted to the fundamental MELCOR input preparation, and therefore most of the figures of Ringhals 3 measurements and operating parameters are excluded here. These are given in another, complete version of the report, for limited distribution, which includes tables for pertinent data of all components. That version contains appendices with a complete listing of the input files as well as tables of data compiled from a RELAP5 file, that was a major basis for the MELCOR input for the cooling loops. The input was tested in steady-state calculations in order to simulate the initial conditions at current nominal operating conditions in Ringhals 3 for 2775 MW thermal power. The results of the steady-state calculations are presented in the report. Calculations with the MELCOR model will then be carried out of certain accident sequences for comparison with results from earlier MAAP4 calculations. That work will be reported separately

  11. 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

  12. 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

  13. 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)

  14. 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)

  15. 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

  16. 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.)

  17. 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

  18. Evaluation of VVER-1200/V-491 reactor pressure vessel integrity during large break LOCA along with SBO using MELCOR 1.8.6

    International Nuclear Information System (INIS)

    Bui Thi Hoa; Tran Chi Thanh

    2015-01-01

    After Fukushima accident and stress test recommended by IAEA for existing reactors, higher safety requirements are enforced upon nuclear power plants during design extension and severe accident conditions. Based on those arguments, Vietnam Government requests a lot of effective safety solutions, in designs proposed for the nuclear power plants in Ninh Thuan province of Vietnam, which can prevent the accident progression toward severe accidents and mitigate severe accident consequences. One of safety requirements is related to delay time of core melt during design extension condition. Especially, if the worst case of accidents occurs, the reactor vessel integrity must be maintained at least 24 hours from the beginning of the accident. With the aim at investigation of Reactor Pressure Vessel (RPV) integrity, in this study, MELCOR 1.8.6 code is used to evaluate the integrity of RPV lower head for VVER-1200/V-491 reactor during a Large Break Loss of Coolant Accident (LBLOCA) in combination with Station Blackout (SBO) event. The study figures out several parameters related to melt down progress such as: rupture position and rupture timing, the amount of hydrogen generated. Availability of the second stage hydro-accumulators (HA2) in the VVER-1200/V-491 is assumed as an additional improvement to delay the timing of core melt as well as to maintain the vessel integrity for long-term. (author)

  19. 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

  20. 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.

  1. 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

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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.

  11. [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.

  12. 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

  13. 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.

  14. 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

  15. 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

  16. 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

  17. Effect of In-Vessel Retention Strategies under Postulated SGTR Accidents of OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Wonjun; Lee, Yongjae; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of); Kim, Hwan-Yeol; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, MELCOR code was used to simulate the severe accident of the OPR1000. MELCOR code is computer code which enables to simulate the progression of the severe accident for light water reactors. It has been developed by Sandia National Laboratories for plant risk assessment and source term analysis since 1982. According to the probabilistic safety analysis (PSA) Level 1 of OPR1000, typical severe accident scenarios of high probability of a transition to severe accident for OPR1000 were identified as Small Break Loss of Coolant Accident (SBLOCA), Station Black out (SBO), Total Loss of Feed Water (TLOFW), and Steam Generator Tube Rupture. While the first three accidents are expected to result in the generation and transportation of the radioactive nuclides within the containment building as consequence of the core damage and subsequent reactor pressure vessel (RPV) failure, the latter accident scenario may be progressed with possible direct release of the radioactive nuclides to the environment by bypassing the containment building. Thus it is of significance to investigate the SGTR accident with a sophisticated severe accident code. This code can simulate the whole phenomena of a severe accident such as thermal-hydraulic response, core heat-up, oxidation and relocation, and fission product release and transport. Thus many researchers have used MELCOR in severe accident studies. In this study, in-vessel retention strategies were applied for postulated SGTR accidents. Mitigation effect and adverse effect of in-vessel strategies was studied in aspect of RPV failure, fission product release and containment thermal-hydraulic and hydrogen behavior. Base case of SGTR accident and three mitigation cases were simulated using MELCOR code 1.8.6. For each mitigation cases, mitigation effect and adverse effect were investigated. Conclusions can be summarized as follows: (1) RPV failure of SGTR base case occurred at 5.62 hours and fission product of RCS released to

  18. Application of FFTBM with signal mirroring to improve accuracy assessment of MELCOR code

    International Nuclear Information System (INIS)

    Saghafi, Mahdi; Ghofrani, Mohammad Bagher; D’Auria, Francesco

    2016-01-01

    Highlights: • FFTBM-SM is an improved Fast Fourier Transform Base Method by signal mirroring. • FFTBM-SM has been applied to accuracy assessment of MELCOR code predictions. • The case studied was Station Black-Out accident in PSB-VVER integral test facility. • FFTBM-SM eliminates fluctuations of accuracy indices when signals sharply change. • Accuracy assessment is performed in a more realistic and consistent way by FFTBM-SM. - Abstract: This paper deals with the application of Fast Fourier Transform Base Method (FFTBM) with signal mirroring (FFTBM-SM) to assess accuracy of MELCOR code. This provides deeper insights into how the accuracy of MELCOR code in predictions of thermal-hydraulic parameters varies during transients. The case studied was modeling of Station Black-Out (SBO) accident in PSB-VVER integral test facility by MELCOR code. The accuracy of this thermal-hydraulic modeling was previously quantified using original FFTBM in a few number of time-intervals, based on phenomenological windows of SBO accident. Accuracy indices calculated by original FFTBM in a series of time-intervals unreasonably fluctuate when the investigated signals sharply increase or decrease. In the current study, accuracy of MELCOR code is quantified using FFTBM-SM in a series of increasing time-intervals, and the results are compared to those with original FFTBM. Also, differences between the accuracy indices of original FFTBM and FFTBM-SM are investigated and correction factors calculated to eliminate unphysical effects in original FFTBM. The main findings are: (1) replacing limited number of phenomena-based time-intervals by a series of increasing time-intervals provides deeper insights about accuracy variation of the MELCOR calculations, and (2) application of FFTBM-SM for accuracy evaluation of the MELCOR predictions, provides more reliable results than original FFTBM by eliminating the fluctuations of accuracy indices when experimental signals sharply increase or

  19. Application of FFTBM with signal mirroring to improve accuracy assessment of MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

    Highlights: • FFTBM-SM is an improved Fast Fourier Transform Base Method by signal mirroring. • FFTBM-SM has been applied to accuracy assessment of MELCOR code predictions. • The case studied was Station Black-Out accident in PSB-VVER integral test facility. • FFTBM-SM eliminates fluctuations of accuracy indices when signals sharply change. • Accuracy assessment is performed in a more realistic and consistent way by FFTBM-SM. - Abstract: This paper deals with the application of Fast Fourier Transform Base Method (FFTBM) with signal mirroring (FFTBM-SM) to assess accuracy of MELCOR code. This provides deeper insights into how the accuracy of MELCOR code in predictions of thermal-hydraulic parameters varies during transients. The case studied was modeling of Station Black-Out (SBO) accident in PSB-VVER integral test facility by MELCOR code. The accuracy of this thermal-hydraulic modeling was previously quantified using original FFTBM in a few number of time-intervals, based on phenomenological windows of SBO accident. Accuracy indices calculated by original FFTBM in a series of time-intervals unreasonably fluctuate when the investigated signals sharply increase or decrease. In the current study, accuracy of MELCOR code is quantified using FFTBM-SM in a series of increasing time-intervals, and the results are compared to those with original FFTBM. Also, differences between the accuracy indices of original FFTBM and FFTBM-SM are investigated and correction factors calculated to eliminate unphysical effects in original FFTBM. The main findings are: (1) replacing limited number of phenomena-based time-intervals by a series of increasing time-intervals provides deeper insights about accuracy variation of the MELCOR calculations, and (2) application of FFTBM-SM for accuracy evaluation of the MELCOR predictions, provides more reliable results than original FFTBM by eliminating the fluctuations of accuracy indices when experimental signals sharply increase or

  20. 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.

