WorldWideScience

Sample records for plant criticality accident

  1. Plant safety review from mass criticality accident

    International Nuclear Information System (INIS)

    Susanto, B.G.

    2000-01-01

    The review has been done to understand the resent status of the plant in facing postulated mass criticality accident. From the design concept of the plant all the components in the system including functional groups have been designed based on favorable mass/geometry safety principle. The criticality safety for each component is guaranteed because all the dimensions relevant to criticality of the components are smaller than dimensions of 'favorable mass/geometry'. The procedures covering all aspects affecting quality including the safety related are developed and adhered to at all times. Staff are indoctrinated periodically in short training session to warn the important of the safety in process of production. The plant is fully equipped with 6 (six) criticality detectors in strategic places to alert employees whenever the postulated mass criticality accident occur. In the event of Nuclear Emergency Preparedness, PT BATAN TEKNOLOGI has also proposed the organization structure how promptly to report the crisis to Nuclear Energy Control Board (BAPETEN) Indonesia. (author)

  2. Prevention of criticality accidents in a fuel cycle plant

    International Nuclear Information System (INIS)

    Gatti, A.M.; Canavese, S.I.; Capadona, N.M.

    1990-01-01

    This work reports the basic considerations on criticality accidents applied to an uranium dioxide fuel cycle production plant. The different fabrication stages are briefly described, with the identification of the neutronically isolated areas. Once the areas have been defined, an evaluation is made, setting up the control parameters to be used in each of them and their variation ranges; normal operation limitations based on experimental data or validating calculations, applied specifically to 5% enriched uranium, are established. Afterwards, defined parameters deviations are analyzed due to incidental conditions in order to prevent criticality accidents under normal conditions and maintenance operations. (Author) [es

  3. Measurement of the Portsmouth Gaseous Diffusion Plant criticality accident alarm

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  4. Measurement of the Portsmouth Gaseous Diffusion Plant criticality accident alarm

    International Nuclear Information System (INIS)

    Tayloe, R.W. Jr.; McGinnis, B.

    1990-01-01

    Measurements of the Portsmouth Gaseous Diffusion Plant's nuclear criticality accident radiation alarm signal response time, sound wave frequency, and sound volume levels were made to demonstrate compliance with ANSI/ANS-8.3-1986. A steady-state alarm signal is produced within one-half second of obtaining a two-out-of-three detector trip. The fundamental alarm sound wave frequency is 440 hertz. The sound volume levels are greater than 10 decibels above background and ranged from 100 to 125 A-weighted decibels. The requirements of the standard were met; however the recommended maximum sound volume level of 115 dBA was exceeded. Emergency procedures require immediate evacuation upon initiation of a facility's radiation alarm. Comparison with standards for allowable time of exposure at different noise levels indicate that the elevated noise level at this location does not represent an occupational injury hazard. 8 refs., 5 figs

  5. Criticality accident:

    International Nuclear Information System (INIS)

    Canavese, Susana I.

    2000-01-01

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

  6. NIRS report of the criticality accident in a uranium conversion test plant in Tokai-mura

    International Nuclear Information System (INIS)

    2001-01-01

    This report is a detailed account of the roles that National Institute of Radiological Sciences (NIRS) played at the criticality accident in the title, which occurred at around 10:35, on Sep. 30, 1999 and resulted in death of two workers after all, and is published to discharge NIRS responsibilities in regards to the accident. The accident caused many residents concern on their health and rumors had both social and economic consequences. The report involves chapters of detailed outline of the accident; demand for acceptance of the victims and communications until the identification of the criticality'' accident; the acceptance and initial treatment; the exposure dose estimation (based on acute symptoms, on physics, on chromosomal analyses and on neutron-activated dental metals, and detailed analyses for dose distribution); decision made for therapeutic strategies; cooperation with the Network Council for Radiation Emergency and with other medical facilities; the urgent import of medicine; treatment and processes (patients, nursing system and radiation injuries); radiation protection in medical facilities; response to nearby residents of the Plant; international response; press release; Uranium Processing Plant Criticality Accident Investigation Committee and the Health Management Committee organized by the Nuclear Safety Commission; handling of information; and radiation emergency medical preparedness at the NIRS (future issues and prospect). The report is hopefully useful in preventing the occurrence of future accidents. (N.I.)

  7. The report of the criticality accident in a uranium conversion test plant in Tokai-mura

    International Nuclear Information System (INIS)

    Murata, Hajime; Akashi, Makoto

    2002-03-01

    The criticality accident in the title occurred at around 10:35, on Sep. 30, 1999, cost the lives of two workers and caused many residents concern on their health. Moreover, rumors had both social and economic consequences. This report is a detailed account of the roles that many individuals and groups in the National Institute of Radiological Sciences (NIRS) performed in a range of the areas, and is published to discharge NIRS responsibilities in regards to the accident. The report involves chapters of detailed outline of the accident; acceptance of the victims and communications until the identification of the ''criticality'' accident; initial treatment; dose estimation (medical, hematological, physical and biological ones and that by dental metals activated by the neutron); decision making for therapeutic strategies; cooperation with the Network Council for Radiation Emergency Medicine and other medical facilities; emergency importation of medical supplies; treatment and progress (nursing system and radiation injuries); protection from radiation in medical facilities; response to nearby residents of the Plant; international response; press release; Uranium Processing Plant Criticality Accident Investigation Committee and the Health Management Committee organized by the Nuclear Safety Commission; handling of information; and radiation emergency medical preparedness at the NIRS (future issues and prospect). The report is hoped to be useful in preventing the occurrence of future accidents. (K.H.)

  8. Replacement of the criticality accident alarm system in the Tokai reprocessing plant

    International Nuclear Information System (INIS)

    Sanada, Yukihisa; Momose, Takumaro; Suzuki, Kei; Kawai, Keiichi

    2008-01-01

    A Criticality Accident Alarm System (CAAS) was installed as part of criticality safety management for use in reducing the radiation workers could be exposed to in the rare case of a criticality accident. The initial CAAS version was installed the Tokai Reprocessing Plant (TRP) in the 1980s. It includes units that can detect gamma-rays or neutron-rays released in criticality accidents (CADs), one of which consists of three plastic scintillation gamma detectors and three solid state neutron detectors with fissile material, and in being highly reliable utilizes the 2 out of 3 voting system. The purpose of this study is to give the design principles and procedures for determining the adequate relocation of the CADs within the TRP. The optimal places for the CADs to be relocated to were determined using a conservative evaluation method. Firstly, equipment needing to be monitored for criticality accidents was selected with consideration given to the risk of excessive exposure to workers. Secondly, the detection threshold of a minimum accident was set to be an increase in power of 10 15 fissions/s occurring within a rise-time of between 0.5 ms and 1 s. The sum of neutron and gamma doses of a minimum accident (10 15 fissions) was 0.3 Gy at an unshielded distance of 1 m. Finally, doses at where the CADs were installed were evaluated using parameters calculated with MCNP and ANISN. As a result, the alarm trip level of both the gamma detector and the neutron detector being set at 2.0 mGy/h enabled minimum criticality accidents to be conservatively detected. These results were then applied to the new CAD positions. (author)

