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Sample records for accident consequences pwr

  1. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    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...

  2. Degraded core accidents for the Sizewell PWR: A sensitivity analysis of the radiological consequences

    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, particular to the releases analysed from Sizewell; for different releases from different locations the sensitivity may change significantly. In the earlier study and analysis was undertaken of the impact on the predicted consequences of potential overestimates in the release fractions of radionuclides. Since the results of that study were published some relatively minor numerical errors have been identified. While none of these affects the conclusions reached in that study the opportunity has been taken in this report to present revised values for those results known to be in error. This revised text and results are presented as an appendix to this report and they replace the corresponding material in the earlier study. (author)

  3. The radiological consequences of degraded core accidents for the Sizewell PWR The impact of adopting revised frequencies of occurrence

    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.

  4. An assessment of the radiological consequences of releases to groundwater following a core-melt accident at the Sizewell PWR

    In the extremely unlikely event of a degraded core accident at the proposed Sizewell PWR it is theoretically possible for the core to melt through the containment, after which activity could enter groundwater directly or as a result of subsequent leaching of the core in the ground. The radiological consequences of such an event are analysed and compared with the analysis undertaken by the NRPB for the corresponding releases to atmosphere. It is concluded that the risks associated with the groundwater route are much less important than those associated with the atmospheric route. The much longer transport times in the ground compared with those in the atmosphere enable countermeasures to be taken, if necessary, to restrict doses to members of the public to very low levels in the first few years following the accident. The entry of long-lived radionuclides into the sea over very long timescales results in the largest contribution to population doses, but these are delivered at extremely low dose rates which would be negligible compared with background exposure. (author)

  5. PWR Core 2 Project accident analysis

    The various operations required for receipt, handling, defueling and storage of spent Shippingport PWR Core 2 fuel assemblies have been evaluated to determine the potential accidents and their consequences. These operations will introduce approximately 16,500 kilograms of depleted natural uranium (as UO2), 139 kilograms of plutonium, 2.8 megacuries of mixed fission products, and 14 kilograms of Zircaloy-4 (cladding and hardware) into the 221-T Canyon Building. Event sequences for potential accidents that were considered included (1) leaking fuel assemblies, (2) fire and explosion, (3) loss of coolant or cooling capability, (4) dropped and/or damaged fuel assemblies, and extrinsic occurrences such as loss of services, missile impact, and natural occurrences (e.g., earthquake, tornado). Accident frequencies were determined by formal analysis to be very low. Accident consequences are greatly mitigated by the safety and containment features designed into the fuel modules and shipping cask, the long cooling time since reactor discharge, and the redundant safety features designed into the facilities, equipment, and operating procedures for the PWR Core 2 Project. Possible hazards associated with the handling of these fuels have been considered and adequate safeguards and storage constraints identified. The operations of M-160 cask unloading and module storage will not involve identifiable risks as great or significantly greater than those for comparable licensed nuclear facilities, nor will hazards or risks be significantly different from comparable past 221-T Plant programs. Therefore, it is concluded that the operations required for receipt, handling, and defueling of the M-160 cask and for the storage and surveillance of the PWR Core 2 fuel assemblies at the 221-T Canyon Building can be performed without undue risk to the safety of the involved personnel, the public, the environment or the facility

  6. Analysis of reactivity accidents in PWR'S

    This note describes the French strategy which has consisted, firstly, in examining all the accidents presented in the PWR unit safety reports in order to determine for each parameter the impact on accident consequences of varying the parameter considered, secondly in analyzing the provisions taken into account to restrict variation of this parameter to within an acceptable range and thirdly, in checking that the reliability of these provisions is compatible with the potential consequences of transgression of the authorized limits. Taking into consideration violations of technical operating specifications and/or non-observance of operating procedures, equipment failures, and partial or total unavailability of safety systems, these studies have shown that fuel mechanical strength limits can be reached but that the probability of occurrence of the corresponding events places them in the residual risk field and that it must, in fact, be remembered that there is a wide margin between the design basis accidents and accidents resulting in fuel destruction. However, during the coming year, we still have to analyze scenarios dealing with cumulated events or incidents leading to a reactivity accident. This program will be mainly concerned with the impact of the cases examined relating to dilution incidents under normal operating conditions or accident operating conditions

  7. 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

    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.

  8. The Chernobyl accident consequences

    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.)

  9. Identification and evaluation of PWR in-vessel severe accident management strategies

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents

  10. Accidents, risks and consequences

    Although the accident at Chernobyl can be considered as the worst accident in the world, it could have been worse. Other far worse situations are considered, such as a nuclear weapon hitting a nuclear reactor. Indeed the accident at Chernobyl is compared to a nuclear weapon. The consequences of Chernobyl in terms of radiation levels are discussed. Although it is believed that a similar accident could not occur in the United Kingdom, that possibility is considered. It is suggested that emergency plans should be made for just such an eventuality. Even if Chernobyl could not happen in the UK, the effects of accidents are international. The way in which nuclear reactor accidents happen is explored, taking the 1957 Windscale fire, Three Mile Island and Chernobyl as examples. Reactor designs and accident scenarios are considered. The different reactor designs are listed. As well as the Chernobyl RBMK design it is suggested that the light water reactors also have undesirable features from the point of view of safety. (U.K.)

  11. REWET, PWR LOCA accident experiments

    1 - Description of test facility: The REWET-II facility was designed for the investigation of the reflooding phase of a LOCA. The main design principle is the accurate simulation of the rod bundle geometry and the primary system elevations. This is necessary in order to have the correct flow channels and hydrostatic pressures for the reflooding process. The reactor vessel is simulated by a stainless steel U-tube construction consisting of downcomer, lower plenum, core and upper plenum. The primary loops contain a pipe simulating the broken loop and a connection line between the upper plenum and the downcomer simulating five intact loops. The containment is simulated by a pressure vessel (not in scale). Steam generators and primary pumps are simulated with flow resistances. The ECC-water can be injected to the downcomer and/or to the upper plenum by a pump or from an accumulator. All the elevation in the reactor vessel simulator are scaled to 1:1 (except the reactor upper head). The scale of the volumes and flow areas is 1:2333 referring to the number of the fuel rod simulators in the facility and the fuel rods in the reference reactor. The rod bundle is either in a hexagonal shroud, the inside distance of the opposite walls is 54.3 mm and the wall thickness 2 mm, or in a round shroud, the inner diameter is 66.0 mm and the wall thickness 2 mm. The simulation of the fuel-rod bundle consists of 19 indirectly electrically heated simulator rods. The heating coils are inside stainless steel claddings in magnesium oxide insulation. The heated length, the outer diameter and the lattice pitch of the fuel-rod simulators as well as the number (= 10) and construction of the rod bundle spacers are the same as in the reference reactor. The upper ends of the rods are attached to the upper tie plate. 2 - Description of test: Pressurized water reactors in use in Finland reactors have certain unique features which make them different from most other PWR designs. The 6 horizontal

  12. Accidents, probabilities and consequences

    Following brief discussion of the safety of wind-driven power plants and solar power plants, some aspects of the safety of fast breeder and thermonuclear power plants are presented. It is pointed out that no safety evaluation of breeders comparable to the Rasmussen investigation has been carried out and that discussion of the safety aspects of thermonuclear power is only just begun. Finally, as an illustration of the varying interpretations of risk and safety analyses, four examples are given of predicted probabilities and consequences in Copenhagen of the maximum credible accident at the Barsebaeck plant, under the most unfavourable meterological conditions. These are made by the Environment Commission, Risoe Research Establishment, REO (a pro-nuclear group) and OOA (an anti-nuclear group), and vary by a factor of over 1000. (JIW)

  13. Study on severe accident mitigation measures for the development of PWR SAMG

    2006-01-01

    In the development of the Severe Accident Management Guidelines (SAMG), it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident (LOCA), Steam Generator Tube Rupture (SGTR), Station Blackout (SBO), and Anticipated Transients without Scram (ATWS) in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management (SAM) such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.

  14. Accident consequence assessment code development

    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)

  15. Consequences of the Chernobyl accident

    The techniques currently used in off-site consequence modelling are applied to the Chernobyl accident. Firstly, the time dependent spread of radioactive material across the European continent is considered, followed by a preliminary assessment of the dosimetric impact (in terms of collective and mean individual doses) on the various countries of Eastern and Western Europe. The consequences of the accident in the USSR are also discussed. Finally, the likely implications of the Chernobyl event on research in the field of environmental consequence assessment are outlined. (author)

  16. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Andrej Prošek; Leon Cizelj

    2013-01-01

    Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO). Long-term SBO in a pressurized water reactor (PWR) leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures) are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pump...

  17. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  18. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

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

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO2–Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  20. PWR auxiliary systems, safety and emergency systems, accident analysis, operation

    The author presents a description of PWR auxiliary systems like volume control, boric acid control, coolant purification, -degassing, -storage and -treatment system and waste processing systems. Residual heat removal systems, emergency systems and containment designs are discussed. As an accident analysis the author gives a survey over malfunctions and disturbances in the field of reactor operations. (TK)

  1. Serious accidents of PWR type reactors for power generation

    This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

  2. Characteristics of the aerosols released to the environment after a severe PWR accident

    In the event of a postulated severe accident on a pressurized water reactor (PWR) involving fuel degradation, gases and aerosols containing radioactive products could be released, with short, medium and long term consequences for the population and the environment. Under such accident conditions, the ESCADRE code system, developed at IPSN (Institute for Nuclear Safety and Protection) can be used to calculate the properties of the substances released and, especially with the AEROSOLS/B2 code, the main characteristics of the aerosols (concentration, size distribution, composition). For conditions representative of severe PWR accidents, by varying different main parameters (structural material aerosols, steam condensation in the containment, etc...), indications are given on the range of characteristics of the aerosols (containing notably Cs, Te, Sr, Ru, etc...) released to the atmosphere. Information is also given on how more accurate data (especially on the chemical forms) will be obtainable in the framework of current or planned experimental programs (HEVA, PITEAS, PHEBUS PF, etc...)

  3. PWR pressure vessel integrity during overcooling accidents

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

  4. PWR accident management realated tests: some Bethsy results

    The BETHSY integral test facility which is a scaled down model of a 3 loop FRAMATOME PWR and is currently operated at the Nuclear Center of Grenoble, forms an important part of the French strategy for PWR Accident Management. In this paper the features of both the facility and the experimental program are presented. Two accident transients: a total loss of feedwater and a 2'' cold leg break in case of High Pressure Safety Injection System failure, involving either Event Oriented - or State Oriented-Emergency Operating Procedures (EO-EOP or SO-EOP) are described and the system response analyzed. CATHARE calculation results are also presented which illustrate the ability of this code to adequately predict the key phenomena of these transients. (authors). 13 figs., 11 refs., 2 tabs

  5. Medical consequences of radiation accidents

    Since 1945, more than 1.8 x 1021 Bq of artificial radionuclides have been released into the atmosphere. Approximately 2.04 x 1018B, i.e. approx. 0.11%, are the result of accidents at nuclear industrial facilities. This percentage is causing increased interest among researchers. This is due to the fact that in the wake of accidental release radionuclides become distributed unevenly across the Earth's surface, and the associated exposures, fluctuating from background level to several grays, an induce both stochastic and deterministic effects in the irradiated population. A comparative analysis of the medical consequences of the twentieth century's most serious nuclear events, namely the authorized dumping of high level radioactive waste into the river Techa in 1950, the explosion of a storage tank containing long lived radioactive waste in the Southern Urals in 1957, the fire at Sellafield in 1957 and the accident at the Chernobyl nuclear power plant in 1986, has shown that when timely countermeasures are taken, the worst immediate and delayed medical consequences of an accident can be avoided. The consequences that have since been ascertained are a brief rise in the mortality rate during the first five years, with a dose in excess of 500 mSv; an increase in the incidence of leukaemia, with an absolute risk of up to 1.1. x 10-4 man·years/Gy; and increased mortality among children with external radiation doses of up to 1000 mSv, and internal doses of 99-190 mSv on the bone surfaces of neonates or 170-600 mSv on the bone surfaces of the mother. There is reliable evidence that, with external gamma radiation doses in excess of 520 mSv, the mortality rate for all malignant tumorous increases by 45-58% compared with the control level. There is also a significant increase in thyroid cancer frequency four to ten years after the incorporation of iodine isotopes by children aged up to 7 years, including an accumulation period in the womb. (author). 12 refs, 7 tabs

  6. Application of PCTRAN-3/U to studying accident management during PWR severe accident

    In order to improve the safety of nuclear power plant, operator action should be taken into account during a severe accident. While it takes a long time to simulate the plant transient behavior under a severe accident in comparison with the design based accident, a transient simulator should have both high speed calculation capability and interactive functions to model the operating procedures. PCTRAN has been developing to be a simple simulator by using a personal computer to simulate plant behavior under an accident condition. While currently available means usually take relatively long time to simulate plant behavior, using a current high-powered personal computer (PC), PCTRAN-3/U code is designed to operate at a speed significantly faster than real-time. The author describes some results of PCTRAN application in studying the efficiency of accident management for a pressurized water reactor (PWR) during an severe accident

  7. Analysis of reactivity insertion accidents in PWR reactors

    A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (Author)

  8. Level 3 PSA and it's implementation for PWR accident

    Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA) is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidents of PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The series of calculations performed are: calculate the source terms of the core damaged, modeling of meteorological conditions and environmental site, exposure pathway modeling, analysis of radionuclide dispersion and transport phenomena in the environment, radionuclide deposition analysis, analysis of radiation dose, protection & mitigation analysis, and risk analysis. The assessment uses a series of subsystems on PC Cosyma software. The results prove that the safety assessment using Level 3 PSA methodology is very effective and comprehensive estimate the impact, consequences, risks, nuclear emergency preparedness, and the reactor accident management especially for severe accidents or beyond design basis accidents of nuclear power plants. The results of the assessment can be used as a feedback to safety assessment of Level 1 PSA and Level 2 PSA. (author)

  9. Psychiatric consequences of road traffic accidents.

    Mayou, R; Bryant, B.; Duthie, R

    1993-01-01

    OBJECTIVE--To determine the psychiatric consequences of being a road traffic accident victim. DESIGN--Follow up study of road accident victims for up to one year. SETTING--Emergency department of the John Radcliffe Hospital, Oxford. SUBJECTS--188 consecutive road accident victims aged 18-70 with multiple injuries (motorcycle or car) or whiplash neck injury, who had not been unconscious for more than 15 minutes, and who lived in the catchment area. MAIN OUTCOME MEASURES--Present state examinat...

  10. Consequences of the Chernobyl accident

    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

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

    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.)