  1. 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

  2. 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)

  3. 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)

  4. 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

  5. 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.)

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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)

  14. 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

  15. 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

  16. 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.

  17. 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

  18. 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

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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)

  9. Application of Melcor code for the calculo of TMLB sequence in PWR with natural circulating into the vessel

    International Nuclear Information System (INIS)

    Marten-Fuertes, F.

    1995-01-01

    The use of computer codes to analyze the phenomena of severe accidents is very important to take decisions in Nuclear Safety. This paper presents the MELCOR code used to calculate the TMLB sequence of PWR with natural circulation into the vessels. The main goal of this code is its application for the PSA (probabilistic safety analysis)

  10. Fusion safety codes International modeling with MELCOR and ATHENA- INTRA

    CERN Document Server

    Marshall, T; Topilski, L; Merrill, B

    2002-01-01

    For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA-INTRA codes and presents their modeling results for the following breaches of a water cooling line into the...

  11. PWR hot leg natural circulation modeling with MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Lee, Jong In [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    1997-12-31

    Previous MELCOR and SCDAP/RELAP5 nodalizations for simulating the counter-current, natural circulation behavior of vapor flow within the RCS hot legs and SG U-tubes when core damage progress can not be applied to the steady state and water-filled conditions during the initial period of accident progression because of the artificially high loss coefficients in the hot legs and SG U-tubes which were chosen from results of COMMIX calculation and the Westinghouse natural circulation experiments in a 1/7-scale facility for simulating steam natural circulation behavior in the vessel and circulation modeling which can be used both for the liquid flow condition at steady state and for the vapor flow condition at the later period of in-vessel core damage. For this, the drag forces resulting from the momentum exchange effects between the two vapor streams in the hot leg was modeled as a pressure drop by pump model. This hot leg natural circulation modeling of MELCOR was able to reproduce similar mass flow rates with those predicted by previous models. 6 refs., 2 figs. (Author)

  12. PWR hot leg natural circulation modeling with MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Lee, Jong In [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    1998-12-31

    Previous MELCOR and SCDAP/RELAP5 nodalizations for simulating the counter-current, natural circulation behavior of vapor flow within the RCS hot legs and SG U-tubes when core damage progress can not be applied to the steady state and water-filled conditions during the initial period of accident progression because of the artificially high loss coefficients in the hot legs and SG U-tubes which were chosen from results of COMMIX calculation and the Westinghouse natural circulation experiments in a 1/7-scale facility for simulating steam natural circulation behavior in the vessel and circulation modeling which can be used both for the liquid flow condition at steady state and for the vapor flow condition at the later period of in-vessel core damage. For this, the drag forces resulting from the momentum exchange effects between the two vapor streams in the hot leg was modeled as a pressure drop by pump model. This hot leg natural circulation modeling of MELCOR was able to reproduce similar mass flow rates with those predicted by previous models. 6 refs., 2 figs. (Author)

  13. Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farmer, Mitchell [Argonne National Lab. (ANL), Argonne, IL (United States); Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-03-01

    System-level code analyses by both United States and international researchers predict major core melting, bottom head failure, and corium-concrete interaction for Fukushima Daiichi Unit 1 (1F1). Although system codes such as MELCOR and MAAP are capable of capturing a wide range of accident phenomena, they currently do not contain detailed models for evaluating some ex-vessel core melt behavior. However, specialized codes containing more detailed modeling are available for melt spreading such as MELTSPREAD as well as long-term molten corium-concrete interaction (MCCI) and debris coolability such as CORQUENCH. In a preceding study, Enhanced Ex-Vessel Analysis for Fukushima Daiichi Unit 1: Melt Spreading and Core-Concrete Interaction Analyses with MELTSPREAD and CORQUENCH, the MELTSPREAD-CORQUENCH codes predicted the 1F1 core melt readily cooled in contrast to predictions by MELCOR. The user community has taken notice and is in the process of updating their systems codes; specifically MAAP and MELCOR, to improve and reduce conservatism in their ex-vessel core melt models. This report investigates why the MELCOR v2.1 code, compared to the MELTSPREAD and CORQUENCH 3.03 codes, yield differing predictions of ex-vessel melt progression. To accomplish this, the differences in the treatment of the ex-vessel melt with respect to melt spreading and long-term coolability are examined. The differences in modeling approaches are summarized, and a comparison of example code predictions is provided.

  14. Identification of gap cooling phenomena from LAVA-4 experiment using MELCOR

    International Nuclear Information System (INIS)

    Park, Jong-Hwa; Kim, Dong-Ha; Kim, See-Darl; Kim, Sang-Baik; Kim, Hee-Dong

    2000-01-01

    During the severe accident, whether the hot debris in. lower head will be cool-down or not is the important issue concerning the plant safety. KAERI has launched the 'LAVA' experimental program to examine the existence of initial gap and its effect on the cooling of hot debris. The objective of this study is to identify the gap cooling phenomena from the analysis of simulation results on LAVA-4 experiment using MELCOR1.8.4 code. Three parameters on the debris coolability in MELCOR are the quenching heat transfer coefficient for the interaction between molten Al 2 O 3 and water, the heat transfer coefficient from debris to wall and the diameter of the particulate debris for calculating the available heat transfer area with water. The sensitivity study was performed with these three parameters. However it was believed that there must be a gap between debris and inside wall during the transient. MELCOR1.8.4 does not consider these gap-cooling phenomena. Therefore a conceptual gap-cooling model has been developed and implemented into the lower plenum model in MELCOR to take into account the gap effect in the lower plenum. When the 'gap model' is implemented, the peak temperature of the vessel wall was reduced and its cooling rate was increased. (author)

  15. A restructuring of the MELCOR fission product packages for the MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S.H.; Kim, K.R.; Kim, D.H.

    2004-01-01

    The RN1/RN2 packages, which are the fission product-related packages in MELCOR, have been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and a modernized data structure. To do this, the data transferring methods of the current MELCOR code are modified and adopted into the RN1/RN2 package. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to user-defined data type, which leads to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN1/RN2 package addressed in this paper includes a module development, subroutine modification, and the treatment of MELGEN, which generates the data file, as well as MELCOR, which is processing the calculation. The verification has been done by comparing the results of the modified code with those of the existing code. As the trends are similar to each other, it implies that the same approach could be extended to the entire code package. It is expected that the code restructuring will accelerate the code domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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

  1. The Phebus FP thermal-hydraulic analysis with Melcor

    International Nuclear Information System (INIS)