  9. An analysis on human factor issues in criticality accident at a uranium processing plant. Investigation on human behavior contributing to the criticality accident. Interim report

    International Nuclear Information System (INIS)

    Sasou, Kuonihide; Goda, Hideki; Hirotsu, Yuko

    1999-01-01

    At 10:30 am, September 30th, 1999, a criticality accident occurred in a conversion building of a uranium processing plant in Tokai, Ibaraki prefecture. 69 people including 3 workers who then worked at the building, 3 fire fighters who dispatched to rescue them were exposed to the radiation. People with a 350 m-radius of the site were recommended to evacuate themselves from the region to a temporarily prepared evacuation center. And about one hundred thousand people within a 10 km-radius were also advised to stay inside of their home. Nuclear Safety Commission's Accident Investigation Committee is investigating causes of this accident and have been revealing that deviation from government-authorized processing method and negligence of its illegal procedure had contributed to the accident. The influence of this accident is expanding not only to the plant operating company, local people but also to Japanese nuclear power policy, the whole nuclear industry in Japan. Especially pervasion of 'Safety Culture' is strongly being required. This report analyses latent factors of some human behavior directly contributing to the criticality accident. It also mentions that 4 critical points on the poor climate for safety in the work place, the inadequate safety management, the unsuitable equipment and the production-biased company's policy are the latent factors of this accident. It also finds that the poor climate and the production-biased policy are the most important factors. It can be said that some people directly or indirectly having caused the accident are the victims of them. (author)

  10. The relationship of JNC and JCO in the uranium processing plant criticality accident

    International Nuclear Information System (INIS)

    Kanamori, Masashi; Yanagibashi, Katsumi; Okamoto, Naritoshi

    2002-12-01

    On September 30th 1999, the criticality accident occurred at JCO's uranium conversion building in Tokai. The accident occurred during reconversion from U 3 O 8 to uranium nitrate solution (UNH) with uranium enriched 18.8% and about 60 kgU. JCO contacted with JNC to supply UNH that is fuel material for the experimental fast breeder reactor 'JOYO'. JNC has contracted with JCO that had started nuclear fuel material processing business following a definite policy of Japanese government and developed SUMITOMO ADU PROCESS'. JNC made the first contract with JCO in 1985 and has made a contact every year. There had never been a problem in their products. JNC inspected products based on contract. JNC discharge our duty as customer inspecting products based on contract. As for safety control, JCO had taken licensing safety review and had been permitted to be 'a processing facility'. Therefore JNC understood that JCO produced following this license. 'The Uranium Processing Plant Criticality Accident Investigation' showed that JCO had been taking a different method from the permit and violating the license. However JNC had never been explained about that and JCO's operation procedures had never described about that. Therefore the Criticality Accident couldn't be avoided. This report describes the relationship of JNC and JCO in the uranium reconversion contract for JOYO, atomic development policy of Japanese government, process to the order and the contents of contract. (author)

  11. Criticality accident in uranium fuel processing plant. Questionnaires from Research Committee of Nuclear Safety

    International Nuclear Information System (INIS)

    Kataoka, Isao; Sekimoto, Hiroshi

    2000-01-01

    The Research Committee of Nuclear Safety carried out a research on criticality accident at the JCO plant according to statement of president of the Japan Atomic Energy Society on October 8, 1999, of which results are planned to be summarized by the constitutions shown as follows, for a report on the 'Questionnaires of criticality accident in the Uranium Fuel Processing Plant of the JCO, Inc.': general criticality safety, fuel cycle and the JCO, Inc.; elucidation on progress and fact of accident; cause analysis and problem picking-up; proposals on improvement; and duty of the Society. Among them, on last two items, because of a conclusion to be required for members of the Society at discussions of the Committee, some questionnaires were send to more than 1800 of them on April 5, 2000 with name of chairman of the Committee. As results of the questionnaires contained proposals and opinions on a great numbers of fields, some key-words like words were found on a shape of repeating in most questionnaires. As they were thought to be very important nuclei in these two items, they were further largely classified to use for summarizing proposals and opinions on the questionnaires. This questionnaire had a big characteristic on the duty of the Society in comparison with those in the other organizations. (G.K.)

  12. Criticality accident in Argentina

    International Nuclear Information System (INIS)

    Oliveira, A.R. de.

    1984-01-01

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

  13. An analysis on human factor issues in criticality accident at a uranium processing plant

    International Nuclear Information System (INIS)

    Sasou, Kunihide; Goda, Hidenori; Hirotsu, Yuko

    2000-01-01

    This report analyses latent factors of a human behavior directly contributing to the criticality accident. It is pouring some 16 kg-U with an enrichment of 18.8% into the precipitation tank. It is the fact that the direct cause of this accident is the workers' unsafe act. However, the authors find lots of latent factors relating to the production-biased company's policy, the poor climate for safety in the work place, the inadequate safety management and the unsuitable equipment. This accident was caused by many organizational factors. This paper also discusses lessons learned from this accident. (author)

  14. Character and consequence of nuclear criticality accident

    International Nuclear Information System (INIS)

    Liu Xinhua; Liu Hua; Wu Deqiang; Li Bing

    2001-01-01

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

  15. Criticality accident alarm system

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  17. Criticality accident in uranium fuel processing plant. Emergency medical care and dose estimation for the severely overexposed patients

    Energy Technology Data Exchange (ETDEWEB)

    Akashi, Makoto; Ishigure, Nobuhito [National Inst. of Radiological Sciences, Chiba (Japan)

    2000-08-01

    A criticality accident occurred in JCO, a plant for nuclear fuel production in 1999 and three workers were exposed to extremely high-level radiation (neutron and {gamma}-ray). This report describes outlines of the clinical courses and the medical cares for the patients of this accident and the emergent medical system for radiation accident in Japan. One (A) of the three workers of JCO had vomiting and diarrhea within several minutes after the accident and another one (B) had also vomiting within one hour after. Based on these evidences, the exposure dose of A and B were estimated to be more than 8 and 4 GyEq, respectively. Generally, acute radiation syndrome (ARS) is assigned into three phases; prodromal phase, critical or manifestation phase and recovery phase or death. In the prodromal phase, anorexia, nausea, vomiting and diarrhea often develop, whereas the second phase is asymptotic. In the third phase, various syndromes including infection, hemorrhage, dehydration shock and neurotic syndromes are apt to occur. It is known that radiation exposure at 1 Gy or more might induce such acute radiation syndromes. Based on the clinical findings of Chernobyl accident, it has been thought that exposure at 0.5 Gy or more causes a lowering of lymphocyte level and a decrease in immunological activities within 48 hours. Lymphocyte count is available as an indicator for the evaluation of exposure dose in early phase, but not in later phase The three workers of JCO underwent chemical analysis of blood components, chromosomal analysis and analysis of blood {sup 24}Na immediately after the arrival at National Institute of Radiological Sciences via National Mito Hospital specified as the third and the second facility for the emergency medical care system in Japan, respectively. (M.N.)