  12. Assessing economic consequences of radiation accidents

    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

  13. Applications of probabilistic accident consequence evaluation in Cuba

    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

  14. Aerosols behavior inside a PWR during an accident

    During very hypothetical accidents occurring in a pressurized water ractor, radioactive aerosols can be released, during core-melt, inside the reactor containment building. A good knowledge of their behavior in the humid containment atmosphere (mass concentration and size distribution) is essential in order to evaluate their harmfulness in case of environment contamination and to design possible filtration devices. Accordingly the Safety Analysis Department of the Atomic Energy Commission uses several computer models, describing the particle formation (BOIL/MARCH), then behavior in the primary circuits (TRAP-MELT), and in the reactor containment building (AEROSOLS-PARFDISEKO-III B). On the one hand, these models have been improved, in particular the one related to the aerosol formation (nature and mass of released particles) using recent experimental results. On the other hand, sensitivity analyses have been performed with the AEROSOLS code which emphasize the particle coagulation parameters: agglomerate shape factors and collision efficiency. Finally, the different computer models have been applied to the study of aerosol behavior during a 900 MWe PWR accident: loss-of-coolant-accident (small break with failure of all safety systems)

  15. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Klaus Umminger

    2012-01-01

    Full Text Available Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs arranged symmetrically around the reactor pressure vessel (RPV. The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

  16. Medical consequences of Chernobyl accident

    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.

  17. Behaviour of organic iodides under pwr accident conditions

    Laboratory experiments were performed to study the behaviour of radioactive methyl iodide under PWR loss-of-coolant conditions. The pressure relief equipment consisted of an autoclave for simulating the primary circuit and of an expansion vessel for simulating the conditions after a rupture in the reactor coolant system. After pressure relief, the composition of the CH3sup(127/131)I-containing steam-air mixture within the expansion vessel was analysed at 80 0C over a period of 42 days. On the basis of the values measured and of data taken from the literature, both qualitative and quantitative assessments have been made as to the behaviour of radioactive methyl iodide in the event of loss-of-coolant accidents. (author)

  18. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

  19. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  20. The Chernobyl nuclear accident and its consequences

    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

  1. Health consequences of nuclear accidents

    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

  2. Biological and medical consequences of nuclear accidents

    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)

  3. Numerical simulation of radioisotope's dependency on containment performance for large dry PWR containment under severe accidents

    Highlights: • Calculation and comparison of activity of BURN-UP code with ORIGEN2 code. • Development of SASTC computer code. • Radioisotopes dependency on containment ESFs. • Mitigation in atmospheric release with ESFs operation. • Variation in radioisotopes source term with spray flow and pH value. -- Abstract: During the core melt accidents large amount of fission products can be released into the containment building. These fission products escape into the environment to contribute in accident source term. The mitigation in environmental release is demanded for such radiological consequences. Thus, countermeasures to source term, mitigations of release of radioactivity have been studied for 1000 MWe PWR reactor. The procedure of study is divided into five steps: (1) calculation and verification of core inventory, evaluated by BURN-UP code, (2) containment modeling based on radioactivity removal factors, (3) selection of potential accidents initiates the severe accident, (4) calculation of release of radioactivity, (5) study the dependency of release of radioactivity on containment engineering safety features (ESFs) inducing mitigation. Loss of coolant accident (LOCA), small break LOCA and flow blockage accidents (FBA) are selected as initiating accidents. The mitigation effect of ESFs on source term has been studied against ESFs performance. Parametric study of release of radioactivity has been carried out by modeling and simulating the containment parameters in MATLAB, which takes BURN-UP outcomes as input along with the probabilistic data. The dependency of iodine and aerosol source term on boric and caustic acid spray has been determined. The variation in source term mitigation with the variation of containment spray flow rate and pH values have been studied. The variation in containment retention factor (CRF) has also been studied with the ESF performance. A rapid decrease in source term is observed with the increase in pH value

  4. Radiological consequences of the Chernobyl reactor accident

    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)

  5. Consequences in Guatemala of the Chernobyl accident

    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

  6. Consequences in Sweden of the Chernobyl accident

    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)

  7. Assessment of severe accident prevention and mitigation features: PWR, large dry containment design

    Plant features and operator actions which have been found to be important in either preventing or mitigating severe accidents in PWRs with large dry containments have been identified. These features and actions were developed from insights derived from reviews of risk assessments performed specifically for the Zion plant and from assessments of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the large dry containment to severe accident containment loads were also identified. In addition, those features of a PWR with a large dry containment, which are important for preventing core damage and are available for mitigating fission-product release to the environment were identified. The report is issued to provide focus to the analyst examining an individual plant. The report calls attention to plant features and operator actions and provides a list of deterministic tributes for assessing those features and actions found to be helpful in reducing the overall risk for Zion and other PWRs with large dry containments. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

  8. Assessment of the efficiency of short term countermeasures following a severe accident on a PWR

    In case of a severe nuclear accident at a PWR plant, countermeasures will be initiated in the short term by authorities to reduce the consequences of the atmospheric radioactive releases on the neighbouring population. Various factors influence the level of protection afforded by countermeasures. For instance, a too late intervention would lead to a Jack of efficiency in terms of dose reduction if the actual evolution of the accident is not considered. Thus, implementation of countermeasures should be optimized. In general, the projected doses (those without countermeasure) are compared with those expected when a particular countermeasure or strategy is implemented. In this paper, an in-depth analysis associates the kinetics of the release with the corresponding evolution of the dosimetric efficiency of countermeasures. This is done at different times in the short term of the accident and for various distances from the plant. Results are presented for different strategies initiated at various times. This work gives useful information for the early management of a major nuclear accident. (authors)

  9. Real and mythical consequences of Chernobyl accident

    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 The 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

  10. Health consequences [of the Chernobyl accident

    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)

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

    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.)

  12. Consequences and problems of the Chernobyl accident

    The data on epidemic situation in connection with the Chernobyl accident, based on the personal medical and dosimetric information on all the persons, subjected to radiation effect, and included in the Russian state medicodosimetric register, are presented. The consequences of the Chernobyl accident become the cause for origination of serious radiation injures by 134 persons (with lethal outcome by 37 patients) and also remote radiation stochastic effects by children (thyroid gland cancer) and by liquidators (thyroid gland leucosis and cancer). The permanent stress and other unfavorable factors conditioned aggravation of chronical and increase in somatic diseases and psychoneurotic disorders

  13. Environmental consequences of releases from nuclear accidents

    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

  14. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Andrej Prošek

    2013-01-01

    Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

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

    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)

  16. Essential severe accident mitigation measures for operating and future PWR's

    Severe Accident mitigation measures are a constituent of the safety concept in Europe not only for operating but also for future light water reactors. While operating reactors mainly have been backfitted with such measure, for future reactors Severe Accident mitigation measures already have to be considered in the design phase. Severe Accident measures are considered as the 4th level of defense for future reactors. This difference has consequences also on the kind of measures proposed to be introduced. While in operating plants Severe Accident mitigation measures are considered for further risk reduction, in future reactors an explicit higher level of safety is required resulting in additional design measures. This higher safety level is expressed in the requirement that there must be no need for evacuation of surrounding populations except in the immediate vicinity of the plant and for long-term restrictions with regard to the consumption of locally grown food. Because of the potential hazard posed by radioactive releases to the environment in the event of an Severe Accident situation depends largely on the airborne material in the containment atmosphere and on the containment integrity, new system features to prevent loss of containment integrity have been introduced in the design of the NPP's. For these tasks it has been necessary to develop and qualify new system technologies and implement them finally into NPP's, e.g. like systems for containment atmosphere H2-control, filtered venting, core retention devices and atmosphere sampling. The following systems are introduced for operating as well as for future plants: · The Hydrogen Control System is based on the Passive Autocatalytic Recombiner (PAR) technology. There is no need for any operator actions because of the self-starting feature of the catalyst if hydrogen is released. · In situ Post Accident Sampling System (In situ-PASS) are introduced for the purpose of obtaining realistic information on airborne

  17. Experiments on natural circulation during PWR severe accidents and their analysis

    Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occurring in the primary coolant system (CS) before substantial core degradation. Natural circulation flow patterns were measured in a one-seventh scale PWR PCS facility at Westinghouse RandD laboratories. The measured flow and temperature distributions are report in this paper. The experiments were analyzed with the COMMIX code and good agreement was obtained between data and calculations. 10 refs., 8 figs., 2 tabs

  18. Radiological consequences of the Chernobyl reactor accident

    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

  19. Remarks on methods of evaluation of aerosol sources related to PWR core meltdown accidents

    The paper tries to demonstrate the conceptional background of the KfK core melting program, which has been started in 1973, and which is scheduled to be terminated by 1986. The paper also summarizes the main findings of the SASCHA program, with the aid of which the enveloping fission product release from the primary system into the containment during a PWR core melt accident has been investigated. The fractions of release from the fuel determined in the experiment are undoubtedly in the range of 70% to 100% for the radiologically most important elements I, Cs, Te. The reduction in release from the primary circuit due to deposition is 50% at the maximum. A considerable portion resuspended must be deducted from that value. The retention of iodine and aerosol particles in the safety containment amounts to several orders of magnitude (up to 5). Likewise, the decrease in the population dose by spread and dilution in the environment and due to other parameters attains several orders of magnitude (up to 7). Consequently, particle retention by a factor of 2 or 3 in the primary circuit is negligible. - Our present knowledge is completely satisfactory for analyzing the so-called source term in core melt accidents. The wish to develop more detailed codes related to core degradation and to activity release from the primary circuit has many understandable causes. However, there is no single technical reason in favor of spending much money in order to materialize this wish. (orig./HP)

  20. The consequences of the Chernobyl reactor accident

    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)

  1. Cosyma a new programme package for accident consequence assessment

    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

  2. Rupther: a simulation approach applied to a PWR vessel failure during a severe accident

    The Rupther program (Rupture Under Thermal Conditions) is a part of the international researches on severe accidents in the PWR type reactors. The aim of the program is the definition of failure simulation validated by experimental data on vessel steel 16MND5 mechanical properties. The paper presents the program and the first results. (A.L.B.)

  3. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    DIAMOND,D.J.; ARONSON,A.; JO,J.; AVVAKUMOV,A.; MALOFEEV,V.; SIDOROV,V.; FERRARESI,P.; GOUIN,C.; ANIEL,S.; ROYER,M.E.

    1999-10-01

    This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.

  4. A general approach to critical infrastructure accident consequences analysis

    Bogalecka, Magda; Kołowrocki, Krzysztof; Soszyńska-Budny, Joanna

    2016-06-01

    The probabilistic general model of critical infrastructure accident consequences including the process of the models of initiating events generated by its accident, the process of environment threats and the process of environment degradation is presented.

  5. Environmental consequences of releases from nuclear accidents

    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. Development of evaluation method for economic consequences of severe accident at NPP

    A nuclear power plant accident has onsite and/or offsite consequences. A common framework into which many of the consequences of an accident may be translated is their economic cost. Though there are some consequence analysis codes, these are limited only to estimate only offsite consequence. In this study, economic losses are estimated for the case of onsite and offsite consequences for the sample PWR plant at Yong-Gwang site. For the estimation of offsite consequence, economic database unique to Korean economic structure is utilized as many as possible. By grouping of various cost components, each cost groups are compared each other. For the detailed estimation of offsite decontamination cost, offsite surface around the plant is divided into five types of surface. This division of surface types is agricultural field, wooded land, roof surface for representing house, asphalt road and water surface. But on the water surface, it is assumed that no decontamination operation is required. The comparison shows that at the severe accident, offsite consequences are more severe than onsite consequences. And in agricultural growing season, the consequence even becomes more severe. It also turned out that the importance and selection of much appropriate criteria will play a major role, because the economic consequences are widely varying depending on how the criteria was chosen

  7. Simulation of a low-pressure severe accident scenario in a PWR with ATHLET-CD

    Hoffmann, Mathias; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2013-07-01

    The plant behavior of a Pressurized Water Reactor (PWR) during a severe accident scenario is analyzed with system code ATHLET-CD Mod. 2.2C in order to assess the code capabilities in terms of the late-phase of the core degradation. For this purpose a severe accident sequence caused by a Station Black-out and a large break in the primary cooling system is simulated both without any accident management measures and with a delayed reflooding of the substantially degraded core. Selected code results are presented in this paper. (orig.)

  8. A knowledge based severe accident handbook for PWR

    During the last decade the level of knowledge about severe accident phenomena has increased dramatically. The improved understanding has been achieved by extensive research but also from feed-back of experience from actual incidents/accidents such as Three Mile Island and Chernobyl. In Sweden, mitigating measures such as filtered venting and external water source were implemented at all nuclear power plants by 1988. In parallel the Emergency Operating Procedures (at Ringhals called Emergency Response Guidelines, ERG, and Beyond ERG, BERG) were developed to include these new features. However, the accident management system has since then been further improved and one important aspect is the long-term accident management. The new information obtained has been one of the basis for a new knowledge based handbook to support the unit leader and the Technical Support Center. The handbook contains information concerning specific issues in the BERG and advice how the organization can manage a long-term severe accident situation

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

    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.)