    Akgane, Kikuo; Kiso, Yoshihiro; Fukahori, Takanori; Yoshino, Mamoru

    1995-01-01

    The severe accident analysis code MELCOR, version 1.8.2, has been applied for thermal-hydraulic pre-test analysis of the first test of the Phebus FP program (test FPT-0) to study the best test parameters and the applicability of the code. The Phebus FP program is an in-pile test program which has been planned by the French Commissariate a L'Energie Atomique and the Commission of the European Union. The experiments are being conducted by an international collaboration to study the release and transport of fission products (FPs) under conditions assumed to be the most representative of those that would occur in a severe accident. The Phebus FP test apparatus simulates a test bundle of an in-pile section, the circuit including the steam generator U-tubes and the containment. The FPT-0 test was designed to simulate the heat-up and subsequent fuel bundle degradation after a loss of coolant severe accident, using fresh fuel. Two options for fuel degradation models in MELCOR have been applied to fuel degradation behavior. the first model assumes that fuel debris will be formed immediately after the fuel support fails by cladding relocation due to the candling process. The other is the uncollapsed bare fuel pellets option, in which the fuel pellets remain standing in a columnar shape until the fuel reaches its melting point, even if the cladding has been relocated by candling. The thermal-hydraulic behaviors in the circuit and containment of Phebus FP are discussed herein. Flow velocities in the Phebus FP circuit are high in order to produce turbulent flow in a small diameter test pipe. The MELCOR calculation has shown that the length of the hot leg and steam generator are adequate to attain steam temperatures or 700 degrees C and 150 degrees C in the respective outlets. The containment atmosphere temperature and humidity derived by once through integral system calculation show that objective test conditions would be satisfied in the Phebus FP experiment

  2. The Phebus FP thermal-hydraulic analysis with Melcor

    Energy Technology Data Exchange (ETDEWEB)

    Akgane, Kikuo; Kiso, Yoshihiro [Nuclear Power Engineering Corporation, Tokyo (Japan); Fukahori, Takanori [Hitachi Engineering Company, Ltd., Hitachi-shi Ibaraki-ken (Japan); Yoshino, Mamoru [Nuclear Engineering Ltd., Tosabori Nishi-ku (Japan)

    1995-09-01

    The severe accident analysis code MELCOR, version 1.8.2, has been applied for thermal-hydraulic pre-test analysis of the first test of the Phebus FP program (test FPT-0) to study the best test parameters and the applicability of the code. The Phebus FP program is an in-pile test program which has been planned by the French Commissariate a L`Energie Atomique and the Commission of the European Union. The experiments are being conducted by an international collaboration to study the release and transport of fission products (FPs) under conditions assumed to be the most representative of those that would occur in a severe accident. The Phebus FP test apparatus simulates a test bundle of an in-pile section, the circuit including the steam generator U-tubes and the containment. The FPT-0 test was designed to simulate the heat-up and subsequent fuel bundle degradation after a loss of coolant severe accident, using fresh fuel. Two options for fuel degradation models in MELCOR have been applied to fuel degradation behavior. the first model assumes that fuel debris will be formed immediately after the fuel support fails by cladding relocation due to the candling process. The other is the uncollapsed bare fuel pellets option, in which the fuel pellets remain standing in a columnar shape until the fuel reaches its melting point, even if the cladding has been relocated by candling. The thermal-hydraulic behaviors in the circuit and containment of Phebus FP are discussed herein. Flow velocities in the Phebus FP circuit are high in order to produce turbulent flow in a small diameter test pipe. The MELCOR calculation has shown that the length of the hot leg and steam generator are adequate to attain steam temperatures or 700{degrees}C and 150{degrees}C in the respective outlets. The containment atmosphere temperature and humidity derived by once through integral system calculation show that objective test conditions would be satisfied in the Phebus FP experiment.

  3. 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)

  4. 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

  5. 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.

  6. 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

  7. 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

  8. 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

  9. Modeling of the installation with the code MELCOR 1.8.4

    International Nuclear Information System (INIS)

    Pomier Baez, L.E.; Nunez Mc Leod, J.E.

    1998-01-01

    Full text: The calculation code MELCOR 1.8.4 is an integrated program that allow to simulate the development of accidents in nuclear plants with refrigerated reactors with light water. This code can simulate the whole spectrum of phenomenons. This work carried out the validation of the packages of the code MELCOR dedicated to evaluate the behaviour under conditions of two-phase flow, through the comparison of the results of the simulation with the experimental data of the installation TPTF (Two-Phase Test Facility) (ROSA-IV) of the Institute JAERI (Japan Atomic Energy Research Institute) of Japan. The main objective of the experiments TPTF is obtain data on the thermohydraulic behaviour from light water reactors (PWR) during an accident with small loss of coolant (SBLOCA), and the capacity of MELCOR code was evaluated in the simulation of these kind of accidents. Diverse options of the code were studies, in order to analyzing the behaviour of the feigned phenomenon. The effect of the change in the nodalization of the nuclear installation was studies, as well as the management of diverse control functions. The results of the evaluation show a good concordance with the experimental data, especially in the prediction of the behaviour of the steam fraction in relation with the mass flow, the quality of the steam and the mixture level in the exit volume that represent two possibilities state in the vessel reactor during the accidental situation. (author) [es

  10. 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)

  11. 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

  12. 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

  13. 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.

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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...

  13. 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

  14. 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.)

  15. 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

  16. 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)

  17. [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.

  18. 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)

  19. 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

  20. 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

  1. 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

  2. Development of an Input Model to MELCOR 1.8.5 for the Oskarshamn 3 BWR

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Lars [Lentek, Nykoeping (Sweden)

    2006-05-15

    An input model has been prepared to the code MELCOR 1.8.5 for the Swedish Oskarshamn 3 Boiling Water Reactor (O3). This report describes the modelling work and the various files which comprise the input deck. Input data are mainly based on original drawings and system descriptions made available by courtesy of OKG AB. Comparison and check of some primary system data were made against an O3 input file to the SCDAP/RELAP5 code that was used in the SARA project. Useful information was also obtained from the FSAR (Final Safety Analysis Report) for O3 and the SKI report '2003 Stoerningshandboken BWR'. The input models the O3 reactor at its current state with the operating power of 3300 MW{sub th}. One aim with this work is that the MELCOR input could also be used for power upgrading studies. All fuel assemblies are thus assumed to consist of the new Westinghouse-Atom's SVEA-96 Optima2 fuel. MELCOR is a severe accident code developed by Sandia National Laboratory under contract from the U.S. Nuclear Regulatory Commission (NRC). MELCOR is a successor to STCP (Source Term Code Package) and has thus a long evolutionary history. The input described here is adapted to the latest version 1.8.5 available when the work began. It was released the year 2000, but a new version 1.8.6 was distributed recently. Conversion to the new version is recommended. (During the writing of this report still another code version, MELCOR 2.0, has been announced to be released within short.) In version 1.8.5 there is an option to describe the accident progression in the lower plenum and the melt-through of the reactor vessel bottom in more detail by use of the Bottom Head (BH) package developed by Oak Ridge National Laboratory especially for BWRs. This is in addition to the ordinary MELCOR COR package. Since problems arose running with the BH input two versions of the O3 input deck were produced, a NONBH and a BH deck. The BH package is no longer a separate package in the new 1

  3. 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

  4. MELCOR 1.8.2 calculations of selected sequences for the ABWR

    International Nuclear Information System (INIS)

    Kmetyk, L.N.

    1994-07-01

    This report summarizes the results from MELCOR calculations of severe accident sequences in the ABWR and presents comparisons with MAAP calculations for the same sequences. MELCOR was run for two low-pressure and three high-pressure sequences to identify the materials which enter containment and are available for release to the environment (source terms), to study the potential effects of core-concrete interaction, and to obtain event timings during each sequence; the source terms include fission products and other materials such as those generated by core-concrete interactions. Sensitivity studies were done on the impact of assuming limestone rather than basaltic concrete and on the effect of quenching core debris in the cavity compared to having hot, unquenched debris present

  5. 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

  6. 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.