  18. JCO criticality accident termination operation

    International Nuclear Information System (INIS)

    Kanamori, Masashi

    2001-12-01

    On September 30 at around 10:35 AM, criticality accident occurred at the JCO's conversion building in Tokai-mura. Since criticality accident had not been anticipated, neither devices for termination of criticality accident nor neutron detectors were available. Immediately after the information of the accident, our emergency staff (Japan Nuclear Cycle development institute staff) went to JCO site, to measure the intensity of neutrons and gammas. There were four main tasks, first one was to measure the radiation intensity, second one was to terminate the criticality accident, third one is to alert the residents surrounding the JCO site, fourth one is to evacuate the employees in the site. These tasks were successfully performed until October 1. This paper describes about how these operations were performed by the relevant staffs. (author)

  19. First Responders and Criticality Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Valerie L. Putman; Douglas M. Minnema

    2005-11-01

    Nuclear criticality accident descriptions typically include, but do not focus on, information useful to first responders. We studied these accidents, noting characteristics to help (1) first responders prepare for such an event and (2) emergency drill planners develop appropriate simulations for training. We also provide recommendations to help people prepare for such events in the future.

  20. Determination of the response function for the Portsmouth Gaseous Diffusion Plant criticality accident alarm system neutron detectors

    International Nuclear Information System (INIS)

    Tayloe, R.W. Jr.; Brown, A.S.; Dobelbower, M.C.; Woollard, J.E.

    1997-03-01

    Neutron-sensitive radiation detectors are used in the Portsmouth Gaseous Diffusion Plant's (PORTS) criticality accident alarm system (CAAS). The CAAS is composed of numerous detectors, electronics, and logic units. It uses a telemetry system to sound building evacuation horns and to provide remote alarm status in a central control facility. The ANSI Standard for a CAAS uses a free-in-air dose rate to define the detection criteria for a minimum accident-of-concern. Previously, the free-in-air absorbed dose rate from neutrons was used for determining the areal coverge of criticality detection within PORTS buildings handling fissile materials. However, the free-in-air dose rate does not accurately reflect the response of the neutron detectors in use at PORTS. Because the cost of placing additional CAAS detectors in areas of questionable coverage (based on a free-in-air absorbed dose rate) is high, the actual response function for the CAAS neutron detectors was determined. This report, which is organized into three major sections, discusses how the actual response function for the PORTS CAAS neutron detectors was determined. The CAAS neutron detectors are described in Section 2. The model of the detector system developed to facilitate calculation of the response function is discussed in Section 3. The results of the calculations, including confirmatory measurements with neutron sources, are given in Section 4

  1. JCO criticality accident termination operation

    OpenAIRE

    金盛 正至

    2010-01-01

    In 2001, we summarized the circumstances surrounding termination of the JCO criticality accident based on testimony in the Mito District Court on December 17, 2001. JCO was the company for uranium fuels production in Japan. That document was assembled based on actual testimony in the belief that a description of the work involved in termination of the accident would be useful in some way for preventing nuclear disasters in the future. This year is the tenth year of the JCO criticality acciden...

  2. Prevention of criticality accidents

    International Nuclear Information System (INIS)

    Canavese, S.I.

    1982-01-01

    These notes used in the postgraduate course on Radiological Protection and Nuclear Safety discuss macro-and microscopic nuclear constants for fissile materials systems. Critical systems: their definition; criteria to analyze the critical state; determination of the critical size; analysis of practical problems about prevention of criticality. Safety of isolated units and of sets of units. Application of standards. Conception of facilities from the criticality control view point. (author) [es

  3. Design of and experience with the gamma-detecting criticality accident alarm system at ALKEM MOX fuel fabrication plant

    International Nuclear Information System (INIS)

    Kindleben, G.

    1988-01-01

    At ALKEM mixed oxide fuel fabrication plant there are two criticality accident alarm systems in operation and another one is planned for different buildings. They use ionization chambers for gamma-measuring. The measuring channels are self controlled with implemented test sources. The order of limit transgression at the detectors is registrated. The interpretation indicates the room of the radiation source, which is signaled by flash lights. Extensive radiation protection shieldings make detector-placing a complex problem with secondary gamma-radiation to be taken into account. Most of the appearing defects can easily be repaired by exchange of components. Some of them have been eliminated by technical modification. Redundancy prevents total system failure. Some false alarms occurred during the operation time of the alarm systems. The main reason is pulse induction, resulting from lightning strike. Measures to prevent such events have been taken, while further measures are being considered

  4. A review of criticality accidents

    International Nuclear Information System (INIS)

    Stratton, W.R.; Smith, D.R.

    1989-03-01

    Criticality accidents and the characteristics of prompt power excursions are discussed. Forty-one accidental power transients are reviewed. In each case where available, enough detail is given to help visualize the physical situation, the cause or causes of the accident, the history and characteristics of the transient, the energy release, and the consequences, if any, to personnel and property. Excursions associated with large power reactors are not included in this study, except that some information on the major accident at the Chernobyl reactor in April 1986 is provided in the Appendix. 67 refs., 21 figs., 2 tabs

  5. Criticality accident in uranium fuel processing plant. Cause analysis and teachings from a viewpoint of a human factor

    International Nuclear Information System (INIS)

    Furuta, Kazuo

    2000-01-01

    On the JCO criticality accident occurred on September 30, 1999, from relatively earlier time since its occurrence it was elucidated that it was formed not by accident and error operation of apparatus and instruments but by unsafe actions of operators beyond regular manual as its direct cause, and that an organizational factor on business managers and safety administration unable to control such unsafe actions of operators at its background. Then, it was judged to be essential to carry out an accident research from a viewpoint of the human factor (HF) for elucidation on essence of the accident, to establish a 'special workshop on the JCO accident research' to investigate elucidation of the accident cause and countermeasure of reoccurrence at a standpoint of HF. As a result, the essential cause of this accident was summarized that safety information such as ideals, information, teachings and so forth necessary for safety management were failed to share among different organizations. As a teaching of this accident, nuclear energy participants must recognize that safety culture is not finished only in specific organization and range but produced by protecting weathering of danger consciousness and effort of mutually exciting and learning by sharing a safety information beyond different organization, range and time. (G.K.)

  6. Criticality accident in uranium fuel processing plant. The estimation of the total number of fissions with related reactor physics parameters

    International Nuclear Information System (INIS)

    Nishina, Kojiro; Oyamatsu, Kazuhiro; Kondo, Shunsuke; Sekimoto, Hiroshi; Ishitani, Kazuki; Yamane, Yoshihiro; Miyoshi, Yoshinori

    2000-01-01

    This accident occurred when workers were pouring a uranium solution into a precipitation tank with handy operation against the established procedure and both the cylindrical diameter and the total mass exceeded the limited values. As a result, nuclear fission chain reactor in the solution reached not only a 'criticality' state continuing it independently but also an instantly forming criticality state exceed the criticality and increasing further nuclear fission number. The place occurring the accident at this time was not reactor but a place having not to form 'criticality' called by a processing process of uranium fuel. In such place, as because of relating to mechanism of chain reaction, it is required naturally for knowledge on the reactor physics, it is also necessary to understand chemical reaction in chemical process, and functions of tanks, valves and pumps mounted at the processes. For this purpose, some information on uranium concentration ratio, atomic density of nuclides largely affecting to chain reaction such as uranium, hydrogen, and so forth in the solution, shape, inner structure and size of container for the solution, and its temperature and total volume, were necessary for determining criticality volume of the accident uranium solution by using nuclear physics procedures. Here were described on estimation of energy emission in the JCO accident, estimation from analytical results on neutron and solution, calculation of various nuclear physics property estimation on the JCO precipitation tank at JAERI. (G.K.)