  10. International experience with a multidisciplinary table top exercise for response to a PWR accident

    Table Top Exercises are used for the training of emergency response personnel from a wide range of disciplines whose duties range from strategic to tactical, from managerial to operational. The exercise reported in this paper simulates the first two or three hours of an imaginary accident on a generic PWR site (named Seaside or Lakeside depending on its location). It is designed to exercise the early response of staff of the utility, government, local authority and the media and some players represent the public. The relatively few scenarios used for this exercise are based on actual events scaled to give off-site consequences which demand early assessment and therefore stress the communication procedures. The exercise is applicable in different cultures and has been used in over 20 short courses held in the USA, UK, Sweden, Prague, and Hong Kong. There are two styles of support for players: a linear program which ensures that all players follow the desired path through the event and an open program which is triggered by umpires (who play the reactor crew from a script) and by requests from other players. In both cases the exercise ends with a Press Conference. Players have an initial briefing and are assigned to roles; those who must speak at interviews and at the Press Conference arc given separate briefing by an expert in Public Affairs. The exercise runs with up to six groups and the communication rate reaches about 30 to 40 messages per hour for each group. The exercise can be applied to test management and communication systems and to study human response to emergencies because the merits of individual players are highlighted in the relatively stressful conditions of the initial stage of an accident. For some players the exercise is the first time that they have been required to carry out their task in front of other people

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

    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. Probabilistic consequence analysis of ATWSs in a PWR plant

    PWR responses (in terms of overpressures, DNBR and other safety-related quantities) to ATWSs are being probabilistically investigated by applying response surface methodology to ALMOD, a computer program for simulation of large amplitude transients. The reactor considered for the analysis is the 1300 MWel reference KWU reactor plant. A comprehensive set of input quantities--including operational, engineering and physical variables--is taken into account. Results are presented for the first phases of station-blackout and loss-of-heat sink ATWSs

  13. Consequences of the Fukushima accident: A preliminary assessment and discussion

    Tsunami due to the earthquake in East Japan Sea eventually leaded to a severe nuclear accident in Fukushima Dai-ichi nuclear power plant. This event immediately became the focus of the whole world. The work to roughly evaluate and predict the consequence of this nuclear accident is summarized in this paper and the work actually provides valuable information in predicting the scale and severity of the accident comparing to the published information on the accident thereafter. (authors)

  14. Calculation of dose consequences of a hypothetical large accident at a nuclear power reactor

    The fission product release to the atmosphere from a nuclear reactor during a hypothetical large accident is discussed, and the consequences in terms of doses to the population are calculated. The reactor is a light-water reactor located at a site representing an idealized, simplified Danish location. Three release categories are discussed: The first is the release in an accident with a core meltdown and a major rupture of the containment. This release is represented by the BWR-2 release of WASH-1400. The second is the release where the containment integrity is maintained, but there is a failure to isolate the containment. This release is represented by the PWR-4 release. The third is obtained as a Best Estimate from Empirical Data, and it is called the BEED release. The release of the BEED case is deduced from empirical evidence - especially the SL-1 accident - in a seperate study which is described in the appendix. The release fractions for the most significant elements such as iodine and cesium decrease by a decade from BWR-2 to PWR-4, and from PWR-4 to BEED, while the release fractions of the noble gases are assumed to be at an almost constant high level. The dose consequences in terms of a long-term committed effective dose equivalent is found to be practically directly proportional to the release fractions, i.e. decreasing by a decade from BWR-2 to PWR-4, and from PWR-4 to BEED. For the acute bone marrow dose the contribution from the noble gases is significant in all three cases, and as the noble gas release is almost constant the decrease from one case to the next is of the order of only half a decade. For the BEED case the noble gases, which give an external gamma dose from the plume, are the most significant radioactive fission products, both for long-term and acute doses. For the BWR-2 and PWR-4 cases the I-131 in the plume is predominant in the acute dose, while Cs-137 deposited on the ground is the main contributor to the long-term dose. (author)

  15. Medical consequences of a nuclear plant accident

    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. Reassessment of PWR pressure-vessel integrity during overcooling accidents

    A continuing analysis of the PTS problem associated with PWR postuated OCA's indicates that the previously accepted degree of conservatism in the fracture-mechanics model needs to be more closely evaluated, and if excessive, reducted. One feature that was believed to be conservative was the use of two-dimensional as opposed to finite-length (three-dimensional) flaws. A flaw of particular interest is one that is located in an axial weld of a plate-type vessel. For those vessels that suffer relatively high radiation damage in the welds, the length of the flaw will be no greater than the length of the weld, and recent calculations indicate that a deep flaw of that length (approx. 2 m) is not effectively infinitely long, contrary to previous thinking. The benefit to be derived from consideration of the 2-m flaw and also a semielliptical flaw with a length-to-depth ratio of 6/1 was investigated by analyzing several postulated transients. In doing so the sensitivity of the benefit to a specified maximum crack arrest toughness and to the duration of the transient was investigated. Results of the analysis indicate that for some conditions the benefit in using the 2-m flaw is substantial, but it decreases with increasing pressure, and above a certain pressure there may be no benefit, depending on the duration of the transient and the limit on crack arrest toughness

  17. Accident management following loss-of-coolant accidents during cooldown in a Westinghouse two-loop PWR

    Operation of pressurised water reactors involves shutdown periods for refuelling and maintenance. In preparation for this, the reactor system is cooled down, depressurised and partially drained. Although reactor coolant pressure is lower than during full-power operation, there remains the possibility of a loss-of-coolant accident (LOCA), with a certain but low probability. While the decay heat to be removed is lower than that from a LOCA at full power, the reduced availability of safety systems implies a risk of failing to maintain core cooling, and hence of core damage. This is recognised though probabilistic safety analyses (PSA), which identify low but non-negligible contributions to core damage frequency from accidents during cooldown and shutdown. Analyses are made for a typical two-loop Westinghouse PWR of the consequences of a range of LOCAs during hot and intermediate shutdown, 4 and 5 h after reactor shutdown respectively. The accumulators are isolated, while power to some of the pumped safety injection systems (SIs) is racked out. The study assesses the effectiveness of the nominally assumed SIs in restoring coolant inventory and preventing core damage, and the margin against core damage where their actuation is delayed. The calculations use the engineering-level MELCOR1.8.5 code, supplemented by the SCDAPSIM and SCDAP/RELAP5 codes, which provide a more detailed treatment of coolant system thermal hydraulics and core behaviour. Both treatments show that the core is readily quenched, without damage, by the nominal SI which assumes operation of only one pump. Margins against additional scenario and model uncertainties are assessed by assuming a delay of 900 s (the time needed to actuate the remaining pumps) and a variety of assumptions regarding models and the number of pumps available in conjunction with both MELCOR and versions of SCDAP. Overall, the study provides confidence in the inherent robustness of the plant design with respect to LOCA during

  18. Accident management following loss-of-coolant accidents during cooldown in a Westinghouse two-loop PWR

    Haste, T.J., E-mail: tim.haste@irsn.f [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Birchley, J. [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Richner, M. [Nordostschweizerische Kraftwerke (NOK) - NPP Beznau, CH-5312 Doettingen (Switzerland)

    2010-06-15

    Operation of pressurised water reactors involves shutdown periods for refuelling and maintenance. In preparation for this, the reactor system is cooled down, depressurised and partially drained. Although reactor coolant pressure is lower than during full-power operation, there remains the possibility of a loss-of-coolant accident (LOCA), with a certain but low probability. While the decay heat to be removed is lower than that from a LOCA at full power, the reduced availability of safety systems implies a risk of failing to maintain core cooling, and hence of core damage. This is recognised though probabilistic safety analyses (PSA), which identify low but non-negligible contributions to core damage frequency from accidents during cooldown and shutdown. Analyses are made for a typical two-loop Westinghouse PWR of the consequences of a range of LOCAs during hot and intermediate shutdown, 4 and 5 h after reactor shutdown respectively. The accumulators are isolated, while power to some of the pumped safety injection systems (SIs) is racked out. The study assesses the effectiveness of the nominally assumed SIs in restoring coolant inventory and preventing core damage, and the margin against core damage where their actuation is delayed. The calculations use the engineering-level MELCOR1.8.5 code, supplemented by the SCDAPSIM and SCDAP/RELAP5 codes, which provide a more detailed treatment of coolant system thermal hydraulics and core behaviour. Both treatments show that the core is readily quenched, without damage, by the nominal SI which assumes operation of only one pump. Margins against additional scenario and model uncertainties are assessed by assuming a delay of 900 s (the time needed to actuate the remaining pumps) and a variety of assumptions regarding models and the number of pumps available in conjunction with both MELCOR and versions of SCDAP. Overall, the study provides confidence in the inherent robustness of the plant design with respect to LOCA during

  19. Fast detections of the accident. Radiological consequences

    This paper shows how the contamination due to the accident of Chernobylsk has been discovered in Sweden. The Swedish national Institute of radio-protection describes in detail the measurements done, and the decisions of radioprotection which have been taken

  20. Basic study on PWR plant behavior under the condition of severe accident

    In this paper, we report on the core cooling effect by natural circulation cooling of the primary cooling system in all core cooling function loss accidents caused by SBO in PWR plant compared with BWR. We also report on the core cooling effect by using air as the final heat sink in place of the seawater by opening the main steam valve of the steam generator. On the other hand, we discuss the behavior of PWR plant in the most serious case that the damage such as LOCA is caused by earthquake and that SBO due to the subsequent tsunami causes the reactor isolation and all function of reactor core cooling system loss. That is the case that LOCA and SBO occur in superimposed manner. We can show the results from the simulation experiments that, in PWR plant, even if it is fell into the reactor core cooling function loss due to SBO, natural circulation cooling can keep the reactor core cool down as long as the feed water is supplied to SG by the turbine-driven auxiliary feed-water pump and also that the cooling effect of even more is expected by ensuring the heat-pass to the atmosphere by opening the main steam valve. We also clarify the plant behaviors under the condition that LOCA and SBO occur in superimposed manner in PWR through the simulation experiments. (author)

  1. Source term and radiological consequences of the Chernobyl accident

    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

  2. The consequences of the Chernobyl nuclear accident in Greece

    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.)

  3. Assessment and limitation of radioactivity transfers in the event of a postulated severe PWR accident

    This report constitutes the supporting material for a lecture on severe accidents which could occur on PWR type nuclear reactors. It is assumed for present purposes that the reader has at least a rudimentary acquaintance with the basics of general physics if not with the operating processes of these reactors. After defining what is meant by a ''severe accident'' on a reactor, the possible phenomenology of such an accident is qualitatively described: loss of coolant and loss of containment integrity. A certain number of elements are then given for the quantitative assessment of these phenomena involving possible radioactivity transfers within and outside the plant. In conclusion, available means are indicated for the limitation and control of these environmental transfers. (author). 5 refs, figs

  4. Consequences of the Chernobyl accident in Lithuania

    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

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

    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(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)

  6. Analysis of a control rod ejection accident in a 900 MWe PWR recycling plutonium with a gray control mode

    This research thesis addresses the study of the control rod cluster ejection accident in a 900 MWe PWR recycling plutonium and operating in grey mode, a class-IV accident in the safety report, which results from the failure of the cluster mechanism pressure enclosure, and results in a quick introduction of a reactivity within the core, and then in a violent power transient during which fuel strength can be put into question again. Two aspects are thus notably addressed: plutonium recycling, and grey mode operation. The objective is to qualitatively and quantitatively assess the evolution of physical parameters during the accident in order to determine the most severe scenarios and to be able to assess the severity of consequences. The author first studies all possible scenarios by means of a 2D+1D+0D calculation scheme in order to determine the most penalizing ones. Then, he develops a precise calculation based on 3D steady calculations, neutron kinetics calculations and thermal kinetics calculations in order to study the previously retained scenarios

  7. Method for consequence calculations for severe accidents

    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)

  8. Appearing consequences of the Chernobyl accident

    Full text: Chernobyl is the greatest world's tragedy after Chirosima. Global results of this tragedy is already being seen. They are the people who have received radiation dose. the first type of cancer 5 years after Chernobyl accident was the thyroid gland cancer, the reason of it, large quantities of radioactive iodine in the air, food products, milk of cattle and finally their collection in the thyroid gland cancer entering the human body. Period all of a sudden after 10 years completed the next latent type of cancer was leykoz. Giving rise to this type of cancer more sensitive to radiation of the body - a violation of the spinal brain function. After 20 years passing from the accident in the first generation one ill child must be born cause of undergoing to radiation father or mother from each three days in Belarus, Russia and Ukraine

  9. Medical demographic consequences of the Chernobyl accident

    A demographic study was made of the population evacuated from the 30-km zone around the nuclear power plant and of the population living in areas over which the radioactive cloud passed and over which the plume was formed. For the farmers evacuated from 11,655 homes in the Chernobyl region, 7,000 new houses, built in the Kiev region, had already been provided within 5 months of the accident, and by the summer of 1987 another 5,000 houses were available. A study of the resettlement of the population carried out a year after the accident showed that more than 60% of those evacuated continued to live in the regions from which the evacuation had taken place; about 5% were resettled in other republics, and 20% within their own republic. (author). 7 figs, 2 tabs

  10. Phenomenology and Course of Severe Accidents in PWR-Plants - Training by Teaching and Demonstration

    A special one day training course on 'Phenomenology and Course of Severe Accidents in PWR-Plants' was developed at GRS initiated by the interest of German utilities. The work was done in the frame of projects sponsored by the German Ministries for Environment, Nature Conservation and Nuclear Safety (BMU) and for Education, Science, Research and Technology (BMBF). In the paper the intention and the subject of this training course will be discussed and selected parts of the training course will be presented. Demonstrations are made within this training course with the GRS simulator system ATLAS to achieve a broader understanding of the phenomena discussed and the propagation of severe accidents on a plant specific basis. The GRS simulator system ATLAS is linked in this case to the integral code MELCOR and pre-calculated plant specific severe accident calculations are used for the demonstration together with special graphics showing plant specific details. Several training courses have been held since the first one in November, 1996 especially to operators, shift personal and the management board of a German PWR. In the meantime the training course was updated and suggestions for improvements from the participants were included. In the future this type of the training course will be made available for members of crisis teams, instructors of commercial training centres and researchers of different institutions too. (authors)

  11. Phenomenology and course of severe accidents in PWR-plants training by teaching and demonstration

    A special one day training course on 'Phenomenology and Course of Severe Accidents in PWR-Plants' was developed at GRS initiated by the interest of German utilities. The work was done in the frame of projects sponsored by the German Ministries for Environment, Nature Conservation and Nuclear Safety (BMW) and for Education, Science, Research and Technology (BMBF). In the paper the intention and the subject of this training course are discussed and selected parts of the training course are presented. Demonstrations are made within this training course with the GRS simulator system ATLAS to achieve a broader understanding of the phenomena discussed and the propagation of severe accidents on a plant specific basis. The GRS simulator system ATLAS is linked in this case to the integral code MELCOR and pre-calculated plant specific severe accident calculations are used for the demonstration together with special graphics showing plant specific details. Several training courses have been held since the first one in November, 1996 especially to operators, shift personal and the management board of a German PWR. In the meantime the training course was updated and suggestions for improvements from the participants were included. In the future this training course will be made available for members of crisis teams, instructors of commercial training centres and researchers of different institutions too. (author)

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

    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)

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

    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

  14. Updating and testing of a PWR model for the Modular Accident Analysis Programe MAAP5

    Marcos Delgado, Elisabet

    2013-01-01

    The present Master’s Thesis is part of the Master’s degree in Nuclear Engineering of the Universitat Politècnica de Catalunya and the ENDESA Escuela de Energía, and it was developed during the internship in a Spanish Pressurized Water Reactor (PWR). The objective of the project is to update and test the nuclear plant model used for the Safety Analysis department which belongs to the Licensing Department mainly for Severe Accidents phenomenology studies to prepare for and respond to emergen...