  7. 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

  8. 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)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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

  18. 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

  19. Overview of MELCOR 1.8.4: Modeling advances and assessment

    International Nuclear Information System (INIS)

    Gauntt, R.O.; Cole, R.K.; Rodriguez, S.B.; Young, M.F.; Gasser, R.D.

    1998-01-01

    The newly released MELCOR 1.8.4 reactor accident analysis code contains many new modeling features as well as improvements to existing models. New model additions to the MELCOR code include a model for predicting enhanced depletion rates for hygroscopic aerosols and a model for predicting the chemisorption of Cesium to the surfaces of piping. Improvements to existing models include: upgrading the core module (COR) to handle flow redistribution resulting from the formation of core blockages, improving the thermal hydraulics (CVH) coupling with COR to handle flow reversal situations, and upgrading the fission product scrubbing model to incorporate the SPARC90 code. Significant upgrading of the COR package core degradation modeling was also included in the new code release version. New and improved models are described in the following paper. In addition, a number of assessment analyses were recently performed, focusing on demonstrating the new and improved capabilities in the code. Results of assessment calculations demonstrating code performance for aerosol (pool) scrubbing, hygroscopic aerosol behavior, and core degradation and hydrogen production are presented. Finally, ongoing code developments activities beyond MELCOR 1.8.4 are described. These include models for treating iodine behavior in containment sumps, pools, and atmosphere, and plans for implementing reflood models and the attendant effects on accident progression. Further improvements and additions to the core degradation modeling in MELCOR are described, including the implementation of enhanced clad failure models to treat clad ballooning and eutectic interaction with grid spacers, and expansion of the COR package to allow for improved representation of UO 2 -Zr eutectic behavior, improved melt relocation treatment, greater detail in describing aspects of BWR core degradation (fuel channel, bypass, and lower plenum), and more flexibility in modeling other structures in the core such as core plate

  20. 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)

  1. 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)

  2. 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)

  3. Benchmarking MELCOR 1.8.2 for ITER Against Recent EVITA Results

    International Nuclear Information System (INIS)

    Merrill, Brad J.

    2007-01-01

    A version of MELCOR 1.8.2 modified for use in ITER Preliminary Safety Report analyses was validated against recent data from the EVITA facility located in Cadarache, France. EVITA Test Series 7 was used for this study to verify MELCOR's ability to predict the pressures, temperatures, cryoplate ice mass, and vacuum vessel (VV) condensate mass for test conditions in EVITA that include injections of steam, nitrogen, and water in to the EVITA VV after the walls had been heated to 165 C and the cryoplate had been cooled to -193 C. In general, the ability of MELCOR to predict the VV pressure and wall temperatures for the steam only and water only injection tests was very good. Predicted ice layer masses where larger than reported for the EVITA cryoplate, in particular for the steam only injection tests (∼40% too high), and the predicted condensate masses were less that measured in EVITA. Both of these discrepancies can be explained by ice porosity. The modified MELCOR 1.8.2 over predicts the EVITA VV pressure for the co-injection tests (e.g., steam plus nitrogen, or water plus nitrogen injections) by almost a factor of two. Based on parametric runs that where made by increasing the predicted cryoplate condensation rate, it is believed that this pressure over prediction is a result of an under predicted cryoplate condensation rate. The particulars of this study are documented in this report as well as conclusions about the impact this study has regarding the use of this version of MELCOR for consequence analyses for ITER safety reports

  4. Simulation of Iodine Behavior by Coupling of a Standalone Model with MELCOR

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-15

    During a severe accident, a large fraction of iodine in the core can be released into the containment. Iodine is important in terms of its high activity in the early phase after a core-melt accident due to its short half-life isotopes and its serious effect on the public health, especially on the thyroid. Therefore, iodine behavior has been extensively studied through the international research programs. Major research areas are iodine chemistry, surface reactions, mass transfer, modeling of iodine chemistry and its applications to severe accident assessment, and accident management. Advanced tools for modeling these phenomena have been developed and validated by several experiments such as ISTP-EPICUR (International Source Term Program - Experimental Program on Iodine Chemistry under Radiation) and PARIS, and OECD-BIP (Behavior of Iodine Project) in which Korea Institute of Nuclear Safety (KINS) has been participating. As a result, a simple iodine model, RAIM (Radio-Active Iodine chemistry Model) was developed, based on the IMOD methodology in order to deal with organic iodides conveniently. RAIM has been also coupled with MELCOR, replacing the pool chemistry model (PCM). This coupling model, MELCOR-RAIM, will be used for an integrated severe accident assessment that takes into account the organic iodine behavior. This model is described herein, and representative simulation results of the model are presented

  5. 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

  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. 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.

  8. 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

  9. 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

  10. 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.)

  11. 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

  12. 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

  13. 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

  14. MELCOR development for existing and advanced reactors

    International Nuclear Information System (INIS)

    Summers, R.M.

    1993-01-01

    Recent efforts in MELCOR development to address previously identified deficiencies have resulted in release of MELCOR 1.8.2, a much-improved version of the code. Major new models have been implemented for direct containment heating, ice condensers, debris quenching, lower plenum debris behavior, core materials interactions' and radial relocation of debris. Significant improvements have also been made in the modeling of interfacial momentum exchange and in the modeling of fission product release, condensation/evaporation, and aerosol behavior. Efforts are underway to address two-phase hydrodynamics difficulties, to improve modeling of water condensation on structures and fine-scale natural circulation within the reactor vessel, and to implement CORCON-Mod3. Improvements are also being made to MELCOR's capability to handle new features of the advanced light water reactor designs, including drainage of water films on connected heat structures, heat transfer from the external surface of the reactor vessel to a flooded cavity, and creep rupture failure of the lower head. Additional development needs in other areas are discussed

  15. 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

  16. 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.

  17. 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

  18. 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.

  19. Improvement of severe accident analysis method for KSNP

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Cho, Song Won; Cho, Youn Soo [Korea Radiation Technology Institute Co., Taejon (Korea, Republic of)

    2002-03-15

    The objective of this study is preparation of MELCOR 1.8.5 input deck for KSNP and simulation of some major severe accidents. The contents of this project are preparation of MELCOR 1.8.5 base input deck for KSNP to understand severe accident phenomena and to assess severe accident strategy, preparation of 20 cell containment input deck to simulate the distribution of hydrogen and fission products in containment, simulation of some major severe accident scenarios such as TLOFW, SBO, SBLOCA, MBLOCA, and LBLOCA. The method for MELCOR 1.8.5 input deck preparation can be used to prepare the input deck for domestic PWRs and to simulate severe accident experiments such as ISP-46. Information gained from analyses of severe accidents may be helpful to set up the severe accident management strategy and to develop regulatory guidance.

  20. 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)

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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.)

  6. 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

  7. 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

  8. 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

  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. 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

  11. 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)

  12. 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

  13. 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

  14. 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

  15. MELCOR 1.8.2 assessment: The MP-1 and MP-2 late phase melt progression experiments

    International Nuclear Information System (INIS)

    Tautges, T.J.