  7. JCO criticality accident termination operation

    International Nuclear Information System (INIS)

    Kanamori, Masashi

    2010-07-01

    In 2001, we summarized the circumstances surrounding termination of the JCO criticality accident based on testimony in the Mito District Court on December 17, 2001. JCO was the company for uranium fuels production in Japan. That document was assembled based on actual testimony in the belief that a description of the work involved in termination of the accident would be useful in some way for preventing nuclear disasters in the future. The description focuses on the witness' own behavior, and what he saw and heard, and thus is written from the perspective of action by one individual. This was done simply because it was easier for the witness to write down his memories as he remembers them. Description of the activities of other organizations and people is provided only as necessary, to ensure that consistency in the descriptive approach is not lost. The essentials of this report were rewritten as a third-person objective description in the summary of the report by the Atomic Energy Society of Japan (AESJ). Since then, comments have been received from sources such as former members of the Nuclear Safety Commission (Dr. Kenji Sumita and Dr. Akira Kanagawa), concerned parties from the former Science and Technology Agency, and reports from the JCO Criticality Accident Investigation Committee of the AESJ, and thus this report was rewritten to correct incorrect information, and add material where that was felt to be necessary. This year is the tenth year of the JCO criticality accident. To mark this occasion we have decided to translate the record of what occurred at the accident site into English so that more people can draw lessons from this accident. This report is an English version of JAEA-Technology 2009-073. (author)

  8. The detection of criticality accidents

    International Nuclear Information System (INIS)

    Prigent, R.; Renard, C.

    It is necessary to shield the personnel from the radiological consequences of a criticality accident. In the past ten years the study programmes have highlighted fresh data which have led to new thinking on the detection philosophy and as a consequence the design of detection equipment. Concurrently, new recommendations have been drawn up by the Safety Criticality Committee. The new detection equipment was developed by the CEA on the basis of the CRAC and SILENE experiments. Its industrialization was entrusted to the Intertechnique Company and the first network installed dates back to 1976. An examination is made of the problem of accident detection, dealing in turn with detection, the characteristics of the equipment and the installation rules. To clarify the various points discussed, a parallel has been drawn between the equipment existing up to 1975 and the new generation developed since then [fr

  9. Recommendations about criticality accident dosimetry

    International Nuclear Information System (INIS)

    1975-07-01

    The aims of criticality accident dosimetry and the characteristics peculiar to a critical burst being defined, the requirements to be fulfilled by a dosimetric system applied to this type of measurements are presented. The devices chosen by the C.E.A. Radiation Survey Divisions, simple and cheap, are described along with the main processes to be carried out in order to evaluate doses after an accident. The apparatus necessary for detector counting and the directions for use are presented in detail, allowing standardization of measurements. A set of linear formula enables to obtain, from these measurements, all required informations about neutron fluences and spectra, along with the suitable components of the dose at the irradiated people locations [fr

  10. Criticality accident of nuclear fuel facility. Think back on JCO criticality accident

    International Nuclear Information System (INIS)

    Naito, Keiji

    2003-09-01

    This book is written in order to understand the fundamental knowledge of criticality safety or criticality accident of nuclear fuel facility by the citizens. It consists of four chapters such as critical conditions and criticality accident of nuclear facility, risk of criticality accident, prevention of criticality accident and a measure at an occurrence of criticality accident. A definition of criticality, control of critical conditions, an aspect of accident, a rate of incident, damage, three sufferers, safety control method of criticality, engineering and administrative control, safety design of criticality, investigation of failure of safety control of JCO criticality accident, safety culture are explained. JCO criticality accident was caused with intention of disregarding regulation. It is important that we recognize the correct risk of criticality accident of nuclear fuel facility and prevent disasters. On the basis of them, we should establish safety culture. (S.Y.)

  11. A Review of Criticality Accidents 2000 Revision

    Energy Technology Data Exchange (ETDEWEB)

    Thomas P. McLaughlin; Shean P. Monahan; Norman L. Pruvost; Vladimir V. Frolov; Boris G. Ryazanov; Victor I. Sviridov

    2000-05-01

    Criticality accidents and the characteristics of prompt power excursions are discussed. Sixty accidental power excursions are reviewed. Sufficient detail is provided to enable the reader to understand the physical situation, the chemistry and material flow, and when available the administrative setting leading up to the time of the accident. Information on the power history, energy release, consequences, and causes are also included when available. For those accidents that occurred in process plants, two new sections have been included in this revision. The first is an analysis and summary of the physical and neutronic features of the chain reacting systems. The second is a compilation of observations and lessons learned. Excursions associated with large power reactors are not included in this report.

  12. A Review of Criticality Accidents 2000 Revision

    International Nuclear Information System (INIS)

    McLaughlin, Thomas P.; Monahan, Shean P.; Pruvost, Norman L.; Frolov, Vladimir V.; Ryazanov, Boris G.; Sviridov, Victor I.

    2000-01-01

    Criticality accidents and the characteristics of prompt power excursions are discussed. Sixty accidental power excursions are reviewed. Sufficient detail is provided to enable the reader to understand the physical situation, the chemistry and material flow, and when available the administrative setting leading up to the time of the accident. Information on the power history, energy release, consequences, and causes are also included when available. For those accidents that occurred in process plants, two new sections have been included in this revision. The first is an analysis and summary of the physical and neutronic features of the chain reacting systems. The second is a compilation of observations and lessons learned. Excursions associated with large power reactors are not included in this report

  13. Radiation monitoring using imaging plate technology: A case study of leaves affected by the Chernobyl nuclear power plant and JCO criticality accidents

    Directory of Open Access Journals (Sweden)

    Kimura Shinzo

    2006-01-01

    Full Text Available This paper describes the use of a photostimulable phosphor screen imaging technique to detect radioactive contamination in the leaves of wormwood (Artemisia vulgaris L and fern (Dryopteris filix-max CL. Schoff plants affected by the Chernobyl nuclear power plant accident. The imaging plate technology is well known for many striking performances in two-dimensional radiation detection. Since imaging plate comprises an integrated detection system, it has been extensively applied to surface contamination distribution studies. In this study, plant samples were collected from high- and low-contaminated areas of Ukraine and Belarus, which were affected due to the Chernobyl accident and exposed to imaging technique. Samples from the highly contaminated areas revealed the highest photo-stimulated luminescence on the imaging plate. Moreover, the radio nuclides detected in the leaves by gamma and beta ray spectroscopy were 137Cs and 90Sr, respectively. Additionally, in order to assess contamination, a comparison was also made with leaves of plants affected during the JCO criticality accident in Japan. Based on the results obtained, the importance of imaging plate technology in environmental radiation monitoring has been suggested.