  15. Modelling of whole-core release of fission products in PWR core melt accidents: Chapter 13

    The computer code FISREL combines the thermal history of a reactor core with experimentally-based release rate constants to calculate whole-core release histories of fission products in PWR core melt accidents. Predictions of the code for releases of volatile fission products during large-break, small-break and transient initiated sequences are presented, and the sensitivities of results to input data examined. A preliminary assessment of the limitations imposed by mass transport on release of vaporized materials in high pressure sequences is given, and the implications of the results for primary system transport are discussed

  16. Thyroid consequences of the Chernobyl nuclear accident.

    Pacini, F; Vorontsova, T; Molinaro, E; Shavrova, E; Agate, L; Kuchinskaya, E; Elisei, R; Demidchik, E P; Pinchera, A

    1999-12-01

    It is well recognized that the use of external irradiation of the head and neck to treat patients with various non-thyroid disorders increases their risk of developing papillary thyroid carcinoma years after radiation exposure. An increased risk of thyroid cancer has also been reported in survivors of the atomic bombs in Japan, as well as in Marshall Island residents exposed to radiation during the testing of hydrogen bombs. More recently, exposure to radioactive fallout as a result of the Chernobyl nuclear reactor accident has clearly caused an enormous increase in the incidence of childhood thyroid carcinoma in Belarus, Ukraine, and, to a lesser extent, in the Russian Federation, starting in 1990. When clinical and epidemiological features of thyroid carcinomas diagnosed in Belarus after the Chernobyl accident are compared with those of naturally occurring thyroid carcinomas in patients of the same age group in Italy and France, it becomes apparent that the post-Chernobyl thyroid carcinomas were much less influenced by gender, virtually always papillary (solid and follicular variants), more aggressive at presentation and more frequently associated with thyroid autoimmunity. Gene mutations involving the RET proto-oncogene, and less frequently TRK, have been shown to be causative events specific for papillary cancer. RET activation was found in nearly 70% of the patients who developed papillary thyroid carcinomas following the Chernobyl accident. In addition to thyroid cancer, radiation-induced thyroid diseases include benign thyroid nodules, hypothyroidism and autoimmune thyroiditis, with or without thyroid insufficiency, as observed in populations after environmental exposure to radioisotopes of iodine and in the survivors of atomic bomb explosions. On this basis, the authors evaluated thyroid autoimmune phenomena in normal children exposed to radiation after the Chernobyl accident. The results demonstrated an increased prevalence of circulating thyroid

  17. ALIBABA: a French Expert System for PWR Containment Analysis in case of Severe Accidents

    In the event of an accident occurring in a French pressurized water reactor (PWR), the authorities should be in position to implement the measures required to protect the surrounding population and the environment from radiological consequences of potential releases. The Institute for Nuclear Safety and Protection is part of the national emergency organization established for this purpose. It provides technical support to the French nuclear safety authority. As a technical adviser, IPSN has defined a methodology intended to help assess the plant status and monitor its development as soon as the accident is detected. On the basis of this assessment, the method forecasts the potential behavior of the installation and estimates the related consequences. The state of the installation is evaluated throughout the accident with special reference to the three barriers stretched out between the radioisotopes and the environment (fuel cladding, reactor coolant system and the containment building). It considers successively their physical state, the state of the safety functions guaranteeing their integrity and finally the state of the systems available to monitor these functions. In order to properly diagnose and predict the state of the barriers, evaluations are necessary to quantify parameters such as the break size on the reactor coolant system, the time to core uncovering and the core degradation. As a result, fission products behavior inside the installation and releases outside the plant are assessed. Several flexible, rapid and user-friendly software tools, which are part of the French SESAME system, have been developed to help the experts with their assessment. products cannot be realistically quantified without a complete knowledge of the state of the containment barrier. The expert system ALIBABA is separate from these tools. It provides complementary qualitative information about the third barrier. Indeed, the ongoing or potential releases of fission products

  18. Simulation of fission products behavior in severe accidents for advanced passive PWR

    Highlights: • A fission product analysis model based on thermal hydraulic module is developed. • An assessment method for fission product release and transport is constructed. • Fission products behavior during three modes of containment response is investigated. • Source term results for the three modes of containment response are obtained. - Abstract: Fission product behavior for common Pressurized Water Reactor (PWR) has been studied for many years, and some analytical tools have developed. However, studies specifically on the behavior of fission products related to advanced passive PWR is scarce. In the current study, design characteristics of advanced passive PWR influencing fission product behavior are investigated. An integrated fission products analysis model based on a thermal hydraulic module is developed, and the assessment method for fission products release and transport for advanced passive PWR is constructed. Three modes of containment response are simulated, including intact containment, containment bypass and containment overpressure failure. Fission products release from the core and corium, fission products transport and deposition in the Reactor Coolant System (RCS), fission products transport and deposition in the containment considering fission products retention in the in-containment refueling water storage tank (IRWST) and in the secondary side of steam generators (SGs) are simulated. Source term results of intact containment, containment bypass and containment overpressure failure are obtained, which can be utilized to evaluate the radiological consequences

  19. Radiological attacks and accidents. Medical consequences

    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)

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

    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

  1. Brief account of the effect of overcooling accidents on the integrity of PWR pressure vessels

    The occurrence in recent years of several (PWR) accident initiating events that could lead to severe thermal shock to the reactor pressure vessel, and the growing awareness that copper and nickel in the vessel material significantly enhance radiation damage in the vessel, have resulted in a reevaluation of pressure-vessel integrity during postulated overcooling accidents. Analyses indicate that the accidents of concern are those involving both thermal shock and pressure loadings, and that an accident similar to that at Rancho Seco in 1978 could, under some circumstances and at a time late in the normal life of the vessel, result in propagation of preexistent flaws in the vessel wall to the extent that they might completely penetrate the wall. More severe accidents have been postulated that would result in even shorter permissible lifetimes. However, the state-of-the-art fracture-mechanics analysis may contain excessive conservatism, and this possibility is being investigated. Furthermore, there are several remedial measures, such as fuel shuffling, to reduce the damage rate, and vessel annealing, to restore favorable material properties, that may be practical and used if necessary. 5 figures

  2. Method for consequence calculations for severe accidents

    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)

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

    The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

  4. Consequence of potential accidents in heavy water plants

    Heavy water plants realize the primary isotopic concentrations of water using H2O-H2S 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 H2S 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, H2S 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)

  5. Offsite Radiological Consequence Analysis for the Bounding Flammable Gas Accident

    Carro, C A

    2003-01-01

    This document quantifies the offsite radiological consequences of the bounding flammable gas accident for comparison with the 25 rem Evaluation Guideline established in DOE-STD-3009, Appendix A. The bounding flammable gas accident is a detonation in a single-shell tank The calculation applies reasonably conservation input parameters in accordance with DOE-STD-3009, Appendix A, guidance. Revision 1 incorporates comments received from Office of River Protection.

  6. MELCOR Accident Consequence Code System (MACCS)

    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

  7. MELCOR Accident Consequence Code System (MACCS)

    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

  8. MELCOR Accident Consequence Code System (MACCS)

    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.

  9. MELCOR Accident Consequence Code System (MACCS)

    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.

  10. Consequences of the Chernobyl accident in Styria

    We present results which document the contamination of Styria (Southern part of Austria) immediately after and in the years following the Chernobyl accident. The radioactivity and distribution of radionuclides in aerosols, rain water, soil, vegetation, animals and various samples of food are described in great detail. One of the key results is that the highest levels of contamination were found in two districts (Liezen, Deutschlandsberg), and the deposition rates for Cs-137 were determined to be in the range from 3 to about 80 kBq/m2. Of particular interest are studies concerning the migration and distribution of radionuclides in soil, the uptake of radiocesium by the aquatic vegetation and the existence of radionuclides in the natural ecosystem up to this day. Effective dose equivalents due to incorporated radiocesium was estimated to be 252.2 μSv for the adult population of Graz (capital of Styria) over the four years follwing the fallout. (authors) 17 papers are presented and are of INIS scope

  11. MELCOR Accident Consequence Code System (MACCS)

    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

  12. Radioecological and dosimetric consequences of Chernobyl accident in France

    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.)

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

    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)

  14. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3·4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident

  15. Development of a parametric containment event tree model of a severe PWR accident

    The study supports the development project of STUK on 'Living' PSA Level 2. The main work objective is to develop review tools for the Level 2 PSA studies underway at the utilities. The SPSA (STUK PSA) code is specifically designed for the purpose. In this work, SPSA is utilized as the Level 2 programming and calculation tool. A containment event tree (CET) model is built for analysis of severe accidents at the Loviisa pressurized water reactor (PWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to include new research results, and so it facilitates the Living PSA concept on Level 2 as well. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting at a low primary system pressure. Severe accident progression from five plant damage states (PDSs) is examined, however the integration with Level 1 is deferred to more definitive, integrated, safety assessments. (34 refs., 5 figs., 9 tabs.)

  16. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong [Seoul Nat' l Univ., Seoul (Korea, Republic of)

    2001-03-15

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3{center_dot}4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident.

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

    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)

  18. OFFSITE RADIOLOGICAL CONSEQUENCE ANALYSIS FOR THE BOUNDING FLAMMABLE GAS ACCIDENT

    This document quantifies the offsite radiological consequences of the bounding flammable gas accident for comparison with the 25 rem Evaluation Guideline established in DOE-STD-3009, Appendix A. The bounding flammable gas accident is a detonation in a SST. The calculation applies reasonably conservative input parameters in accordance with guidance in DOE-STD-3009, Appendix A. The purpose of this analysis is to calculate the offsite radiological consequence of the bounding flammable gas accident. DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', requires the formal quantification of a limited subset of accidents representing a complete set of bounding conditions. The results of these analyses are then evaluated to determine if they challenge the DOE-STD-3009-94, Appendix A, ''Evaluation Guideline,'' of 25 rem total effective dose equivalent in order to identify and evaluate safety-class structures, systems, and components. The bounding flammable gas accident is a detonation in a single-shell tank (SST). A detonation versus a deflagration was selected for analysis because the faster flame speed of a detonation can potentially result in a larger release of respirable material. A detonation in an SST versus a double-shell tank (DST) was selected as the bounding accident because the estimated respirable release masses are the same and because the doses per unit quantity of waste inhaled are greater for SSTs than for DSTs. Appendix A contains a DST analysis for comparison purposes

  19. The Chernobyl reactor accident - provisional results and consequences

    Those involved at present in the analysis and estimation of consequences of the Chernobyl reactor accident are in a dilemma: While a worried and uncertain Western German public is calling for information the Soviet Union was practicing a rigorously restrictive information policy. Both the severity of the reactor accident and the complexity of events do urgently require the acquisition and evaluation of facts which will provide the basis for an objective factual discussion of issues and possible measures. The paper abstracted is trying to assess the alleged causes of the accident and estimate possible consequences. However, all attempts of that kind are based but on incomplete and dubious information as of May 21st, 1986. (orig.)

  20. Cold leg condensation model for analyzing loss-of-coolant accident in PWR

    Liao, Jun, E-mail: liaoj@westinghouse.com; Frepoli, Cesare; Ohkawa, Katsuhiro

    2015-04-15

    Highlights: • Direct contact cold leg condensation model for full spectrum LOCA evaluation model. • The cold leg condensation model addresses both large break LOCA and small break LOCA. • The model is assessed against both large break and small break LOCA experiments. • Scalability of the cold leg condensation model to full scale PWR is discussed. - Abstract: Direct contact condensation in the cold leg of pressurized water reactor is an important phenomenon during a postulated loss-of-coolant accident. The amount of condensation in the cold legs impacts the thermal hydraulic behavior of the reactor coolant system and eventually the integration of reactor nuclear core. A cold leg condensation model was developed for the WCOBRA/TRAC-TF2 safety analysis code. The model correlated the COSI test data and addressed the scaling issues with respect to geometry, pressure, and steam and water flow rates expected during a typical PWR LOCA. The correlation was found to be in good agreement with separate effects and integral effects experimental data and implemented in the WCOBRA/TRAC-TF2 safety analysis code. The cold leg condensation model was assessed against various small break and large break LOCA separate effects tests such as COSI experiments, ROSA experiments and UPTF experiments. Those experiments cover a wide range of cold leg dimensions, system pressures, mass flow rates, and fluid properties. All the predicted condensation results match reasonably well with the experimental data. Scalability discussions on the diameter, flow area, length, superficial velocity, Reynolds number of both cold leg and SI line, and Froude number of SI line in the Westinghouse COSI test facility were provided. The distortion of the SI jet Reynolds number is moderate. The scaling analysis together with the validation matrix covering a wide range of cold leg diameter, SI flow rate and SI Reynolds number support the scalability of the developed cold leg condensation model to the full

  1. Remote medical consequences of Chernobyl NPP accident in Armenia

    In result of global radio-ecological disaster at the Chernobyl NPP in Armenia there has appeared a great 'risk group' of persons, who had participated in liquidation of the accident consequences. The results of medical observation of this cohort carried out in dynamics in Scientific Center of Radiation Medicine and Burns during 25 years are brought in the work

  2. Hanford Waste Tank Bump Accident and Consequence Analysis

    BRATZEL, D.R.

    2000-06-20

    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.

  3. Hanford Waste Tank Bump Accident and Consequence Analysis

    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

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

    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

  5. Consequences of Chernobyl accident in Europe

    ,000 'liquidators' ranged between 170 mSv in 1986 and 15 mSv in 1989. Among the >100,000 evacuees the average whole body dose prior to evacuation was 15 mSv. The average lifetime Chernobyl whole body doses in European countries outside the former Soviet Union range from 0.006 mSv in Portugal to 2.4 mSv in Bulgaria. In the Northern Hemisphere the average Chernobyl lifetime dose is 0.14 mSv, i.e. about 0.08% of the natural dose. The average global whole body dose of natural radiation during 70 years is about 170 mSv, and 700 mSv in typically high background areas. Epidemiological studies from Hiroshima and Nagasaki suggest that no increase in cancer mortality should be expected at a single whole body dose (in addition to natural background radiation) of <200 mSv, delivered during a fraction of a second. Doses of about 200 mSv accumulated over tens of years of exposure would be even less effective. Ten years after the Chernobyl catastrophe the total radiation death toll is 31 - 38 persons, among them 3 persons were the members of the public. The total expected number of thyroid cancer deaths is about 500. In Poland, a country closest to Chernobyl outside the former Soviet Union, during two days, starting on the second day after arrival of radioactive cloud, 18.5 million persons were administered a prophylactic dose of stable iodine in form of 'Lugol solution', to block the uptake of radioiodine by the thyroid. This caused a thyroid dose reduction by a factor of up to 5, without any intra-thyroid side effects. Economic loses related to necessary and unnecessary remedial measures are estimated to reach in Belarus between 1986 and 2015 US$ 191.7 billion, of which US$ 86.32 billion are costs of financial and other compensation ('privileges') for peoples living at contaminated regions. It is estimated that in Ukraine in regions where 'Chernobyl radiation dose' is less than 1 mSv/year about 1.73 million persons receives the 'privileges'. Psychosomatic consequences of radiophobia induced by

  6. Status on development and verification of reactivity initiated accident analysis code for PWR (NODAL3)

    A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the nodal few-group neutron diffusion theory in 3-dimensional Cartesian geometry for a typical pressurized water reactor (PWR) static and transient analyses, especially for reactivity initiated accidents (RIA). The spatial variables are treated by using a polynomial nodal method (PNM) while for the neutron dynamic solver the adiabatic and improved quasi-static methods are adopted. A simple single channel thermal-hydraulics module and its steam table is implemented into the code. Verification works on static and transient benchmarks are being conducting to assess the accuracy of the code. For the static benchmark verification, the IAEA-2D, IAEA-3D, BIBLIS and KOEBERG light water reactor (LWR) benchmark problems were selected, while for the transient benchmark verification, the OECD NEACRP 3-D LWR Core Transient Benchmark and NEA-NSC 3-D/1-D PWR Core Transient Benchmark (Uncontrolled Withdrawal of Control Rods at Zero Power). Excellent agreement of the NODAL3 results with the reference solutions and other validated nodal codes was confirmed. (author)