    1994-05-01

    MELCOR is a fully integrated, engineering-level computer code being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As a part of an ongoing assessment program, MELCOR has been used to model the MP-1 and MP-2 experiments, which provided data for late-phase melt progression in PWR geometries. Core temperature predicted by MELCOR were within 250--500 K of measured data in both MP-1 and MP-2. Relocation in the debris bed and metallic crust regions of MP-2 was predicted accurately compared to PIE data. Temperature gradients in lower portions of the test bundle were not predicted well in both MP-1 and MP-2, due to the lack of modeling of the heat transfer path to the cooling jacket in those portions of the test bundles. Fifteen sensitivity studies were run on various core (COR), control volume hydrodynamics (CVH) and heat structures (HS) package parameters. No unexpected sensitivities were found, and in particular there were no sensitivities to reduced time step, finer nodalization or to computer platform. Calculations performed by the DEBRIS and TAC2D codes for MP-1 and MP-2 showed better agreement with measured data than those performed by MELCOR. This was expected, through, due to the fully 2-dimensional modeling used in the other codes

  16. 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.

  17. 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

  18. 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

  19. Analysis of Peach Bottom station blackout with MELCOR

    International Nuclear Information System (INIS)

    Dingman, S.E.; Cole, R.K.; Haskin, F.E.; Summers, R.M.; Webb, S.W.

    1987-01-01

    A demonstration analysis of station blackout at Peach Bottom has been performed using MELCOR and the results have been compared with those from MARCON 2.1B and the Source Term Code Package (STCP). MELCOR predicts greater in-vessel hydrogen production, earlier melting and core collapse, but later debris discharge than MARCON 2.1B. The drywell fails at vessel breach in MELCOR, but failure is delayed about an hour in MARCON 2.1B. These differences are mainly due to the MELCOR models for candling during melting, in-core axial conduction, and continued oxidation and heat transfer from core debris following lower head dryout. Three sensitivity calculations have been performed with MELCOR to address uncertainties regarding modeling of the core-concrete interactions. The timing of events and the gas and radionuclide release rates are somewhat different in the base case and the three sensitivity cases, but the final conditions and total releases are similar

  20. 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

  1. 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

  2. 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.

  3. Risk Analysis of Fukushima Accident using MACCS2

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

    It has been three years since Fukushima Daiichi accident had occurred. Many efforts have been done for a restoration, however, radioactive materials are still released resulting in a crucial additional damage to a human health and economics and the scale of damage is not much evaluated. Therefore, an estimation of damage degree caused by the released radioactive materials right after a nuclear accident is essential to cope with additional radioactive problems. Here, we report the risk analysis of Fukushima Dai-ichi accident using MELCOR Accident Consequence Code System 2 (MACCS2), which is the Nuclear Regulatory Commission's (NRC's) code for evaluating off-site consequences. It is used in level-3 Probabilistic Risk Analyses (PRA), for planning purposes, for cost-benefit analyses and so on. The purpose of this study is to estimate radiological doses and health risks of Fukushima Daiichi accident through short- and long-term of lifetime using MACCS2. In summary, the health risk for inhabitants near Fukushima Daiichi NPP has been evaluated by considering the long term radiation effect using MACCS2 code. The result indicates that the occurrence and death rate of the cancer have been increased by the radioactive materials released from Fukushima Daiichi accident. The result obtained in this study may provide new insights for taking action after the nuclear reactor accident to mitigate the released radioactive materials and to prepare the countermeasure.

  4. 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)

  5. 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

  6. 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)

  7. 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)

  8. 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

  9. 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?.

  10. 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.

  11. 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)

  12. 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

  13. Benchmarking MELCOR 1.8.2 for ITER Against Recent EVITA Results

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J

    2007-11-01

    A version of MELCOR 1.8.2 modified for use in ITER Preliminary Safety Report analyses was validated against recent data from the EVITA facility located in Cadarache, France. EVITA Test Series 7 was used for this study to verify MELCOR’s ability to predict the pressures, temperatures, cryoplate ice mass, and vaccum vessel (VV) condensate mass for test conditions in EVITA that include injections of steam, nitrogen, and water in to the EVITA VV after the walls had been heated to 165 ºC and the cryoplate had been cooled to -193 ºC. In general, the ability of MELCOR to predict the VV pressure and wall temperatures for the steam only and water only injection tests was very good. Predicted ice layer masses where larger than reported for the EVITA cryoplate, in particular for the steam only injection tests (~40% too high), and the predicted condensate masses were less that measured in EVITA. Both of these descrpancies can be explained by ice porosity. The modified MELCOR 1.8.2 over predicts the EVITA VV pressure for the co-injection tests (e.g., steam plus nitrogen, or water plus nitrogen injections) by almost a factor of two. Based on parametric runs that where made by increasing the predicted cryoplate condensation rate, it is believed that this pressure over prediction is a result of an under predicted cryoplate condensation rate. The particulars of this study are documented in this report as well as conclusions about the impact this study has regarding the use of this verions of MELCOR for consequence analyses for ITER safety reports.

  14. 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.

  15. 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.)

  16. 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

  17. Review of psychological consequences of nuclear accidents and empirical study on peoples reactions to radiation protection activities in an imagined situation

    International Nuclear Information System (INIS)

    Haukkala, A.; Eraenen, L.

    1994-10-01

    The report consist of two parts: a review of studies on psychological consequences of nuclear and radiation accidents in population and an empirical study of peoples reactions to protection actions in an event of hypothetical accident. Review is based on research results from two nuclear reactor accidents (Three Mile Island 1979, Chernobyl 1986) and a radiation accident in Goiania, Brazil 1987. (53 refs, 2 figs.,7 tabs.)

  18. OCCUPATIONAL ACCIDENTS WITH BIOLOGICAL MATERIALS IN CLINICAL ANALYSIS LABORATORY: CAUSES AND CONSEQUENCES

    Directory of Open Access Journals (Sweden)

    T. M. Azevedo

    2014-07-01

    Full Text Available Accidents involving biological material can cause diseases to the professional healthcare and also bring psychosocial effects. The aim of this study was to characterize the accidents occurring with biological material with professional of clinical laboratories of Sinop-MT. Data were collected by a questionnaire consisting of sociodemographic and health variables. 21 (87.5% of respondents stated that they never suffered any kind of accident. One of the injured workers reported that there was involvement in your emotional life. It is observed underreporting of occupational accidents by employees affected, making it difficult to increase research on the subject and actions about the problem.

  19. MELCOR Applications to SOARCA and Fukushima

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.

    2014-03-01

    This PowerPoint presentation was organized as follows: Background; Overview of Fukushima Accidents; Comparisons of SOARCA Study with Fukushima accidents; Equipment functioning in real-world accidents; and, Conclusions.

  20. A 1ST Step Integration of the Restructured MELCOR for the MIDAS Computer Code

    International Nuclear Information System (INIS)

    Park, S. H.; Kim, D. H.; Cho, S. W.