  14. Accident prevention in power plants

    International Nuclear Information System (INIS)

    Steyrer, H.

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

  15. Workplace Accidents and Self-Organized Criticality

    OpenAIRE

    Mauro, John C.; Diehl, Brett; Marcellin, Richard F.; Vaughn, Daniel J.

    2018-01-01

    The occurrence of workplace accidents is described within the context of self-organized criticality, a theory from statistical physics that governs a wide range of phenomena across physics, biology, geosciences, economics, and the social sciences. Workplace accident data from the U.S. Bureau of Labor Statistics reveal a power-law relationship between the number of accidents and their severity as measured by the number of days lost from work. This power-law scaling is indicative of workplace a...

  16. Four years after the JCO criticality accident

    International Nuclear Information System (INIS)

    Sumita, Kenji

    2003-01-01

    It has been about four years since the first criticality accident in Japan. The JCO accident site was not so far from this auditorium. I have been asked to give a short review of important results from the various technical investigations on the accident that have been performed during the past four years. I will also give a short introduction to the changes that have been made in the nuclear safety regulation systems of the Japanese Government. (author)

  17. Dosimetric management during a criticality accident

    International Nuclear Information System (INIS)

    Lebaron-Jacobs, L.; Fottorino, R.; Racine, Y.; Miele, A.; Barbry, F.; Briot, F.; Distinguin, S.; Le Goff, J.P.; Berard, P.; Boisson, P.; Cavadore, D.; Lecoix, G.; Persico, M.H.; Rongier, E.; Challeton-De Vathaire, C.; Medioni, R.; Voisin, P.; Exmelin, L.; Flury-Herard, A.; Gaillard-Lecanu, E.; Lemaire, G.; Gonin, M.; Riasse, C.

    2008-01-01

    A working group from health occupational and clinical biochemistry services on French sites has issued essential data sheets on the guidelines to follow in managing the victims of a criticality accident. Since the priority of the medical management after a criticality accident is to assess the dose and the distribution of dose, some dosimetric investigations have been selected in order to provide a prompt response and to anticipate the final dose reconstruction. Comparison exercises between clinical biochemistry laboratories on French sites were carried out to confirm that each laboratory maintained the required operational methods for hair treatment and the appropriate equipment for 32 P activity in hair and 24 Na activity in blood measurements, and to demonstrate its ability to rapidly provide neutron dose estimates after a criticality accident. As a result, a relation has been assessed to estimate the dose and the distribution of dose according to the neutron spectrum following a criticality accident. (authors)

  18. Lessons learned from early criticality accidents

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1996-01-01

    Four accidents involving the approach to criticality occurred during the period July, 1945, through May, 1996. These have been described in the format of the OPERATING EXPERIENCE WEEKLY SUMMARY which is distributed by the Office of Nuclear and Facility Safety. Although the lessons learned have been incorporated in standards, codes, and formal procedures during the last fifty years, this is their first presentation in this format. It is particularly appropriate that they be presented in the forum of the Nuclear Criticality Technology Safety Project Workshop closest to the fiftieth anniversary of the last of the four accidents, and that which was most instrumental in demonstrating the need to incorporate lessons learned

  19. Criticality accident in uranium fuel processing plant. Efficacy of insurance and third party inspection system. Capability margins of insurance, in view of bitter experience at JCO

    Energy Technology Data Exchange (ETDEWEB)

    Izumi, Taiichiro [Songaihokenryoritsusanteikai (Japan)

    2000-08-01

    Among persons relating to safety engineering, most of them point out merit and demerit of safety theory in Japan since a long term before. Under national policy aiming at growth and expansion due to a policy under leading of government after war, running for about 50 years remained some strains containing a number of contradictions and absurdities in various fields. Here was described mainly on how to be done safety accident protection and inspection on a base of happenings incidentally seen at a chance of the criticality accident. Therefore, here were also established some viewpoints such as transferable risk, limit from insurance feature, genealogy of insurance, under-writing, and risk management, to mention effectiveness of the third party inspection with closed relationship with accident insurance. (G.K.)

  20. Criticality accident in uranium fuel processing plant. Efficacy of insurance and third party inspection system. Capability margins of insurance, in view of bitter experience at JCO

    International Nuclear Information System (INIS)

    Izumi, Taiichiro

    2000-01-01

    Among persons relating to safety engineering, most of them point out merit and demerit of safety theory in Japan since a long term before. Under national policy aiming at growth and expansion due to a policy under leading of government after war, running for about 50 years remained some strains containing a number of contradictions and absurdities in various fields. Here was described mainly on how to be done safety accident protection and inspection on a base of happenings incidentally seen at a chance of the criticality accident. Therefore, here were also established some viewpoints such as transferable risk, limit from insurance feature, genealogy of insurance, under-writing, and risk management, to mention effectiveness of the third party inspection with closed relationship with accident insurance. (G.K.)

  1. Medical consequences of a nuclear plant accident

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  2. Criticality accident dosimetry with ESR spectroscopy.

    Science.gov (United States)

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

    1996-01-01

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

  3. A neutron dosemeter for nuclear criticality accidents.

    Science.gov (United States)

    d'Errico, F; Curzio, G; Ciolini, R; Del Gratta, A; Nath, R

    2004-01-01

    A neutron dosemeter which offers instant read-out has been developed for nuclear criticality accidents. The system is based on gels containing emulsions of superheated dichlorodifluoromethane droplets, which vaporise into bubbles upon neutron irradiation. The expansion of these bubbles displaces an equivalent volume of gel into a graduated pipette, providing an immediate measure of the dose. Instant read-out is achieved using an array of transmissive optical sensors which consist of coupled LED emitters and phototransistor receivers. When the gel displaced in the pipette crosses the sensing region of the photomicrosensors, it generates a signal collected on a computer through a dedicated acquisition board. The performance of the device was tested during the 2002 International Accident Dosimetry Intercomparison in Valduc, France. The dosemeter was able to follow the initial dose gradient of a simulated accident, providing accurate values of neutron kerma; however, the emulsion was rapidly depleted of all its drops. A model of the depletion effects was developed and it indicates that an adequate dynamic range of the dose response can be achieved by using emulsions of smaller droplets.

  4. A neutron dosemeter for nuclear criticality accidents

    International Nuclear Information System (INIS)

    D'Errico, F.; Curzio, G.; Ciolini, R.; Del Gratta, A.; Nath, R.