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

    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

  8. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    Pasichnyk, I.; Perin, Y.; Velkov, K. [Gesellschaft flier Anlagen- und Reaktorsicherheit - GRS mbH, Boltzmannstasse 14, 85748 Garching bei Muenchen (Germany)

    2013-07-01

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  9. An evaluation of the failure of a PWR lower head during a core meltdown accident

    This paper presents an analysis of the failure of lower vessel head during a core meltdown accident. The analysis is limited to PWR systems with no penetration tubes attached to the lower vessel head. The case considered is characterized by a small quantity of corium and a relatively slow discharge into the lower plenum. The assumption of the breakup of the jet stream results in the solidification of debris particles and the formation of a debris bed thermally attacking the lower head wall. Detailed analyses were performed to determine the debris/water interaction, ablation of the lower head wall, and the time of vessel failure. Parameters which have significant effect on the results were identified. Parametric studies were performed to reflect uncertainties associated with the various phenomenological processes occurring during corium relocation into the lower head

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

    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

  11. Validation and verification of accident consequence assessment models

    Homma, T.; Togawa, O. [Japan Atomic Energy Research Inst., Tokyo (Japan); Takahashi, T. [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Arkhipov, A.N. [Chernobyl Science and Technology Centre for International Research (Ukraine)

    2001-03-01

    An accident consequence assessment code, OSCAAR, primarily designed by Japan Atomic Energy Research Institute (JAERI) for use in probabilistic safety assessment (PSA) of nuclear reactors in Japan, was applied to use for siting, emergency planning, and development of design criteria, and in the comparative risk studies of different energy systems. After verifying the code system through the international code comparison organized by CEC and OECD/NEA, the validation and improvements of the individual models and the verification of the whole OSCAAR code system were made. The cooperative research between Chernobyl Science and Technology Center for International Research (CHESCIR) and JAERI provided a valuable opportunity to test the performance of the accident consequence assessment models by comparing the model predictions with data obtained in the Chernobyl accidents. The predictive capabilities of OSCAAR were demonstrated using the accident source term and meteorological data for estimating the early exposure to the public occurred during and shortly after plume passage. The calculations indicated that ground-shine dose and inhalation dose, particularly from large nonvolatile particulates were the main contributors in the early stage of the accident. (S. Ohno)

  12. Reducing the consequences of reactor accidents with a green belt

    Considerable attention is being paid to reducing the consequences of low-probability accidents in nuclear power plants. A scheme based on the pollution absorption properties of trees is proposed to reduce early and continued mortalities among the general public due to an accident in a nuclear power plant. The consequences of a hypothetical case in which a large, cold, ground-level release of radionuclides into the atmosphere takes place have been analyzed in the absence and in the presence of a green belt (rows of trees). The results show that in the presence of a suitably designed green belt around a nuclear power plant, the consequences in terms of early and continued mortality as well as an interdiction area, involving relocation of population and supply of food stuff from an uncontaminated region, can be reduced by orders of magnitude. This could also help in substantially reducing the magnitude of emergency preparedness in the public domain

  13. Application of the WECHSL code to PWR and BWR specific accident scenarios

    The WECHSL Mod3 version is used to perform an accident analysis for a 1300 MW PWR and a BWR. The analysis starts after the melt has penetrated the reactor pressure vessel and is contained in the dry reactor cavity. The initial melt temperature is estimated to be 2673 K. In the initial phase of the melt/concrete interaction, the dominant energy source in the melt is the energy released in the zirconium oxidation reactions with the concrete decomposition products. Hence the concrete composition will determine the Zr-oxidation and the gas release rates as well as the composition of the released gases. Recent experiments and analyses have shown that the solidus temperature of the oxidic melt decreases much more rapidly with addition of concrete oxide than modelled previously. The solidus temperature of the oxide phase drops rapidly as concrete oxides are incorporated into the melt, approaching the concrete solidus at only about 10 to 20 weight percent of concrete oxides. The calculations are performed using the old estimate and the new solidus temperatures for both reactor types in order to study the influence of that oxide solidus temperature. The condensed Zr/SiO2 chemistry is only relevant for the PWR because of the high content of SiO2 in the siliceous concrete basemat. Compared to former analyses for the PWR the much faster zirconium oxidation leads to a higher temperature of about 100 K in the early phase of melt/concrete interaction and therefore the crust formation process starts later than in the former analyses leading to a longer duration of high gas release rates dominated by H2 because of more effective heat transfer to the concrete in this period of time. The concrete basemat of the BWR consists of pure limestone with a decomposition temperature which is higher than the solidus temperature of the metallic melt. This high concrete decomposition temperature prevents a crust formation at the metal-concrete boundary. Hence a very efficient heat transfer leads

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

    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

  15. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors

    Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

  16. The Influence of Seasonal Characteristics on the Accident Consequences Analysis

    In order to examine the influence of seasonal characteristics on accident consequences, we defined a limited number of basic spectra based on the relative importance of source term release parameters and meteorological conditions on offsite health effects and economic impacts. We then investigated the variation in numbers and frequency of early health effects and economic impacts resulting from the severe accidents of the YGN 3 and 4 nuclear power plants from spectrum to spectrum by using MACCS code. These investigations were for meteorological conditions defined as typical on an annual basis. Also, we investigated the variation in numbers and frequency of early health effects and economic impacts for the same standard spectra for meteorological conditions defined as typical on a seasonal basis recognizing that there are four seasons with distinct meteorological characteristics. Results show that there are large differences in consequences from spectrum to spectrum, although an equal amount and mix of radioactive material is released to the atmosphere in each case. Therefore, release parameters and meteorological data have to be well characterized in order to estimate accident consequences resulting from an accident accurately. Also, there are large differences in the estimated number of health effects and economic impacts from season to season due to distinct seasonal variations in meteorological conditions in Korea. In fall, the early fatalities and early fatality risk show minimum values due to enhanced dispersion arising from increased atmospheric instability, and the early fatalities show maximum value in summer due to a large rainfall rate. On the contrast, the economic cost shows maximum value in fall and minimum in summer due to different atmospheric dispersion and rainfall rate. Therefore, it is necessary to consider seasonal characteristics in developing emergency response strategies for reducing offsite early health risks in the event of a severe

  17. Consequences and experiences - ten years after the Chernobyl accident

    On 26 April 1986. the most serious accident in the history of the nuclear industry occurred at the Chernobyl nuclear power plant in the former Soviet Union, near the present borders of Ukraine, Belarus and Russia.Material released into the atmosphere dispersed and eventually deposited back on the surface of the earth,were it was measurable over the whole northern hemisphere. Millions of people and all segments of life and economy have been affected by the accident. Radioactive contamination has reached several tens of MBq/m2 in the area of 30 km diameter around the reactor in 1986., and plants and animals have been exposed to short lived radionuclides up to external doses of several tens of Gy. In the early phase after the accident, 237 persons were suspected to have acute radiation syndrome as a consequence of the Chernobyl accident, but diagnoses has been confirmed in 134 cases. In that phase 28 person have died as a consequence of exposure. There are significant non - related health disorders and symptoms, such as anxiety, depression and various psychosomatic disorders attributable to mental stress among the population in the region

  18. Estimated consequences from severe spent nuclear fuel transportation accidents

    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. Nuclear installations abroad the accident risks and their potential consequences

    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

  20. Consequences and effectiveness of relocation after nuclear accidents

    Extensive parameter studies have been performed with the program package COSYMA for probabilistic accident consequence assessments to quantify by means of PRA methods the interdependence of those quantities, which influence the extent, the duration, the efficiency and the monetary costs of relocation. As most important quantities, the amount of radionuclides released, the dose intervention levels for relocation, the (avoided) radiation doses in the population and the associated costs have been identified. Decontamination measures have also been included in the investigations, since they reduce the duration of relocation. The expression of all relevant accident consequences in monetary units allowed to investigate the applicability of cost/benefit analysis for deriving the most favourable intervention levels. It could be shown that weighting with different factors of collective doses calculated from different individual dose bands, and thus incorporating subjective judgements, significantly extends and improves the method. (orig./HP)

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

    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)

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

    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.)

  3. The investigation of Passive Accident Mitigation Scheme for advanced PWR NPP

    Highlights: • We put forward a new PAMS and analyze its operation characteristics under SBO. • We conduct comparative analysis between PAMS and Traditional Secondary Side PHRS. • The PAMS could cope with SBO accident and maintain the plant in safe conditions. • PAMS could decrease heat removal capacity of PHRS. • PAMS has advantage in reducing cooling rate and PCCT temperature rising amplitude. - Abstract: To enhance inherent safety features of nuclear power plant, the advanced pressurized water reactors implement a series of passive safety systems. This paper puts forward and designs a new Passive Accident Mitigation Scheme (PAMS) to remove residual heat, which consists of two parts: the first part is Passive Auxiliary Feedwater System (PAFS), and the other part is Passive Heat Removal System (PHRS). This paper takes the Westinghouse-designed Advanced Passive PWR (AP1000) as research object and analyzes the operation characteristics of PAMS to cope with the Station Blackout Accident (SBO) by using RELAP5 code. Moreover, the comparative analysis is also conducted between PAMS and Traditional Secondary Circuit PHRS to derive the advantages of PAMS. The results show that the designed scheme can remove core residual heat significantly and maintain the plant in safe conditions; the first part of PAMS would stop after 120 min and the second part has to come into use simultaneously; the low pressurizer (PZR) pressure signal would be generated 109 min later caused by coolant volume shrinkage, which would actuate the Passive Safety Injection System (PSIS) to recovery the water level of pressurizer; the flow instability phenomenon would occur and last 21 min after the PHRS start-up; according to the comparative analysis, the coolant average temperature gradient and the Passive Condensate Cooling Tank (PCCT) water temperature rising amplitude of PAMS are lower than those of Traditional Secondary Circuit PHRS

  4. Control room dose analysis for Maanshan PWR plant during design basis loss of coolant accident

    To address the issue identified in USNRC's Generic Letter 2003-1 that the unfiltered air in-leakage rate through plant's control room during design basis accident may exceed that assumed in the licensing analysis and thus threat the control room habitability, the control room radiation dose analysis of Maanshan PWR plant has to be re-performed to determine the allowable unfiltered air in-leakage rate. The allowable unfiltered air in-leakage rate is to be determined in such a way that the calculated whole body dose in the control room during the most limiting design basis accident must meet the criteria set forth in 10 CFR 50 Appendix A General Design Criteria (GDC) 19. The determined allowable air in-leakage rate is then employed as an acceptable limit to be met by the control room in-leakage test. In this study, the Maanshan plant control room dose analysis model during loss of coolant accident (LOCA) has been established based on USNRC's RADTRAD computer code. Different release and transport paths have been incorporated in this model, including containment leakage, engineered safety feature (ESF) leakage, and control room filtered and un-filtered air in-leakage. The RADTRAD calculation results are compared with Final Safety Analysis Report (FSAR) results to assure that overall consistency is reached. Finally, considering the uncertainties and margin to be maintained between RADTRAD calculation results and GDC-19 dose limits, an allowable unfiltered air in-leakage rate for control room habitability application during LOCA has been well defined. (author)

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

    The aim of the paper is to give an assessment of the conditions which would develop in the large, dry containment of a modern Westinghouse-type PWR during a severe accident where all safety systems are unavailable. The analysis is based principally on the results of calculations using the CONTAIN code, with a 4 cell model of the containment, for a station blackout (TMLB') scenario in which the vessel is assumed to fail at high pressure. In particular, the following are noted: (i) If much of the debris is in contact with water, so that decay heat can boil water directly, then the pressure rises steadily to reach the assumed containment failure point after 11/2 to 2 days. If most of the debris becomes isolated from water, for example, because of water is held up on the containment floors and in sumps and drains, the pressure rises too slowly to threaten the containment on this timescale. (ii) If a core-concrete interaction occurs, most of the associated fission product release takes place soon after relocation of molten fuel to the containment. The aerosols which transport these (and other non-gaseous fission products released earlier in the accident) in the containment agglomerate and settle. As a result, 0.1% or less of the aerosols remain airborne a day after the start of the accident. (iii) Hydrogen and carbon monoxide, which would accumulate in the containment are not expected to burn because the atmosphere would be inerted by steam. If, however, enough of the steam is condensed, for example, by recovering the containment sprays, a burn could occur but the resulting pressure spike is unlikely to threaten the containment unless a transition to detonation occurs. 6 refs., 6 tabs., 12 figs

  6. Thermal-hydraulic behavior of a PWR under accident conditions complementary test results from UPTF and PKL

    Two complementary test facilities - the Upper Plenum Test Facility (UPTF) and the Primaerkreislauf test facility (PKL) - were constructed to investigate the thermal-hydraulic response of a pressurized water reactor (PWR) during postulated accidents. The UPTF is a geometrical full-scale simulation of the primary system of a 1300-MW PWR. The upper plenum, the downcomer and the four connected loops as well as the pressurizer are represented on a 1:1 scale. The integral test facility PKL also simulates a 1300-MW PWR, whereby the power and volume is reduced by a factor of 1:145 (elevations 1:1). The PKL test facility models the entire primary system, relevant parts of the secondary side and all important engineered safety and auxiliary systems. Whereas the UPTF was mainly designed to perform separate-effect tests focusing on multidimensional thermal-hydraulic phenomena in full-scale simulated components, the main objective of the PKL tests has been the investigation of the thermal-hydraulic system behavior on the primary and secondary side. So far the program objectives represent a reasonable completion and in summary the experimental results from both test facilities provide an essential contribution for a better understanding of assumed accident sequences in a PWR. Test results which demonstrate the complementary character of the UPTF and the PKL test programs as well as the interaction between the two test facilities are presented in this paper. (author)

  7. Illustration interface of accident progression in PWR by quick inference based on multilevel flow models

    In this paper, a new accident inference method is proposed by using a goal and function oriented modeling method called Multilevel Flow Model focusing on explaining the causal-consequence relations and the objective of automatic action in the accident of nuclear power plant. Users can easily grasp how the various plant parameters will behave and how the various safety facilities will be activated sequentially to cope with the accident until the nuclear power plants are settled into safety state, i.e., shutdown state. The applicability of the developed method was validated by the conduction of internet-based 'view' experiment to the voluntary respondents, and in the future, further elaboration of interface design and the further introduction of instruction contents will be developed to make it become the usable CAI system. (authors)

  8. Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

    Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S2CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

  9. Health consequences of the Chernobyl accident: a review

    Full text of publication follows: on April 26, 1996, the accident at Chernobyl nuclear power plant led to the release into the atmosphere of considerable quantities of radionuclides. Most contaminated regions were in the southern Belarus, northern Ukraine and Bryansk and Kaluga regions of Russia. Main population groups exposed to the radioactivity released during the accident were the personnel at the Chernobyl plant and the rescue teams present on-site during the first hours, the cleanup workers (numbering about 600000) who participated in the decontamination and cleaning operations in the 30 km zone around the site, the residents of the same zone who were evacuated (numbering about 115000) and the inhabitants of contaminated zones (≥1 Ci/km2). Dose and dose rate levels as well as exposure pathways differ from one population group to another. A review of scientific articles published in the international literature till 1998 has been carried out. Apart the 28 deaths due to acute radiation sickness which occurred in the personnel of the plant and rescue teams within several days or weeks after the accident, two main public health consequences of the Chernobyl accident have been observed. First an unprecedented epidemic of thyroid cancers was detected in children first in 1992 in Belarus then in the Ukraine and to a lesser extent in Bryansk region. The spontaneous incidence of these tumours was multiplied by 100 in most contaminated regions. Although the role of the accident in this epidemic is now recognised, questions are raised regarding the respective role of radioactive agents and other environmental or genetic factors, and its evolution in the future. Regarding other kinds of solid cancers and leukemia, no excess has been clearly demonstrated in the residents of contaminated areas nor in liquidators. Second, results of available epidemiological investigations show an increased risk of psychological distress in residents of highly contaminated areas

  10. Tank Bump Accident Potential and Consequences During Waste Retrieval

    BRATZEL, D.R.