    2006-01-01

    KAERI is developing a localized severe accident code, MIDAS, based on MELCOR. MELCOR uses pointer variables for a fixed-size storage management to save the data. It passes data through two depths, its meaning is not understandable by variable itself. So it is needed to understand the methods for data passing. This method deteriorates the readability, maintainability and portability of the code. As a most important process for a localized severe accident analysis code, it is needed convenient method for data handling. So, it has been used the new features in FORTRAN90 such as a dynamic allocation for the restructuring. The restructuring of the data saving and transferring method of the existing code makes it easy to understand the code. Before an entire restructuring of the code, a restructuring for each package was developed and tested. And then integration of each restructured package was being processed one by one. In this paper, the integrating scope includes the BUR, CF, CVH, DCH, EDF, ESF, MP, SPR, TF and TP packages. As most of them use data within each package and a few packages share data with other packages. The verification was done through comparing the results before and after the restructuring

  1. Assessment Of Source Term And Radiological Consequences For Design Basis Accident And Beyond Design Basis Accident Of The Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong; Tran Tri Vien

    2011-01-01

    The paper presents results of the assessment of source terms and radiological consequences for the Design Basis Accident (DBA) and Beyond Design Basis Accident (BDBA) of the Dalat Nuclear Research Reactor. The dropping of one fuel assembly during fuel handling operation leading to the failure of fuel cladding and the release of fission products into the environment was selected as a DBA for the analysis. For the BDBA, the introduction of a step positive reactivity due to the falling of a heavy block from the rotating bridge crane in the reactor hall onto a part of the platform where are disposed the control rod drives is postulated. The result of the radiological consequence analyses shows that doses to members of the public are below annual dose limit for both DBA and BDBA events. However, doses from exposure to operating staff and experimenters working inside the reactor hall are predicted to be very high in case of BDBA and therefore the protective actions should be taken when the accident occurs. (author)

  2. Fukushima accident: the consequences in Japan, France and in Japan; Accident de Fukushima: les repercusions au Japon, en France et dans le Japon

    Energy Technology Data Exchange (ETDEWEB)

    Foucher, N.; Sorin, F.

    2011-03-15

    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.)

  3. Hygienic evaluation of radiation consequences after the Chernobyl accident in highly populated areas of Ukraine

    International Nuclear Information System (INIS)

    Karachev, I.I.; Tkachenko, N.V.; Markelova, L.K.

    1993-01-01

    This paper addresses radiation exposure to the people in the Ukraine as a result of the Chernobyl accident. As a result of this accident all Ukrainian regions have been heavily contaminated, even though the contamination density obtained in different regions are considerably different. Soils have become contaminated and plants grown in the soils transfer radionuclides to people

  4. Accident Damage Analysis Module (ADAM) – Technical Guidance, Software tool for Consequence Analysis calculations

    OpenAIRE

    FABBRI LUCIANO; BINDA MASSIMO; BRUINEN DE BRUIN YURI

    2017-01-01

    This report provides a technical description of the modelling and assumptions of the Accident Damage Analysis Module (ADAM) software application, which has been recently developed by the Joint Research Centre (JRC) of the European Commission (EC) to assess physical effects of an industrial accident resulting from an unintended release of a dangerous substance

  5. Comparison of MELCOR modeling techniques and effects of vessel water injection on a low-pressure, short-term, station blackout at the Grand Gulf Nuclear Station

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1995-06-01

    A fully qualified, best-estimate MELCOR deck has been prepared for the Grand Gulf Nuclear Station and has been run using MELCOR 1.8.3 (1.8 PN) for a low-pressure, short-term, station blackout severe accident. The same severe accident sequence has been run with the same MELCOR version for the same plant using the deck prepared during the NUREG-1150 study. A third run was also completed with the best-estimate deck but without the Lower Plenum Debris Bed (BH) Package to model the lower plenum. The results from the three runs have been compared, and substantial differences have been found. The timing of important events is shorter, and the calculated source terms are in most cases larger for the NUREG-1150 deck results. However, some of the source terms calculated by the NUREG-1150 deck are not conservative when compared to the best-estimate deck results. These results identified some deficiencies in the NUREG-1150 model of the Grand Gulf Nuclear Station. Injection recovery sequences have also been simulated by injecting water into the vessel after core relocation started. This marks the first use of the new BH Package of MELCOR to investigate the effects of water addition to a lower plenum debris bed. The calculated results indicate that vessel failure can be prevented by injecting water at a sufficiently early stage. No pressure spikes in the vessel were predicted during the water injection. The MELCOR code has proven to be a useful tool for severe accident management strategies

  6. Assumptions used for evaluating the potential radiological consequences of a less of coolant accident for pressurized water reactors - June 1974

    International Nuclear Information System (INIS)

    Anon.

    1974-01-01

    Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility. The design basis loss of coolant accident is one of the postulated accidents used to evaluate the adequacy of these structures, systems, and components with respect to the public health and safety. This guide gives acceptable assumptions that may be used in evaluating the radiological consequences of this accident for a pressurized water reactor. In some cases, unusual site characteristics, plant design features, or other factors may require different assumptions which will be considered on an individual case basis. The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position

  7. Assumptions used for evaluating the potential radiological consequences of a loss of coolant accident for boiling water reactors - June 1974

    International Nuclear Information System (INIS)

    Anon.

    1974-01-01

    Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility. The design basis loss of coolant accident is one of the postulated accidents used to evaluate the adequacy of these structures, systems, and components with respect to the public health and safety. This guide gives acceptable assumptions that may be used in evaluating the radiological consequences of this accident for a pressurized water reactor. In some cases, unusual site characteristics, plant design features, or other factors may require different assumptions which will be considered on an individual case basis. The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position

  8. 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

  9. KINS Research Activities on the iodine behavior in containment during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hanchul; Kim, Dosam [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Oh, Jaeyong; Yun, Jongil [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Cho, Songwon [Korea Radiation Technology Institute, Daejeon (Korea, Republic of)

    2012-03-15

    Iodine is a major contributor to the potential health risk for the public following a severe accident from a nuclear power plant. Volatile iodine and organic iodides can be generated from the containment sump through various kinds of reactions and be released to the environment. This iodine behavior has been an important topic for the international research programs run by the OECD/NEA and EU-SARNET2. Korea Institute of Nuclear Safety (KINS) also has joined ISTP-EPICUR (Experimental Program on Iodine Chemistry under Radiation) and OECD-BIP (Behavior of Iodine Project). In the course of researching this issue with these experimental programs, a simple iodine model, RAIM, has been developed and coupled with the MELCOR code for radiological consequence analysis. This methodology is likely to provide a technical basis for developing the regulatory requirements concerning a severe accident including accident source term, which is one of urgent domestic needs.

  10. Evaluation of severe accident risks: Methodology for the containment, source term, consequence, and risk integration analyses. Volume 1, Revision 1

    International Nuclear Information System (INIS)

    Gorham, E.D.; Breeding, R.J.; Brown, T.D.; Harper, F.T.; Helton, J.C.; Murfin, W.B.; Hora, S.C.