    2004-01-01

    A neutron dosemeter which offers instant read-out has been developed for nuclear criticality accidents. The system is based on gels containing emulsions of superheated dichlorodifluoromethane droplets, which vaporise into bubbles upon neutron irradiation. The expansion of these bubbles displaces an equivalent volume of gel into a graduated pipette, providing an immediate measure of the dose. Instant read-out is achieved using an array of transmissive optical sensors which consist of coupled LED emitters and phototransistor receivers. When the gel displaced in the pipette crosses the sensing region of the photo microsensors, it generates a signal collected on a computer through a dedicated acquisition board. The performance of the device was tested during the 2002 International Accident Dosimetry Intercomparison in Valduc (France)). The dosemeter was able to follow the initial dose gradient of a simulated accident, providing accurate values of neutron kerma; however, the emulsion was rapidly depleted of all its drops. A model of the depletion effects was developed and it indicates that an adequate dynamic range of the dose response can be achieved by using emulsions of smaller droplets. (authors)

  5. Fukushima nuclear power plant accident was preventable

    Science.gov (United States)

    Kanoglu, Utku; Synolakis, Costas

    2015-04-01

    On 11 March 2011, the fourth largest earthquake in recorded history triggered a large tsunami, which will probably be remembered from the dramatic live pictures in a country, which is possibly the most tsunami-prepared in the world. The earthquake and tsunami caused a major nuclear power plant (NPP) accident at the Fukushima Dai-ichi, owned by Tokyo Electric Power Company (TEPCO). The accident was likely more severe than the 1979 Three Mile Island and less severe than the Chernobyl 1986 accidents. Yet, after the 26 December 2004 Indian Ocean tsunami had hit the Madras Atomic Power Station there had been renewed interest in the resilience of NPPs to tsunamis. The 11 March 2011 tsunami hit the Onagawa, Fukushima Dai-ichi, Fukushima Dai-ni, and Tokai Dai-ni NPPs, all located approximately in a 230km stretch along the east coast of Honshu. The Onagawa NPP was the closest to the source and was hit by an approximately height of 13m tsunami, of the same height as the one that hit the Fukushima Dai-ichi. Even though the Onagawa site also subsided by 1m, the tsunami did not reach to the main critical facilities. As the International Atomic Energy Agency put it, the Onagawa NPP survived the event "remarkably undamaged." At Fukushima Dai-ichi, the three reactors in operation were shut down due to strong ground shaking. The earthquake damaged all offsite electric transmission facilities. Emergency diesel generators (EDGs) provided back up power and started cooling down the reactors. However, the tsunami flooded the facilities damaging 12 of its 13 EDGs and caused a blackout. Among the consequences were hydrogen explosions that released radioactive material in the environment. It is unfortunately clear that TEPCO and Japan's principal regulator Nuclear and Industrial Safety Agency (NISA) had failed in providing a professional hazard analysis for the plant, even though their last assessment had taken place only months before the accident. The main reasons are the following. One

  6. Process criticality accident likelihoods, consequences and emergency planning

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1992-01-01

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

  7. Accidents at nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The accidents which accurred at Wuergassen, Browns Ferry and Three Mile Island are each briefly described and discussed. The last is naturally treated in much more detail than the first two. Damage to the fuel elements is briefly considered and the release of fission products, radiation doses to the population and their expected consequences are discussed. The accidents are evaluated and related to risk evaluations, especially in WASH-1400. (JIW)

  8. Plant critical concept

    International Nuclear Information System (INIS)

    O'Regan, P.J.

    1996-01-01

    The objective of the effort summarized in this paper is to support O and M cost reduction efforts by focusing resources on components and processes critical to plant performance. This effort will identify where resources on nonplant critical components and processes can be reduced or eliminated. This method will use a functional assessment as the basis for component-specific evaluations and ranking. This effort consists of two stages conducted in series. The first stage is to deterministically identify that set of plant components that are relevant from a plant performance perspective (i.e., safety, economics, reliability). The second stage probabilistically ranks that set of plant components from an importance perspective, where importance pertains to the particular application and is probabilistically weighted. The results of a pilot study identified that only a relatively small set of components are truly critical from an integrated plant performance perspective. These results are consistent with work being conducted at other nuclear power plants, as well as other commercial facilities. Initial implementation of this effort is estimated to reduce O and M costs on the order of $1 million per year. Subsequent applications are anticipated to increase that savings to $4--$5 million per year

  9. Criticality accident studies and methodology implemented at the CEA

    International Nuclear Information System (INIS)

    Barbry, Francis; Fouillaud, Patrick; Reverdy, Ludovic; Mijuin, Dominique

    2003-01-01

    Based on the studies and results of experimental programs performed since 1967 in the CRAC, then SILENE facilities, the CEA has devised a methodology for criticality accident studies. This methodology integrates all the main focuses of its approach, from criticality accident phenomenology to emergency planning and response, and thus includes aspects such as criticality alarm detector triggering, airborne releases, and irradiation risk assessment. (author)

  10. The criticality accident in Tokaimura and medical aspects of radiation emergency

    International Nuclear Information System (INIS)

    Chen Xiaohua; Mao Bingzhi

    2003-01-01

    A criticality accident occurred on September 30, 1999 at the uranium processing plant in Tokaimura Japan, which is the most severe accident since Chernobyl catastrophe. 213 people were exposed to radiation, among them 2 workers were exposed to 16-23 Gy and 6-10 Gy individually, one worker was 2 Gy, 2 people was 10 mSv and 208 person was 0-5 mSv. Author was invited to attend an international symposium on 'The Criticality Accident in Tokaimura Medical Aspects of Radiation Emergency' in Chiba Japan on December 2000. An overview of the accident, dose estimation and neutron relative biological effects are discussed in this article

  11. Improved dose estimates for nuclear criticality accidents

    International Nuclear Information System (INIS)

    Wilkinson, A.D.; Basoglu, B.; Bentley, C.L.; Dunn, M.E.; Plaster, M.J.; Dodds, H.L.; Yamamoto, T.

    1995-01-01

    Slide rules are improved for estimating doses and dose rates resulting from nuclear criticality accidents. The original slide rules were created for highly enriched uranium solutions and metals using hand calculations along with the decades old Way-Wigner radioactive decay relationship and the inverse square law. This work uses state-of-the-art methods and better data to improve the original slide rules and also to extend the slide rule concept to three additional systems; i.e., highly enriched (93.2 wt%) uranium damp (H/ 235 U = 10) powder (U 3 O 8 ) and low-enriched (5 wt%) uranium mixtures (UO 2 F 2 ) with a H/ 235 U ratio of 200 and 500. Although the improved slide rules differ only slightly from the original slide rules, the improved slide rules and also the new slide rules can be used with greater confidence since they are based on more rigorous methods and better nuclear data

  12. Severe accidents in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Valle Cepero, R.; Castillo Alvarez, J.; Ramon Fuente, J.