    2000-09-27

    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).

  11. Analysis of the core reflooding of a PWR reactor under a loss-of-coolant postulated accident

    The main purpose of this work is to analyse the termohydraulic behaviour of emergency cooling water, during reflooding of a PWR core submitted to a postulated loss-of-coolant accident, with the scope of giving the boundary conditions needed to verify fuel element and containment integrity. The analytical model presented was applied to the simulation of Angra I core reflooding phase, after a double-ended break between pressure vessel and discharge of one of the main coolant pumps. For this accident, with a discharge coefficient of C sub(D) = 0.4, the highest peak cladding temperature is expected. (author)

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

    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

  13. Offsite radiological consequence analysis for the bounding flammable gas accident

    The purpose of this analysis is to calculate the offsite radiological consequence of the bounding flammable gas accident. DOE-STD-3009-94, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', requires the formal quantification of a limited subset of accidents representing a complete set of bounding conditions. The results of these analyses are then evaluated to determine if they challenge the DOE-STD-3009-94, Appendix A, ''Evaluation Guideline,'' of 25 rem total effective dose equivalent in order to identify and evaluate safety class structures, systems, and components. The bounding flammable gas accident is a detonation in a single-shell tank (SST). A detonation versus a deflagration was selected for analysis because the faster flame speed of a detonation can potentially result in a larger release of respirable material. As will be shown, the consequences of a detonation in either an SST or a double-shell tank (DST) are approximately equal. A detonation in an SST was selected as the bounding condition because the estimated respirable release masses are the same and because the doses per unit quantity of waste inhaled are generally greater for SSTs than for DSTs. Appendix A contains a DST analysis for comparison purposes

  14. A rapid dose assessment and display system applicable to PWR accident

    The necessity of developing a rapid dose assessment system has been emphasized for an effective emergency response of mitigation of off-site radiological consequences. A microcomputer program based on a rapid dose assessment model of the off-site radiological consequences is developed for various accident sinarios for the Nuclear Power Plants in Korea. This model, which is consists of the user answering-question input format as a menu driven method and the output format of table and graphic types, is helpful to decision-making on Emergency Preparedness by being more rapidly able to implement the off-site dose assessment and to interpret the result. (Author)

  15. Assessment methods and minimization of radiological consequences of nuclear accidents

    The uncertainty and sensitivity analyses with the program system COSYMA for assessing the radiological consequences of nuclear accidents, performed since 1997 in close co-operation with the University of Delft, NL, and the NRPB, UK, have been terminated and fully documented. Work on the real-time on-line decision support system RODOS for off-site emergency management after nuclear accidents has concentrated on the preparation of the operational version PV 4.0; it will be released by mid 2000. It has been developed and customised to the various regions of Europe in close co-operation with some 40 contract partners in East and West Europe. The operational use of the RODOS system at a central place in Germany and in emergency centres of other West and East European countries is in progress. (orig.)

  16. Chernobyl victims: realistic evaluation of medical consequences of Chernobyl accident

    Objective assessment of early and delayed medical consequence of the Chernobyl accident is presented. Mortality of people due to acute radiation disease, burns and mechanical injuries are attributed to the early effects. Oncological and genetic diseases are considered as the delayed effects. Delayed radiation effects on the residents of contaminated territories were estimated by epidemiologic examination taking into account the dose due to radioactive fallout. Certain regions of Belarus, Russia and Ukraine were mostly exposed to contamination. Contamination density by 137Cs is considered and radiation doses due to natural sources and Chernobyl accident are compared. Disease incidence is analysed for carcinoma and genetic diseases. Health hazard caused by non-radiation accidental factors (psychological stress, victim psychology thrusting, groundless evacuation) is assessed

  17. Reports of the Chernobyl accident consequences in Brazilian newspapers

    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)

  18. Environmental radiological consequences of a loss of coolant accident

    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)

  19. Radiological consequences of the Three Mile Island accident

    The radiological consequences of the Three Mile Island (TMI) nuclear accident are discussed in detail. The nature, quantity and timing of the radioactive materials released to the atmosphere are established; mainly radioactive noble gases were emitted. A description is given of the radiological monitoring that occurred and the measured levels both inside and outside the plant are given as a function of time from the accident. In particular, the radioiodine release and its subsequent detection in milk analysis is described. The methods of establishing the population dosage are discussed; it is concluded that the collective dose equivalent is in the range 1600-3300 person rems. This implies a projected cancer (fatal and non-fatal) incidence of less than 1.5 in the offsite population within the 50 miles of the TMI site; the expected occurrence in this population is 541,000 cancers. The exposure of workers to radiation levels within the plant is also reported. (U.K.)

  20. Application of Westinghouse NEXUS/ANC9 cross-section model for PWR accident analyses

    NEXUS/ANC9 is the latest licensed PWR core design code system developed by Westinghouse. This system has demonstrated capabilities of modeling advanced core designs with improved accuracy in core reactivity and power distribution predictions. NEXUS/ANC9 system is being rolled out to replace the current APA system (ALPHA/PHOENIX-P/ANC) for routine core calculations. In addition to the standard core design calculations, investigations are underway to explore the possibility to expand the NEXUS/ANC9 application for safety analysis, especially at accident conditions. The main focus of the investigation is the evaluation of the NEXUS/ANC9 cross-section representation model conditions like high void and significant change of core pressure. Comparisons of the predicted parameters among ANC9, PARAGON lattice code and MCNP calculations are presented. The results show that NEXUS/ANC9 is able to model the cross-section behavior and accurately reproduce lattice code results at all simulated conditions. (author)

  1. The accident consequence model of the German safety study

    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.)

  2. Thyroid Consequences of the Fukushima Nuclear Reactor Accident

    Nagataki, Shigenobu

    2012-01-01

    Background A special report, ‘The Fukushima Accident’, was delivered at the 35th Annual Meeting of the European Thyroid Association in Krakow on September 11, 2011, and this study is the follow-up of the special report. Objectives To present a preliminary review of potential thyroid consequences of the 2011 Fukushima nuclear reactor accident. Methods Numerous new data have been presented in Japanese, and most of them are available on the website from each research institute and/or from each m...

  3. Primary disability of the Chernobyl Accident consequences liquidators

    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

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

    (NOEMAIL), K; Jonathan Lowrie, J; David Thoman (NOEMAIL), D; Austin Keller (NOEMAIL), A

    2008-07-30

    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.

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

    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

  6. Impact of improved neutronic methodology on the cladding response during a PWR reactivity initiated accident

    Hursin, Mathieu, E-mail: mathieu.hursin@psi.ch [Department of Nuclear Engineering of University of California at Berkeley, 4101 Etcheverry Hall, Berkeley, CA (United States); Downar, Thomas J., E-mail: downar@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI (United States); Montgomery, Robert, E-mail: Robert.Montgomery@pnnl.gov [Anatech Corp, San Diego, CA (United States)

    2013-09-15

    Highlights: • DeCART provides radial, azimuth, and axial power distribution during RIA analysis. • Coupling to FALCON is developed to evaluate the impact of such information. • Burnup calculation in the “two step” approach causes cladding load discrepancies. • The effect of azimuthal power variation has a 10% impact on the cladding load. -- Abstract: When applied to reactivity initiated accidents (RIAs) analysis, codes such as DeCART can provide a detailed radial, azimuth, and axial power distribution within a fuel rod. The work reported here is aimed at quantifying the sensitivity of the cladding thermo-mechanical response, calculated by the fuel performance code FALCON to the more accurate and detailed neutronic solution provided by DeCART for full PWR core RIA analysis. As a basis of comparison, the neutronics analysis is also performed with the U.S. NRC PARCS code, which is representative of the methodology used by the industry. Based on the DeCART solutions, several fuel rods are chosen for analysis with FALCON according to several relevant criteria. For each of the selected fuel rods, a FALCON study is performed using the boundary conditions provided by the neutronic solvers to predict the cladding response in terms of Strain Energy Density (SED) to the power pulse during the transient. The results of the analysis led to the following conclusions: • The largest impact on the cladding response can be attributed to the differences in the kinetic parameters in PARCS and DeCART. • The modeling of fuel pin exposure in the current industry standard “two step” methodology can result in some significant discrepancies in terms of SED during RIA analysis. • The effect of azimuthal power variation within a given fuel rod has a 10% impact on the SED and should be taken into consideration during RIA analysis, especially for high exposure fuel.

  7. Impact of improved neutronic methodology on the cladding response during a PWR reactivity initiated accident

    Highlights: • DeCART provides radial, azimuth, and axial power distribution during RIA analysis. • Coupling to FALCON is developed to evaluate the impact of such information. • Burnup calculation in the “two step” approach causes cladding load discrepancies. • The effect of azimuthal power variation has a 10% impact on the cladding load. -- Abstract: When applied to reactivity initiated accidents (RIAs) analysis, codes such as DeCART can provide a detailed radial, azimuth, and axial power distribution within a fuel rod. The work reported here is aimed at quantifying the sensitivity of the cladding thermo-mechanical response, calculated by the fuel performance code FALCON to the more accurate and detailed neutronic solution provided by DeCART for full PWR core RIA analysis. As a basis of comparison, the neutronics analysis is also performed with the U.S. NRC PARCS code, which is representative of the methodology used by the industry. Based on the DeCART solutions, several fuel rods are chosen for analysis with FALCON according to several relevant criteria. For each of the selected fuel rods, a FALCON study is performed using the boundary conditions provided by the neutronic solvers to predict the cladding response in terms of Strain Energy Density (SED) to the power pulse during the transient. The results of the analysis led to the following conclusions: • The largest impact on the cladding response can be attributed to the differences in the kinetic parameters in PARCS and DeCART. • The modeling of fuel pin exposure in the current industry standard “two step” methodology can result in some significant discrepancies in terms of SED during RIA analysis. • The effect of azimuthal power variation within a given fuel rod has a 10% impact on the SED and should be taken into consideration during RIA analysis, especially for high exposure fuel

  8. Evaluation of the structural integrity of the CPR1000 PWR containment under steam explosion accidents

    Highlights: • Detailed three dimensional finite element model with a full consideration of the complex structures, the complex materials and the complex loadings. • Damage to the concrete structure was calculated. • Influence of the thermal loading on the concrete barrel was evaluated. - Abstract: Detailed three-dimensional finite element models were set up to study the dynamic response and the possible damage of the CPR1000 PWR containment under steam explosion accidents with a full consideration of the complex geometric structures, the complex mechanical behavior of materials and the complex blast loadings. The structural integrity of the containment and the internal structures was evaluated under five typical steam explosion scenarios. In addition, the influence of the thermal loading was investigated by setting up a thermal-mechanical coupling finite element model. It is found that under steam explosions, only a small portion of energy is transferred to the concrete containment and therefore, the influence of the explosions on the containment is insignificant. Although the internal facilities and the structures are damaged severely by the blast loadings, the damage to the containment is negligible and the structural integrity is ensured. The thermal loading has a noticeable influence on the loading-capability of the containment structure. It is shown even a pure thermal load, i.e., a 150 °C temperature variation across the containment wall, can cause some damage to the concrete containment. The damage is further deepened by a simultaneous blast loading due to a steam explosion. However, the maximum depth of the damage is small compared with the thickness of the wall and the integrity of the concrete containment is still ensured

  9. Analysis of the loss of pool cooling accident in a PWR spent fuel pool with MAAP5

    Highlights: • A PWR spent fuel pool was modeled by using MAAP5. • Loss of pool cooling severe accident scenarios were studied. • Loss of pool cooling accidents with two mitigation measures were analyzed. - Abstract: The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents the analysis of loss of pool cooling accident scenarios and the discussion of mitigation measures for the SFP at a pressurized water reactor (PWR) NPP with the MAAP5 code. Analysis of uncompensated loss of water due to the loss of pool cooling with different initial pool water levels of 12.2 m (designated as a reference case) and 10.7 m have been performed based on a MAAP5 input model. Scenarios of the accident such as overheating of uncovered fuel assemblies, oxidation of claddings and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission products were predicted with the assumption that mitigation measures were unavailable. The results covered a broad spectrum of severe accident evaluations in the SFP. Furthermore, as important mitigation measures, the effects of recovering the SFP cooling system and makeup water in SFP on the accident progressions have also been investigated respectively based on the events of pool water boiling and spent fuels uncovery. Based upon the reference case, three cases with the recovery of SFP cooling system and three other cases with makeup water in SFP have been studied. The results showed that, severe accident might happen if SFP cooling system was not restored timely before the spent fuels started to become uncovered; spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate larger than the average evaporation rate defined as the division of pool water mass above the

  10. Consequences of the nuclear power plant accident at Chernobyl

    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

  11. Consequences of the nuclear power plant accident at Chernobyl

    Ginzburg, H.M.; Reis, E. (Health Resources and Services Administration, Rockville, MD (USA))

    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.

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

    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

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

    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

  14. Health effects estimation code development for accident consequence analysis

    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)

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

    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

  16. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    The safety challenges associated with sodium-cooled fast reactors have been recognized since the beginning of nuclear power and include the high power density in the core, the need for a reactor coolant and heat transfer system with high heat removal capability, the variation of power across the core requiring the use of ducted assemblies, and the condition that the fuel is not in the most neutronically reactive configuration during normal operation such that relocation can result in positive reactivity excursions, even possibly exceeding prompt critical conditions and energetic events. The potential for accidents with such severe consequences has been a negative factor with respect to the use of the sodium-cooled fast reactor. With the development of inherent safety principles, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, it is possible to greatly increase the level of safety, to the point where it is highly unlikely, or perhaps even not possible, for accidents to result in releases of hghly radioactive materials to the containment or the surrounding environment. (author)

  17. Towards more realistic assessment of reactor accident consequences

    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

  18. Accident consequence assessments with different atmospheric dispersion models

    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.)