    1993-12-01

    NUREG-1150 examines the risk to the public from five nuclear power plants. The NUREG-1150 plant studies are Level III probabilistic risk assessments (PRAs) and, as such, they consist of four analysis components: accident frequency analysis, accident progression analysis, source term analysis, and consequence analysis. This volume summarizes the methods utilized in performing the last three components and the assembly of these analyses into an overall risk assessment. The NUREG-1150 analysis approach is based on the following ideas: (1) general and relatively fast-running models for the individual analysis components, (2) well-defined interfaces between the individual analysis components, (3) use of Monte Carlo techniques together with an efficient sampling procedure to propagate uncertainties, (4) use of expert panels to develop distributions for important phenomenological issues, and (5) automation of the overall analysis. Many features of the new analysis procedures were adopted to facilitate a comprehensive treatment of uncertainty in the complete risk analysis. Uncertainties in the accident frequency, accident progression and source term analyses were included in the overall uncertainty assessment. The uncertainties in the consequence analysis were not included in this assessment. A large effort was devoted to the development of procedures for obtaining expert opinion and the execution of these procedures to quantify parameters and phenomena for which there is large uncertainty and divergent opinions in the reactor safety community

  11. Evaluation of Coolant Injection Procedure in the Severe Accident Management Strategy of APR1400

    International Nuclear Information System (INIS)

    Cho, Yongjin; Lim, Kukhee; Song, Sungchu; Lee, Sukho; Hwang, Taesuk

    2013-01-01

    A coolant injection strategy in the severe accident management guideline (SAMG) of APR1400 relates to immediate coolant injection into RCS (Reactor Coolant System) or injection following the recovery of secondary coolant inventory. This strategy could play important role in accident mitigation and radiological consequences. In this study, appropriateness of the strategy was evaluated using MELCOR1.8.6 and several sensitivity studies of the key parameters were performed. Analysis for APR1400 using MELCOR 1.8.6 was performed to evaluate the effectiveness of accident management strategies and the following conclusions were identified. Sequential operation of secondary and RCS injection may not be the best strategy and the simultaneous injection of secondary and RCS injection could be more preferable. At least, the RCS injection should start before complete drainage of water in the safety injection tank using mobile pumps. In this study, the effectiveness of timing of operator action has been examined and the amount of injection flowrate needs to be studied in the future

  12. A critical review of Jan Beyea's report: A study of some of the consequences of hypothetical reactor accidents at Barsebaeck

    International Nuclear Information System (INIS)

    Gjoerup, H.L.; Hedemann Jensen, P.; Jensen, N.O.; Pejtersen, V.; Lundtang Petersen, E.; Petersen, T.; Thykier-Nielsen, S.; Heikel Vinther, F.

    1978-04-01

    This report contains a critical review of Jan Beyea's report: A study of some of the consequences of hypothetical reactor accidents at Barsebaeck (Princeton University, January 1978). Unreasonable assumptions concerning dry deposition, plume rise, meteorological considerations, dose-response relationship and probability distributions were found in the report. It is found that the conclusions of the Beyea report are the result of a mathematical exercise rather than the results of a realistic risk evaluation for Barsebaeck. (author)

  13. Human exposure to radiation following the release of radioactivity from a reactor accident: a quantitative assessment of the biological consequences

    International Nuclear Information System (INIS)

    Smith, H.; Stather, J.W.

    1976-11-01

    The objective of this review is to provide a biological basis upon which to assess the consequences of the exposure of a population to radioactivity released after a reactor accident. Depending upon the radiation dose, both early and late somatic damage could occur in the exposed population and hereditary effects may occur in their descendants. The development of dose-effect relationships has been based upon the limited amount of information available on humans, supplemented by data obtained from experiments on animals. (author)

  14. Elimination of the consequences of radiation accidents at the Mayak production association in the 1950s and 1960s

    International Nuclear Information System (INIS)

    Drozhko, E.G.; Romanov, G.N.

    2002-01-01

    The paper describes the consequences of radiation accidents happened at Mayak production association located in Chelyabinsk region, Urals, Russia, and countermeasures applied for reduction of radiation exposure of local population. The assessment of the efficiency of countermeasures based on the averted dose criterion is presented. It is stated that the most efficient measures on prevention of the population exposure were relocation of the population and construction of the Techa Reservoir Cascade. (author)

  15. Scientometric analysis of the means of scientific communication of the problem of medical consequences of Chernobyl Nuclear accident

    International Nuclear Information System (INIS)

    Artamonova, N.O.; Kulyinyich, G.V.; Pavlyichenko, Yu.V.; Gorvan', A.Je.; Zakrut'ko, L.Yi.; Novgorods'ka, L.M.; Byilan, L.G.

    2014-01-01

    In this paper evaluation of the structure and trends in the development of the Ukrainian scientific communication tools on the medical consequences of the Chernobyl nuclear accident using bibliometric methods has been given. The main developers of methodical documents are allocated, the dynamics of the distribution of methodical references, information letters and innovations is estimated. The importance of scientific communications tools in dissemination and use of new medical knowledge is demonstrated

  16. Evaluation of sanitary consequences of Chernobylsk accident in France. Epidemiological surveillance plan, state of knowledge, risks evaluation and perspectives

    International Nuclear Information System (INIS)

    Verger, P.; Cherie-Challine, L.

    2000-12-01

    This report jointly written by IPSN and InVS, reviews the sanitary consequences in France of the Chernobyl accident, which occurred in 1986. The first point is dedicated to a short presentation of the knowledge relative to the sanitary consequences of the Chernobyl accident in the high contaminated countries and to the risk factors of the thyroid cancer. Secondly, this report describes the main systems of epidemiological surveillance of health implemented in France in 1986 and in 1999, as well as the data of the incidence and mortality of thyroid cancer observed in France since 1975. In addition, this report presents an analysis of the risk of thyroid cancer related to radioactive contamination in France, for young people of less than 15 years of age who where living in 1986 in the highest contaminated areas of France (Eastern territories). For this purpose, the theoretical number of thyroid cancers in excess is evaluated for this population, on the basis of different available risk model. Finally starting from the results of risk assessment, there is a discussion about the relevance and the feasibility of different epidemiological methods in view of answering the questions related to the sanitary consequences of the Chernobyl accident. In conclusion, this report recommends to reinforce the surveillance of thyroid cancer in France. (author)

  17. Summary of the consequences for safety which result from the Three-Mile-Island accident

    International Nuclear Information System (INIS)

    Smidt, D.

    1982-01-01

    The paper focusses on the Three-Mile-Island (TMI) accident in terms of reactor safety, and describes the first stage of the event's course (the first 2 hours and 18 minutes), the second stage (up to 16 hours after accident onset) and the stage till ultimate transition to stationary cooling. Conclusions are drawn for plant design and control room concepts. In conclusion, problems of staff training for critical situations are discussed. (HAG) [de

  18. RADIATION HYGIENIC CONSEQUENCES OF THE ACCIDENT AT THE CHERNOBYL NPP AND THE TASKS OF THEIR MINIMIZATION

    Directory of Open Access Journals (Sweden)

    G. G. Onischenko

    2009-01-01

    Full Text Available The paper presents data on the role and results of activities of Rospotrebnadzor bodies and institutions in the field of ensuring population radiation protection during various periods since accident at the Chernobyl NPP. Radiation hygienic characterization of territories affected by radioactive contamination from the accident, population exposure dose range, issues of ensuring radiological well-being of population and ways of their solution are being presented in the paper.

  19. The psychological results of the consequence of the ChNPP accident

    International Nuclear Information System (INIS)

    Salamatov, V.A.; Volodina, I.A.

    1992-01-01

    The ChNPP accident grought about the following stress reactions among the exposed population registered: psychosomatic reactions (headaches, gastro-intestinal tract disorders pains in the heart etc.), psychoemotional reactions (panic, sleep disturbance, restlessness etc.), behaviour reaction (aggressiveness, apathy) the whole structure of the human psychic is stated to have been affected by the Chernobyl' accident. The changes in the emotion-and-will sphere appear most distinctly. 2 tab

  20. The radiological accident of Goiania and its consequences for the development of law

    International Nuclear Information System (INIS)

    Franco, N.M.F.L.