    1996-01-01

    For the assessment of the safety of nuclear power plants it is of great importance the analyses of severe accidents since they allow to estimate the possible failure models of the containment, and also permit knowing the magnitude and composition of the radioactive material that would be released to the environment in case of an accident upon population and the environment. This paper presents in general terms the basic principles for conducting the analysis of severe accidents, the fundamental sources in the generation of radionuclides and aerosols, the transportation and deposition processes, and also makes reference to de main codes used in the modulation of severe accidents. The final part of the paper contents information on how severe accidents are dialed with the regulatory point view in different countries

  13. Accidents with nuclear power plants, ch. 11

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    A recalculation of the consequences of nuclear power plant accidents is presented taking into account different parameters or different quantities than those usually accepted. A case study of a nuclear power plant planned for the Eems-river estuary in the Netherlands is presented

  14. Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

    International Nuclear Information System (INIS)

    Watanabe, Norio; Tamaki, Hitoshi

    2000-04-01

    On September 30, 1999, a criticality accident occurred at the Tokai-mura uranium processing plant operated by JCO Co., Ltd., which resulted in the first nuclear accident involving a fatality, in Japan, and forced the residents in the vicinity of the site to be evacuated and be sheltered indoors. There have now been 21 criticality accidents reported in nuclear fuel processing facilities in foreign countries: seven in the United States, one in the United Kingdom and thirteen in Russia. Most of them occurred during the period from mid-1950's to mid-1960's, but one criticality accident tool place in Russian in 1997. This report reviews and compiles the published information on these accidents, including the latest information, focusing on the event sequence, the consequence of accident, and the cause of accident. The observations from the reviews are summarized as follows: Twenty of the 21 accidents occurred with the fissile material in a liquid. Twenty of the 21 accidents occurred in vessels/tanks with unfavorable geometry but one occurred in the vessel with favorable geometry. There were seven fatalities that were involved in five accidents. Three accidents involved a re-criticality condition caused by inadequate operator actions and two of them led to the death of the operators. One accident reached a re-criticality condition several hours after the first excursion was terminated by injecting borated water into the affected vessel. This accident implies the possibility that the borated water injection might not be effective to the criticality termination due to solubility of boric acid. Mechanisms of the criticality termination vary as follows: ejection or splashing of the solution at the time of power excursion, boiling or evaporation, addition of neutron poisons, or manual draining of solutions. (author)

  15. Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Norio [Planning and Analysis Division, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Tamaki, Hitoshi [Department of Safety Research Technical Support, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2000-04-01

    On September 30, 1999, a criticality accident occurred at the Tokai-mura uranium processing plant operated by JCO Co., Ltd., which resulted in the first nuclear accident involving a fatality, in Japan, and forced the residents in the vicinity of the site to be evacuated and be sheltered indoors. There have now been 21 criticality accidents reported in nuclear fuel processing facilities in foreign countries: seven in the United States, one in the United Kingdom and thirteen in Russia. Most of them occurred during the period from mid-1950's to mid-1960's, but one criticality accident tool place in Russian in 1997. This report reviews and compiles the published information on these accidents, including the latest information, focusing on the event sequence, the consequence of accident, and the cause of accident. The observations from the reviews are summarized as follows: Twenty of the 21 accidents occurred with the fissile material in a liquid. Twenty of the 21 accidents occurred in vessels/tanks with unfavorable geometry but one occurred in the vessel with favorable geometry. There were seven fatalities that were involved in five accidents. Three accidents involved a re-criticality condition caused by inadequate operator actions and two of them led to the death of the operators. One accident reached a re-criticality condition several hours after the first excursion was terminated by injecting borated water into the affected vessel. This accident implies the possibility that the borated water injection might not be effective to the criticality termination due to solubility of boric acid. Mechanisms of the criticality termination vary as follows: ejection or splashing of the solution at the time of power excursion, boiling or evaporation, addition of neutron poisons, or manual draining of solutions. (author)

  16. Dose evaluation on the basis of {sup 24}Na activity in the human body for the criticality accident at JCO Tokai nuclear fuel processing plant

    Energy Technology Data Exchange (ETDEWEB)

    Momose, T.; Tsujimura, N.; Tasaki, T.; Kanai, K.; Hayashi, N.; Shinohara, K. [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2001-11-01

    {sup 24}Na in the human body, activated by neutrons emitted at the JCO criticality accident, was observed for 62 subjects, where 148 subjects were measured by the whole body counter of JNC Tokai Works. The 148 subjects, including JCO employees and the contractors, residents neighboring the site and emergency service officers, were measured by the whole-body counter. The neutron-energy spectrum around the facility was calculated using neutron transport codes (ANISN and MCNP), and the relation between an amount of activated sodium in human body and neutron dose was evaluated from the calculated neutron energy spectrum and theoretical neutron capture probability by the human body. The maximum {sup 24}Na activity in the body was 7.7 kBq (83 Bq({sup 24}Na)/g({sup 23}Na)) and the relevant effective dose equivalent was 47 mSv. (author)

  17. Dose evaluation on the basis of 24Na activity in the human body for the criticality accident at JCO Tokai nuclear fuel processing plant

    International Nuclear Information System (INIS)

    Momose, T.; Tsujimura, N.; Tasaki, T.; Kanai, K.; Hayashi, N.; Shinohara, K.

    2001-01-01

    24 Na in the human body, activated by neutrons emitted at the JCO criticality accident, was observed for 62 subjects, where 148 subjects were measured by the whole body counter of JNC Tokai Works. The 148 subjects, including JCO employees and the contractors, residents neighboring the site and emergency service officers, were measured by the whole-body counter. The neutron-energy spectrum around the facility was calculated using neutron transport codes (ANISN and MCNP), and the relation between an amount of activated sodium in human body and neutron dose was evaluated from the calculated neutron energy spectrum and theoretical neutron capture probability by the human body. The maximum 24 Na activity in the body was 7.7 kBq (83 Bq( 24 Na)/g( 23 Na)) and the relevant effective dose equivalent was 47 mSv. (author)

  18. Plant critical concept

    International Nuclear Information System (INIS)

    O'Regan, P.J.

    1995-01-01

    The achievement of operation and maintenance (O ampersand M) cost reductions is a prime concern for plant operators. Initiatives by the nuclear industry to address this concern are under way and/or in development. These efforts include plant reliability studies, reliability-centered maintenance, risk ranking and testing philosophies, performance-based testing philosophies, graded quality assurance, and so forth. This paper presents the results of an effort to develop a methodology that integrates and applies the common data and analysis requirements for a number of risk-based and performance-based initiatives. This initial phase of the effort applied the methodology and its results to two initiatives. These were the procurement function and the preventive maintenance function. This effort integrated multiple programs and functions to identify those components that are truly critical from an integrated plant performance perspective. The paper describes the scope of the effort, the development of a methodology to identify plant critical components, and the application of these results to the maintenance rule compliance, preventive maintenance, and procurement functions at the candidate plant

  19. Chemical dosimetry system for criticality accidents.

    Science.gov (United States)

    Miljanić, Saveta; Ilijas, Boris

    2004-01-01

    Ruder Bosković Institute (RBI) criticality dosimetry system consists of a chemical dosimetry system for measuring the total (neutron + gamma) dose, and a thermoluminescent (TL) dosimetry system for a separate determination of the gamma ray component. The use of the chemical dosemeter solution chlorobenzene-ethanol-trimethylpentane (CET) is based on the radiolytic formation of hydrochloric acid, which protonates a pH indicator, thymolsulphonphthalein. The high molar absorptivity of its red form at 552 nm is responsible for a high sensitivity of the system: doses in the range 0.2-15 Gy can be measured. The dosemeter has been designed as a glass ampoule filled with the CET solution and inserted into a pen-shaped plastic holder. For dose determinations, a newly constructed optoelectronic reader has been used. The RBI team took part in the International Intercomparison of Criticality Accident Dosimetry Systems at the SILENE Reactor, Valduc, June 2002, with the CET dosimetry system. For gamma ray dose determination TLD-700 TL detectors were used. The results obtained with CET dosemeter show very good agreement with the reference values.