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

    The effect of PWR plant normal and accident operating conditions on polymers forms the basis of nuclear qualification of safety-related containment equipment. This study was carried out on the request of safety organizations. Its purpose was to check whether accident simulations carried out sequentially during equipment qualification tests would lead to the same deterioration as that caused by an accident involving simultaneous irradiation and thermodynamic effects. The IPSN, DAS and the United States NRC have collaborated in preparing this study. The work carried out by ORIS Company as well as the results obtained from measurement of the mechanical properties of 8 industrial polymers are described in this report. The results are given in the conclusion. They tend to show that, overall, the most suitable test cycle for simulating accident operating conditions would be one which included irradiation and consecutive thermodynamic shock. The results of this study and the results obtained in a previous study, which included the same test cycles, except for more severe thermo-ageing, have been compared. This comparison, which was made on three elastomers, shows that ageing after the accident has a different effect on each material

  20. Study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP

    The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)

  1. Evaluation of severe accident risks: Quantification of major input parameters: MAACS (MELCOR Accident Consequence Code System) input

    Sprung, J.L.; Jow, H-N (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Helton, J.C. (Arizona State Univ., Tempe, AZ (USA))

    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.

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

    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

  3. PSYCHIATRIC CONSEQUENCES OF STRESS AFTER A VEHICLE ACCIDENT

    Dickov, Aleksandra; Martinović-Mitrović, Sladjana; Vučković, Nikola; Siladji-Mladenović, Djendji; Mitrović, Dragan; Jovičević, Mirjana; Mišić-Pavkov, Gordana

    2009-01-01

    Background: Vehicle accidents are a common cause of disease and death among people over 30 years of age. Essentially, reaction to stress due to the vehicle accident does not differ from the reaction to other stress factors. There are still no uniform viewpoints about the kind of sequels and their percentage representation after vehicle accidents. Subjects and methods: The research was provided as a prospective study, included 150 subjects who had vehicle accident minimum 2 years prior to t...

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

    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

  5. Numerical simulation of the insulation material transport to a PWR core under loss of coolant accident conditions

    Höhne, Thomas, E-mail: T.Hoehne@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Fluid Dynamics, P.O. Box 510119, D-01328 Dresden (Germany); Grahn, Alexander; Kliem, Sören; Rohde, Ulrich; Weiss, Frank-Peter [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Fluid Dynamics, P.O. Box 510119, D-01328 Dresden (Germany)

    2013-05-15

    Highlights: ► Detailed results of a numerical simulation of the insulation material transport to a PWR core are shown. ► The spacer grid is modeled as a strainer which completely retains the insulation material carried by coolant. ► The CFD calculations showed that the fibers at the upper spacer grid plane are not uniformly distributed. ► Furthermore the pressure loss does not exceed a critical limit. ► The PWR core coolablity can be guaranteed all the time during the transient. -- Abstract: In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modeled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the

  6. Numerical simulation of the insulation material transport to a PWR core under loss of coolant accident conditions

    Highlights: ► Detailed results of a numerical simulation of the insulation material transport to a PWR core are shown. ► The spacer grid is modeled as a strainer which completely retains the insulation material carried by coolant. ► The CFD calculations showed that the fibers at the upper spacer grid plane are not uniformly distributed. ► Furthermore the pressure loss does not exceed a critical limit. ► The PWR core coolablity can be guaranteed all the time during the transient. -- Abstract: In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modeled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the

  7. Analysis of severe accident on OPR1000 PWR plant at low power and shutdown states with MAAP5 code

    The objective of this paper is to provide a brief description of severe accident analysis using computer codes in Korean OPR1000 Plant at low power and shutdown states. The results of the analysis are utilized in preparing the shutdown severe accident management guidelines (LPSD SAMG). As part of the efforts to prepare LPSD SAMG, analysis of severe accident is performed at low power and shutdown states with MAAP5 code. The Korean OPR1000 plant, a PWR plant with 2 hot legs and 4 cold legs is considered as a reference plant in the analysis. In this study, the scenarios are selected based on the plant operational states (POS) and dominant initiating events (IE) which cause the core damages. Typical scenarios are the loss of shutdown cooling (LSCS) at various primary coolant levels and stuck-opening of valves which prevent the low temperature over pressurization (LTOP) of primary system. As the analysis results, the core uncovery is expected in 2∼6 hours. The maximum temperature of core exit exceeds 649degC (SAMG entry temperature) in 3∼7 hours. The molten corium starts to relocate into lower head in 5∼13 hours and reactor vessel failure is occurred in 11∼14 hours. The above mentioned timings are utilized to choose the possible actions and the timing to implement those actions LPSD SAMG. Also based on the results, the environmental conditions that instruments may encounter in a severe accident are determined. (author)

  8. Analysis of high burnup fuel behavior under rod ejection accident in the Westinghouse-designed 950 MWe PWR

    As there has arisen a concern that failure of the high burnup fuel under the reactivity-insertion accident (RIA) may occur at the energy lower than the expected, duel behavior under the rod ejection accident in a typical Westinghouse-designed 950 MWe PWR was analyzed by using the three dimensional nodal transient neutronics code, PANBOX2 and the transient fuel rod performance analysis code, FRAP-T6. Fuel failure criteria versus the burnup was conservatively derived taking into account available test data and the possible fuel failure mechanisms. The high burnup and longer cycle length fuel loading scheme of a peak rod burnup of 68 MWD/kgU was selected for the analysis. Except three dimensional core neutronics calculation, the analysis used the same core conditions and assumptions as the conventional zero dimensional analysis. Results of three dimensional analysis showed that the peak fuel enthalpy during the rod ejection accident is less than one third of that calculated by the core is less than 4 percent. Therefore, it can be said that the current design limit of less than 10 percent fuel failure and maintaining the core coolable geometry would be adequately satisfied under the rod ejection accident, even though the conservative fuel failure criteria derived from the test data are applied. (author)

  9. The aftermath of nuclear accidents on mental health; Consequences des accidents radiologiques sur la sante mentale

    Pirard, Ph.; Brenot, J.; Verger, P. [CEA Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire

    1998-10-01

    Technological disasters bring about psychological effects in exposed populations of various durability and intensity. This article reviews the epidemiological studies which assess psychological and psychiatric consequences of the Three Mile Island, Goieanna and Chernobyl accidents. It shows, in different accidental and cultural contexts, a statistically significant and durable increase of psychological symptoms in various exposed population groups, which points out an actual psychological distress. Diagnosed psychiatric effects are less constant, but much less studied. Most affected groups are mothers of young children, relocated persons, persons with less social support or in financial trouble. The psychological distress can further psychiatric disorders and give rise to behavioural changes towards health. More research is necessary to delineate the nature and the determinants of the observed symptoms and disorders. It implies to design better tools for the assessment of individual exposure and the diagnosis of mental health effects. (authors)

  10. The primal application research of figure assimilation theory in the nuclear accident consequence forecast

    The deepgoing research of figure assimilation theory promotes many subjects' rapid development. This article outlooks the application of figure assimilation technique in the nuclear accident consequence forecast. The nuclear accident consequence forecast is a complicated system which needs rapidity and precision, so it is quiet difficult. but after the insertion of figure assimilation, it pushes on one step about the question. (authors)

  11. Radiological consequences of the Three Mile Island accident

    The Three Mile Island Accident is described. The pathway and quantity of radioactive materials released, the radiological monitoring results and health effects are discussed. It was concluded that while radiological releases were small in view of the magnitude of fuel damage, the accident indicated that better health physics instrumentation and personnel training is required. (H.K.)

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

    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

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

    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)

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

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

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

    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)

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

    Jørgensen, Kirsten

    2016-01-01

    prevention or safety methodologies and procedures established for major accidents are applicable to simple accidents. The article goes back to basics about accidents causes, to review the nature of successful prevention techniques and to analyze what have been constraints to getting this knowledge used more...... 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 for the...... systematic observation of hazard sources in order to ascertain whether safety barriers and management deliveries are present. Safety management and safety culture, together with the INFO cards are important factors in the prevention process. The conclusion is that we must look at safety as a part of being a...

  17. The physical and chemical degradation of PWR fuel rods in severe accident conditions

    An experimental study of the interaction between Zircaloy-4 cladding and UO2 in PWR fuel rods heated to high temperatures with a negligible differential pressure across the cladding wall is described. The fuel rods were of dimensions appropriate to the 17x17 PWR fuel sub-assembly and were heated in a non-oxidising environment (vacuum) up to approx. 1850 deg. C either isothermally or through heating ramps. Observations were made concerning the extent and nature of the reaction zone between Zircaloy-4 and UO2 over the temperature range 1500-1850 deg. C for times ranging from 1 min to 125 min. The location, morphology and the chemical composition of the phases formed are described along with the kinetics of their formation. (author)

  18. Consequences of the nuclear power plant accident at Chernobyl.

    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 m...

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

    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.)

  20. Genetic consequences of the Chernobyl accident for Belarus republic

    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.)

  1. Accident management following loss of residual heat removal during mid-loop operation in a Westinghouse two-loop PWR

    Birchley, J. [Paul Scherrer Institut, CH-5232 Villigen (Switzerland)], E-mail: jonathan.birchley@psi.ch; Haste, T.J. [Paul Scherrer Institut, CH-5232 Villigen (Switzerland); Richner, M. [Nordostschweizerische Kraftwerke (NOK) - NPP Beznau, CH-5312 Doettingen (Switzerland)

    2008-09-15

    The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic studies as a major contributor to core melt frequency and source term risk during shutdown conditions. The wide range of plant states encountered and the unavailability of certain safety features make it difficult to guarantee that safety systems operation will always be sufficient to terminate the accident evolution. In this context analyses are performed using MELCOR 1.8.5 for loss of residual heat removal (RHR) at various times during mid-loop operation of a Westinghouse two-loop PWR. In the absence of recovery of RHR or other accident management (AM) measures, the sequences necessarily lead to a long term core uncovery, heat-up and degradation, loss of geometry and eventual failure of the reactor pressure vessel (RPV). The results show an extensive time window before uncovery and additionally before core damage, which increase progressively with increasing time after shutdown at which loss of RHR occurs. Significant oxidation of the cladding may result in concentrations of hydrogen sufficient for deflagration. The slow evolution implies an opportunity for the plant operators to initiate AM measures even after core uncovery has started. The analyses indicate a substantial time window during the uncovery within which the injection can recover the core without damage. The upper end of the window is determined by the temperature at which heat from cladding oxidation becomes a dominant factor, marking a critical point for the effectiveness of this recovery mode. The results provide confidence in the inherent robustness of the plant with respect to accident sequences of this type.

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

    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.)

  3. ICARE/CATHARE V1. Application to a PWR 900 MWe severe accident sequence

    Zabiego, Magali; Fichot, Florian; Guillard, Valia; Barrachin, Marc; Melis, Stephane; Chatelard, Patrick; Camous, Francine [Commissariat a l' Energie Atomique, Institut de Protection et de Surete Nucleaire, DRS/SEMAR/LECTA, Centre d' Etudes de Cadarache, St-Paul-Les-Durance Cedex (France); Lefevre, Bertrand [Communications et Systems, Centre d' Etudes de Cadarache, St-Paul-Lez-Durance (France)

    2000-11-01

    In the first part of this paper, a brief description of the V1 version of the ICARE/CATHARE software is presented. The new models developed by IPSN are described here. An application of this version to a French PWR 900 MWe is then shown. Although a description of the whole circuit is possible with ICARE/CATHARE V1, the present application is restricted to the reactor vessel. The results show the progression of the core degradation and the relocation of an important amount of corium in the reactor lower head. This calculation demonstrates the capacity of the code to provide a physically grounded simulation of the whole scenario. (author)

  4. Quantification of in-containment fission products source term for 1000 MWe PWR under loss of coolant accident

    Highlights: • Kinetic modeling for in-containment fission product activity. • Modeling and simulation of in-containment source term after LOCA. • Quantification of airborne in-containment activity. • BURNUP activity calculation and comparison with literature. • Study the effect of ESFs and coolant retention with mixing rate. - Abstract: The aim of this work is the modeling and simulation of in-containment fission products (FPs) quantification and behavior under loss of coolant accident (LOCA) in terms of NUREG-1465 key aspects. For this purpose, a kinetic model has been developed to determine the quantification and behavior of in-containment source term after loss of coolant accident for typical 1000 MWe PWR. A more realistic approach of continuous release of fission products from damaged core has been implemented with coolant retention. The simulation for in-containment fission product quantification influenced by containment atmosphere and containment system response has been carried out. Dramatic results have been obtained upon comparison study of fission product behaviors with different computational values. Moreover a contradiction in mixing rate (wx) value has been observed with a factor of 10 in comparison with Saeed et al. (2012)

  5. Investigation of the different scenarios occurring in a PWR in case of a TMLB accident

    Severe accidents in light water reactors fall into one of two main categories, depending on whether or not core meltdown is accompanied by a pressure buildup in the primary system. The way in which the accident develops is, in fact, largely conditioned by this pressure aspect: temperature distribution in the core and primary system resulting from natural convection gas streams; fuel clad failure mode, etc... One major effect of pressure buildup on the accident scenario is primary system failure under the combined actions of pressure and temperature. The purpose of the present paper is to present, after a detailed thermalhydraulic study, an analysis of the timing and location of the system failures in case of a TMLB accident on CPY french type reactor

  6. Radiation-biological consequences of the Chernobyl accident

    The paper points out essential aspects of the actual or potential impact of the Chernobyl reactor accident on human health in the areas immediately affected. In particular, radiation-induced diseases in the population are pointed out, which were caused by radioactive iodine. Epidemiological studies try to establish an increased incidence of leukaemia, lymphomas, and thyroid gland tumours. (DG)

  7. Radiological consequence of Chernobyl nuclear power accident in Japan

    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/m3 observed in Fukui, and the same kinds of radioactive nuclides as those in Europe were detected. (Kako, I.)

  8. The possibility of building nuclear power plant free from severe accident risk: PWR NPP with advanced all passive safety cooling systems (AAP SCS)

    A complete set of advanced all passive safety cooling systems (AAP SCS) for PWR NPP, actuated by natural force has been put forward in the article. Here the natural force mainly means the fore, which created by change of pressure distribution in the first loop of PWR as a result of operational regime conversion from one to another, including occurrence of accident situation. Correspondent safety cooling system will be actuated naturally and then put it into passive operation after occurring some kind of accident, so accidental situation will be mitigated right after it's occurrence and core residual heat will be naturally moved from the active core to the ultimate heat sink. There is no need to rely on automatic control system, any active equipment and human actions in all working process of the AAP SCS, which can reduce the probability of severe accident to zero, so as to exclude the need of evacuation plan around AAP nuclear power plant and eliminate the public's concern and doubt about nuclear power safety. Implementation of the AAP SCS concept is only based on use of evolutionary measures and state-of-the-art technology. So at present time it can be used for design of new-type third generation PWR nuclear power plant without severe accident risk, and for modernization of existing second generation nuclear power plant. (authors)

  9. The Chernobyl Accident 20 Years On: An Assessment of the Health Consequences and the International Response

    Baverstock, Keith; Williams, Dillwyn

    2006-01-01

    Background The Chernobyl accident in 1986 caused widespread radioactive contamination and enormous concern. Twenty years later, the World Health Organization and the International Atomic Energy Authority issued a generally reassuring statement about the consequences. Accurate assessment of the consequences is important to the current debate on nuclear power. Objectives Our objectives in this study were to evaluate the health impact of the Chernobyl accident, assess the international response ...