    1998-01-01

    The radiological accident of Goiania and its repercussions caused intense debate in Brazilian society, which extended to the legislative sphere. One of the principal outcomes of this debate was the inclusion in the new Brazilian Constitutional Charter of legal provisions covering the control of nuclear energy and of radiation sources. Internationally, the 1986 Vienna Convention on Early Notification of a Nuclear Accident and the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency were invoked following the accident and proved to be effective in facilitating international co-operation and solidarity to deal with the aftermath of the accident. A number of international treaties on assistance in the event of nuclear accidents, the management of radioactive waste and the management of spent fuel are currently in force. The Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste, adopted in 1997, is the most recent treaty promoting the sustainability of nuclear activities. Looking ahead, an international legal framework is needed to build upon and improve the principles of a culture of radiation safety. (author)

  1. Methodology Using MELCOR Code to Model Proposed Hazard Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Gavin Hawkley

    2010-07-01

    This study demonstrates a methodology for using the MELCOR code to model a proposed hazard scenario within a building containing radioactive powder, and the subsequent evaluation of a leak path factor (LPF) (or the amount of respirable material which that escapes a facility into the outside environment), implicit in the scenario. This LPF evaluation will analyzes the basis and applicability of an assumed standard multiplication of 0.5 × 0.5 (in which 0.5 represents the amount of material assumed to leave one area and enter another), for calculating an LPF value. The outside release is dependsent upon the ventilation/filtration system, both filtered and un-filtered, and from other pathways from the building, such as doorways (, both open and closed). This study is presents ed to show how the multiple leak path factorsLPFs from the interior building can be evaluated in a combinatory process in which a total leak path factorLPF is calculated, thus addressing the assumed multiplication, and allowing for the designation and assessment of a respirable source term (ST) for later consequence analysis, in which: the propagation of material released into the environmental atmosphere can be modeled and the dose received by a receptor placed downwind can be estimated and the distance adjusted to maintains such exposures as low as reasonably achievableALARA.. Also, this study will briefly addresses particle characteristics thatwhich affect atmospheric particle dispersion, and compares this dispersion with leak path factorLPF methodology.

  2. Assessment of the MELCOR 1.8.6 condensation heat transfer model under the presence of noncondensable gases

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Ji Min; Lee, Dong Hun; Jeong, Jae Jun [Pusan National University, Busan (Korea, Republic of)

    2016-05-15

    Condensation heat transfer under the presence of noncondensable gases (NCGs) is an important issue in nuclear safety because the presence of even a small quantity of NC gases in the vapor largely reduces the condensation rate. The extensive assessment of the condensation model of the safety analysis codes has been also performed. When NCGs are present, the condensation phenomenon is largely reduced by accumulated NCGs near the condensing surface. Since the total pressure remains constant, the partial pressure of vapor at the liquid-vapor interface is lower than that in the bulk mixture, providing the driving force for vapor diffusion towards the liquid-vapor interface. The main objective of the present study is the assessment of the condensation heat transfer model of the severe accident code MELCOR 1.8.6 under the presence of NCGs. In this study, the condensation heat transfer model of the MELCOR 1.8.6 is assessed using various experiments which have 4 different types of geometry. Through the comparison of the results, it was shown that the MELCOR code generally under-predicts the condensation heat transfer except the condensation on outer surface of vertical pipes and improvement is needed for other geometries.

  3. 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)

  4. 10 years from the Chernobyl nuclear reactor accident: consequences and lesson learned

    International Nuclear Information System (INIS)

    1996-01-01

    Published jointly by the Czech State Office for Nuclear Safety and the Czech National Radiation Protection Institute, the publication gives a succinct account of the cause of the Chernobyl accident and its impact on the former Soviet Union, and concentrates on the effects of the accident on the Czech Republic. The topics dealt with in this respect include, among others: radionuclide contents of foods with particular emphasis on milk products for babies, assessment of surface contamination of the Czech Republic due to the accident, internal contamination of the population as determined by whole-body measurements, assessment of the effective dose equivalents from external irradiation and effective dose equivalent commitments from internal irradiation, cesium radioisotopes in natural ecosystems, and the use of post-Chernobyl monitoring to test radionuclide migration models within the IAEA VAMP programme. (P.A.). 12 tabs., 30 figs., 64 refs

  5. Evaluation of food chain transfer data for use in accident consequence assessment

    International Nuclear Information System (INIS)

    Coughtrey, P.J.; Kirton, J.A.; Mitchell, N.G.

    1991-01-01

    Input data for the food chain transport component of radiological assessment models are summarised in the context of the sources of information available prior to the Chernobyl accident and those derived after the accident. Particular attention is devoted to interception and retention soil-to-plant, and plant-to-animal transfer, and to the applicability of environmental data to both equilibrium and time-dependent models. It is argued that much of the current uncertainty in parameter values for use in radiological assessment models reflects lack of understanding of processes involved in the various stages of transfer of radionuclides to man. The Chernobyl accident highlighted this lack of fundamental knowledge and illustrated a number of areas where further research and model development is justified. These areas are identified and suggestions given for appropriate research to support model development

  6. The influence of the natural environment on the radiological consequences of a nuclear accident

    International Nuclear Information System (INIS)

    Grauby, A.

    1989-01-01

    Much work has been done in the agri-food field on the evaluation of the health implications of a nuclear accident. Much less has been done on environmental dynamics. Research on the behaviour of radioactive deposits from the Chernobyl accident, however, has revealed the importance of natural conditions for the transfer of radionuclides in the environment. In particular, two phases have been observed. In the first, which lasts until a point of equilibrium is reached in the environment, radionuclides follow the water cycle. The second phase unfolds following the incorporation of ruthenium and cesium in soil and sediment. Certain environmental factors (natural ecosystems, relief, etc.) and the different meteorological conditions (climate and season) should be investigated in more detail. Better understanding of the effect of natural conditions could be used to predict, in time and space, any specific risks arising during nuclear accidents whether located near or far [fr

  7. Evaluation of sanitary consequences of Chernobyl accident in France: epidemiological monitoring device, state of knowledge, evaluation of risks and perspectives

    International Nuclear Information System (INIS)

    Verger, P.; Champion, D.; Gourmelon, P.; Hubert, Ph.; Joly, J.; Renaud, Ph.; Tirmarche, M.; Vidal, M.; Cherie-Challine, L.; Boutou, O.; Isnard, H.; Jouan, M.; Pirard, Ph.

    2000-12-01

    The objectives of this document are firstly, to present the situation of knowledge both on the sanitary consequences of the Chernobyl accident and on the risk factors of thyroid cancers, these ones constituting one of the most principal consequences observed in Belarus, in Ukraine and Russia; secondly, the give the principal system contributing to the epidemiological surveillance of effects coming from a exposure to ionizing radiations, in France and to give the knowledge on incidence and mortality of thyroid cancer in France; thirdly, to discuss the pertinence and the feasibility of epidemiological approaches that could be considered to answer questions that the public and authorities ask relatively to the sanitary consequences of Chernobyl accident in France; fourthly to male a calculation of thyroid cancer risk in relation with Chernobyl fallout in France from works and studies made from 1986 on the consequences of this disaster in terms of radioecology and dosimetry at the national level. Besides, the improvement of thyroid cancer surveillance is also tackled. (N.C.)

  8. Consequences in Norway of a hypothetical accident at Sellafield: Potential release - transport and fallout

    International Nuclear Information System (INIS)

    Ytre-Eide, M. A.; Standring, W.J.F.; Amundsen, I.; Sickel, M.; Liland, A.; Salt