  10. An Analysis of the Cumulative Uncertainty Associated with a Quantitative Consequence Assessment of a Major Accident

    JIRSA PAVEL

    2005-01-01

    The task of the article is to quantify the uncertainty of the possible results of the accident consequence assessment of the chemical production plant and to provide some description of potentional problems with literature references and examples to help to avoid the erroneous use of available formulas. Based on numbers presented in the article we may conclude, that the main source of uncertainty in the consequence analysis of chemical accident assessment is surprisingly not only the dispers...

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

    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

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

    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

  13. Russian National Chernobyl Register as information and and analytical for Chernobyl accident medical consequences estimation

    The paper is devoted to using of the National Radiation and Epidemiology Register basic part, namely the Russian State Medical-Dosimetric Register of the people affected by the Chernobyl accident, to estimate the medical consequences of the accident. First part of article presents the common description and current state of Register. The estimation of medical consequences of the accident for clean-up workers is given in second part. The prognosis of radiation effects and definition of basic epidemiology factors to propose optimal medicalrehabilitation measures is discussed

  14. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    Joy L. Rempe; Darrell L. Knudson

    2014-05-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation

  15. TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

    Joy L. Rempe; Darrell L. Knudson

    2013-03-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation

  16. Extra-regulatory accident safety evaluation for the PWR S/F transport and storage system

    In the field of high speed crash, high speed impact analyses and test were performed for two systems, the dual purpose metal cask and the concrete cask considering the aircraft crash condition. Through the tests, the procedure and methodology of the assessment were successfully validated. In the field of transient fire, the computer simulation method for transient fire was drawn through the overseas status and methodology analysis. In the field of cumulative damage evaluation for transport accident, the analysis technique for assessment for cumulative damages which occurred from successive accident conditions was developed and proposed. And the sequential tests for the dual purpose cask were performed

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

    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

  18. External and internal accidents in PWR power plants. Comparison of current regulations in Belgium, United States, France, Federal Republic of Germany and United Kingdom

    In this report a comparison is made of the rules and practices applied in various countries (Belgium, France, Federal Republic of Germany, United Kingdom and United States of America) in designing PWR plants to resist natural hazards (first part of the report) and hazards associated with human activities (second part). The third part of the report deals with the practices in different countries concerning the protection against accidents of internal origin

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

    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.

  20. The radioecological consequences of Chernobyl accident for fish

    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 137Cs 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 137Cs 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

  1. Consequences of the Chernobyl accident in France. Thematic sheets; Les consequences de l'accident de Tchernobyl en France. Fiches thematiques

    NONE

    2006-07-01

    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)

  2. Accident at the Chernobyl nuclear power plant and its consequences

    In the early morning of April 26, 1986, as the culmination of an almost incredible series of errors that began 24 hours earlier, Unit 4 of the Chernobyl nuclear complex, a so-called RBMK-1000 reactor, suffered the worst accident in the history of commercial nuclear power. There was an uncontrolled nuclear excursion, release of a large amount of energy, possibly comparable to hundreds of pounds of TNT, blowing the top off the reactor. There was no containment, in the traditional American sense, so the roof of the building was blown out, an unprecedented amount of radioactivity was released to the biosphere, and a graphite fire was ignited, which burned for days. The radiation that was released spread through Eastern Europe (the world first learned of it through Swedish observations), bringing with it both official and unofficial protests that the Soviet Union had made no announcement of the radiation release until they were, in effect, caught. In fact, after a few days, the Soviets seemed to recognize that nuclear safety is a matter of international concern, and became quite open in their search for cooperation. They invited officials of the International Atomic Energy Agency (IAEA) to visit the area and to fly over the plant, and agreed, in the end, to make a complete disclosure of the details of the accident at a special meeting of IAEA in Vienna, August 25 to 29, 1986. In preparation for that meeting they distributed a lengthy (400 pages) report on the event. This paper reviews this report

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

    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.

  4. The Chernobyl accident 20 years on: an assessment of the health consequences and the international response.

    Baverstock, Keith; Williams, Dillwyn

    2007-01-01

    Twenty years after the Chernobyl accident the WHO and the International Atomic Energy Authority issued a reassuring statement about the consequences. Our objectives in this study were to evaluate the health impact of the Chernobyl accident, assess the international response to the accident, and consider how to improve responses to future accidents. So far, radiation to the thyroid from radioisotopes of iodine has caused several thousand cases of thyroid cancer but very few deaths; exposed children were most susceptible. The focus on thyroid cancer has diverted attention from possible nonthyroid effects. The international response to the accident was inadequate and uncoordinated, and has been unjustifiably reassuring. Accurate assessment in future health effects is not currently possible in the light of dose uncertainties, current debates over radiation actions, and the lessons from the late consequences of atomic bomb exposure. Because of the uncertainties from and the consequences of the accident, it is essential that investigations of its effects should be broadened and supported for the long term. The United Nations should initiate an independent review of the actions and assignments of the agencies concerned, with recommendations for dealing with future international-scale accidents. These should involve independent scientists and ensure cooperation rather than rivalry. PMID:17680126

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

    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)

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

    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)

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

    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

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

    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

  9. Analysis for relocation strategy using the method of probabilistic accident consequence assessment

    Relocation is one of the long-term protective actions in case of nuclear emergency to mitigate the consequences of an accidental release of radionuclides. The strategy of relocation is characterized by its protective benefit, cost and the corresponding residual dose in planning. This paper describes the application of a probabilistic accident consequence assessment model to the calculation of these quantities and the planning of relocation. Calculations of the consequence have been made of a postulated accident with source terms derived from a generic level 2 PSA. The results provided the insights for the development optimum dose criteria for introducing and terminating relocation. (author)

  10. Dose rate calculations inside the containment shell of a PWR in case of an accident

    The aim of the study is the evaluation of γ dose rates. In view to treat the different aspects of partial or total melting of the reactor core, leading to emission of a great quantity of fission radioactive products in the containement shell, implying the loss of integrity of the first two containement barriers, the study comprises: 1) Calculation of the γ dose rate as a function of time, normalized in considering a total release of fission products present in the core at the moment of the accident: it is done by taking into account the fission products in suspension in the containement shell, deposited on the internal surfaces and trapped by the water drain. 2) Transfer hypotheses of fission products (the release kinematics are linked to the accident scenario). The hypotheses of emission and deposit in this work are those of Rasmussen

  11. JERICHO computer code: PWR containment response during severe accidents description and sensitivity analysis

    The JERICHO code has been developed in order to study the thermodynamic behaviour inside the reactor containment building for the complete spectrum of accident sequences likely to occur in such a reactor, including models for the various mass and energy transfer phenomena, for water spray, for hydrogen and carbon monoxide flammability limits and combustion, as well as for containment venting. Sensitivity analyses have been performed on a severe accident sequence, (namely, small LOCA with failure of the emergency core cooling and containment spray systems), involving core melting and subsequent concrete containment basemat erosion. The effect of various models, such as mass and energy transfer to the structures, has been studied. The influence of the concrete composition, of the fission product deposition and of the thermal degradation of the reactor cavity concrete walls on long term thermodynamic behaviour has also been investigated

  12. A Study of Containment Sprays and Charcoal Filters for the Removal of Iodine Following a PWR Loss of Coolant Accident

    This paper summarizes a study of the radioiodine problem associated with the siting of the current generation of nuclear power stations in the United States, and compares the effectiveness of containment spray systems and charcoal filter systems for the removal of the radioiodine following a PWR loss of coolant accident. The study showed that, based on the present reactor siting criteria in the United States, the ultimate iodine removal system must provide an iodine dose reduction factor (DRF) of 13 in the first two hours and then eventually achieve a DRF of 50. This means that about 2% of the iodine need never be removed from the containment atmosphere if the other 98% is removed sufficiently fast. The performance of iodine removal systems is shown as a function of operating time and design capacity. For removal systems operated in a recirculating mode, the system's iodine removal efficiency is relatively unimportant when compared with the effect of the presence of a small quantity of a penetrating component. The slow response time of a charcoal filter system limits the maximum dose reduction factor that can be practically obtained in the first two hours to between 5 and 8. However, the fast response of a containment spray system provides the capability of achieving an iodine dose reduction factor of about 50 within the first two hours after an accident. If a small quantity of a penetrating component is present, the charcoal filter system will eventually catch up to the spray system, and both systems will approach the same dose reduction factor. (author)

  13. The consequences of the Kyshtym accident for Flora and Fauna

    Flora and fauna irradiated in areas radioactively contaminated by the Kyshtym accident accumulated the bulk of their dose more or less in the first year, with the irradiation being at its most intensive in autumn 1957 and the winter of 1957/58, when plants and many animal species were in the physically dormant state. During the ''acute'' phase the maximum doses absorbed (at a contamination level of 4 000 Ci 90 Sr/km2) were as follows (in krad): mouse-type rodents and fish 4, birch bud meristems 20, pine bud meristems 40, pine needles 80, dormant leaf buds and gramineae seeds on the soil surface 160. The main radiobiological effects appeared in the spring of 1958 and were observed for several years after; subsequently, in the presence of chronic irradiation at a low dose rate, predominantly genetic effects were observed, conifers being the most radiosensitive among the plants. Changes in the structure of herbaceous communities occurred at doses over 20 krad (1 500 Ci 90Sr/km2). In subsequent years we observed changes in the structure and numbers of fish and mouse-type rodent populations. The radioactive contamination caused an increase in the rate of mutational processes in plant and animal populations. However, for populations as a whole the increased frequency observed for most mutations (chromosome aberrations, chemical mutations) did not play a major role, since they were speedily eliminated by natural selection. No deformities of a genetic nature were found on the contaminated territory. In the 30 years since the accident the biological characteristics of the contaminated area have not differed (except for coniferous forests) from those of the surrounding regions. Natural ecosystems are very radioresistant, and extremely high doses are needed to damage them seriously and irreversibly. (author)

  14. About some regularities consequences of accidents at NPP

    In the article are considered the principal reasons of differences in observed and calculation frequencies of emergency events in complex technological systems, including NPP. Analyzed simplification inherent in the probabilistic model of work of the reactor. The results of application of power-law probability distribution are presented for the estimation of consequences of catastrophes in complex and dangerous technologies

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

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

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

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

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

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

  18. Help guides for post-accident consequence management: farm activities and exiting the emergency phase

    After having recalled the main actions foreseen in the PPIs (plans particuliers d'intervention, intervention specific plans) in case of radionuclide release in the environment after a nuclear accident, i.e. sheltering and ingestion of steady iodine, and also indicated the different phases of consequence management (preparation, emergency and post-accident phases), this report describes and comments the contents of two guides published by the IRSN (the French Radioprotection and Nuclear Safety Institute) and dealing with the management of post-accident consequences. The first one is a guide to aid to decision-making for the management of the agricultural sector in case of nuclear accident, and the second one is a guide for the preparation of the end of the emergency phase in which actions to be performed during the first week after the end of accidental releases are described

  19. A study of core melting phenomena in reactor severe accident of PWR

    Jeun, Gyoo Dong; Park, Shane; Kim, Jong Sun; Kim, Sung Joong [Hanyang Univ., Seoul (Korea, Republic of); Kim, Jin Man [Korea Maritime Univ., Busan (Korea, Republic of)

    2001-03-15

    In the 4th year, SCDAP/RELAP5 best estimate input data obtained from the TMI-2 accident analysis were applied to the analysis of domestic nuclear power plant. Ulchin nuclear power plant unit 3, 4 were selected as reference plant and steam generator tube rupture, station blackout SCDAP/RELAP5 calculation were performed to verify the adequacy of the best estimate input parameters and the adequacy of related models. Also, System 80+ EVSE simulation was executed to study steam explosion phenomena in the reactor cavity and EVSE load test was performed on the simplified reactor cavity geometry using TRACER-II code.

  20. Definition of loss-of-coolant accident radiation source. [PWR; BWR

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist.

  1. The influence of seasonal conditions on the radiological consequences of a nuclear accident

    The impact of an accidental release of radioactivity to the environment can be strongly influenced by prevailing environmental conditions. Thus, potential variations in accident consequences caused by variable seasonal, meteorological or climatic conditions are of significance to the development and application of protective measures and emergency response plans. These proceedings present the results of a workshop organized by the NEA to examine such aspects of emergency response to a nuclear accident

  2. Consequences of tractor accidents in the agriculture in Republic of Macedonia

    Dimitrovski, Zoran

    2008-01-01

    In this paper are the results from the research of the consequences of tractor accidents in the agriculture in Republic of Macedonia. During the research from 1999 till 2003, 610 people have been injured in Republic of Macedonia in agricultural production, and the tractors have been the main reason. 544 people have been injured in tractor traffic accidents and 66 have been injured during tractor operating in agricultural condition. From the total number, 101 people have died during this perio...

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

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

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

    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

  5. Consequences of the Chernobyl accident for reindeer husbandry in 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. Reactor accident at the Chernobyl nuclear power plant-Block 4. Effects, countermeasures and consequences

    The findings of the Summary Report on the Chernobyl accident issued by IAEA in September 1986 (International Nuclear Safety Advisory Group (INSAG): Summary Report on the Post-Accident Review Meeting on the Chernobyl Accident. Safety Series No. 78-INSAG-1 Vienna, International Atomic Energy Agency (IAEA). Sept. 1986) are updated, reviewing more recent publications providing more complete information on the events both within and outside the plant. The available information on the resulting radioactive pollution of agriculture and the food chain is discussed considering also the consequences for the future in comparison with the other sources of radioactivity in the environment. 21 refs.; 3 figs.; 3 tabs

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

    G. V. Arkhangelskaya

    2012-01-01

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

  8. Assessment of radiation consequences of cabins in a nuclear accident of the nuclear ship

    The author discussed about the spread routes of radioactive nuclides from reactor cabin to other cabins and their distributions in these cabins. Methods and formulas to estimate radioactivities of nuclides and doses received by crews in cabins were established. The radiation consequences of cabins in a nuclear accident was quantified and evaluated. The assessments indicates that the consequences of cabins is light and the doses to the staff will not exceed the dose limits prescribed in standards in a design basis accident, and the consequences of cabins is serious and the doses to the staff will exceed the dose limits prescribed in standards in serious accident. Some suggestions on emergency management and radiation protection were given

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

    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

  10. RADIOLOGICAL AND MEDICAL CONSEQUENCES OF THE CHERNOBYL ACCIDENT

    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.