  20. Accident analysis for nuclear power plants

    International Nuclear Information System (INIS)

    2002-01-01

    Deterministic safety analysis (frequently referred to as accident analysis) is an important tool for confirming the adequacy and efficiency of provisions within the defence in depth concept for the safety of nuclear power plants (NPPs). Owing to the close interrelation between accident analysis and safety, an analysis that lacks consistency, is incomplete or is of poor quality is considered a safety issue for a given NPP. Developing IAEA guidance documents for accident analysis is thus an important step towards resolving this issue. Requirements and guidelines pertaining to the scope and content of accident analysis have, in the past, been partially described in various IAEA documents. Several guidelines relevant to WWER and RBMK type reactors have been developed within the IAEA Extrabudgetary Programme on the Safety of WWER and RBMK NPPs. To a certain extent, accident analysis is also covered in several documents of the revised NUSS series, for example, in the Safety Requirements on Safety of Nuclear Power Plants: Design (NS-R-1) and in the Safety Guide on Safety Assessment and Verification for Nuclear Power Plants (NS-G-1.2). Consistent with these documents, the IAEA has developed the present Safety Report on Accident Analysis for Nuclear Power Plants. Many experts have contributed to the development of this Safety Report. Besides several consultants meetings, comments were collected from more than fifty selected organizations. The report was also reviewed at the IAEA Technical Committee Meeting on Accident Analysis held in Vienna from 30 August to 3 September 1999. The present IAEA Safety Report is aimed at providing practical guidance for performing accident analyses. The guidance is based on present good practice worldwide. The report covers all the steps required to perform accident analyses, i.e. selection of initiating events and acceptance criteria, selection of computer codes and modelling assumptions, preparation of input data and presentation of the

  1. A critical assessment of energy accident studies

    International Nuclear Information System (INIS)

    Felder, Frank A.

    2009-01-01

    A comparison of two studies conducted ten years apart on energy accidents provides important insights into methodological issues and policy implications. Recommendations for further improvements in energy accident studies are developed including accounting for differences between average and incremental accident damages, testing for appropriate levels of aggregation of accidents, making references and databases publicly available, more precisely defining and reporting different types of economic damages, accounting for involuntary and voluntary risks, reporting normalized damages, raising broader public policy and planning implications and updating existing accident databases.

  2. A critical assessment of energy accident studies

    Energy Technology Data Exchange (ETDEWEB)

    Felder, Frank A. [Edward J. Bloustein School of Planning and Public Policy, Rutgers, The State University of New Jersey, 33 Livingston Avenue, New Brunswick, NJ 08901 (United States)

    2009-12-15

    A comparison of two studies conducted ten years apart on energy accidents provides important insights into methodological issues and policy implications. Recommendations for further improvements in energy accident studies are developed including accounting for differences between average and incremental accident damages, testing for appropriate levels of aggregation of accidents, making references and databases publicly available, more precisely defining and reporting different types of economic damages, accounting for involuntary and voluntary risks, reporting normalized damages, raising broader public policy and planning implications and updating existing accident databases. (author)

  3. Contribution to evaluating nuclear power plant accidents

    International Nuclear Information System (INIS)

    Razga, J.; Horacek, P.

    1990-01-01

    Large-scale accidents pose the highest risk in the use of nuclear power. They are the major factor that has to be taken into account when assessing the effect of nuclear power plants on human health and on the environment. In Czechoslovak conditions, the effectiveness of provisions made to reduce the hazard of large-scale nuclear power plant accidents must be considered from the following aspects: effect on human health, consequences of long-term disabling of the infrastructure, potential of human and material reserves in coping with the accident, consequences of power failure for the electricity system, effect on agricultural production and catering, risk of ground and surface water contamination in the Labe or Danube river basin, and international political aspects. (Z.M.). 3 tabs., 18 refs

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

    Energy Technology Data Exchange (ETDEWEB)

    McLaughlin, T.P.

    1991-01-01

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

  5. Installation places of criticality accident detectors in the plutonium conversion development facility

    International Nuclear Information System (INIS)

    Sanada, Yukihisa; Tsujimura, Norio; Shimizu, Yoshio; Izaki, Kenji; Furuta, Sadaaki

    2008-01-01

    At the Plutonium Conversion Development Facility (PCDF) in the Nuclear Fuel Cycle Engineering Laboratories, the co-conversion technologies to purify the mixed plutonium and uranium nitrate solution discharged from a reprocessing plant have been developed. The probability of a criticality accident in PCDF is extremely low. However, the criticality accident alarm system (CAAS) has been in place since 1982 to reduce the radiation dose to workers in case of such a rare criticality accident. The CAAS contains criticality accident detector units (CADs), one unit consisting of three plastic scintillation detectors, and using the 2 out of 3 voting system for the purpose of high reliability. Currently, eight CADs are installed in PCDF evaluating the dose using a simple equation allowing for a safety margin. The purpose of this study is to show the determination procedures for the adequate relocation of the CADs which adequately ensures safety in PCDF. (author)

  6. Review of the CRAC and SILENE Criticality Accident Studies

    International Nuclear Information System (INIS)

    Barbry, F.; Fouillaud, P.; Grivot, P.; Reverdy, L.

    2009-01-01

    In 1967, the Commissariat et l'Energie Atomique (French Atomic Energy Agency) performed its first research on criticality accidents for the purpose of limiting their impact on people, the environment, and nuclear facilities themselves. A criticality accident is accompanied by intense neutron and gamma emissions and release of radioactive fission products-gases and aerosols-gene rating risk of irradiation and contamination. This work has supplemented earlier work in criticality safety, which concentrated on critical mass measurements and computations. Understanding of the consequences of criticality accidents was limited. Emergency planning was hampered by lack of data. Information became available from pulsed reactor experiments, but the experiments were restricted to the established reactor configurations. The objectives of research performed at the Valduc criticality laboratory, mainly on aqueous fissile media, using the CRAC and SILENE facilities, by multidisciplinary teams of physicists, dosimetry specialists, and radio-biologists, were to model criticality accident physics, estimate irradiation risks and radioactive releases, detect excursions, and organize emergency response. The results of the Valduc experiments have contributed toward improved understanding of criticality accident phenomenology and better evaluation of the risks associated with such accidents. (authors)

  7. Review of the CRAC and SILENE Criticality Accident Studies

    Energy Technology Data Exchange (ETDEWEB)

    Barbry, F.; Fouillaud, P.; Grivot, P.; Reverdy, L. [CEA Valduc, Serv Rech Neutron and Critcite, 21 - Is-sur-Tille (France)

    2009-02-15

    In 1967, the Commissariat et l'Energie Atomique (French Atomic Energy Agency) performed its first research on criticality accidents for the purpose of limiting their impact on people, the environment, and nuclear facilities themselves. A criticality accident is accompanied by intense neutron and gamma emissions and release of radioactive fission products-gases and aerosols-gene rating risk of irradiation and contamination. This work has supplemented earlier work in criticality safety, which concentrated on critical mass measurements and computations. Understanding of the consequences of criticality accidents was limited. Emergency planning was hampered by lack of data. Information became available from pulsed reactor experiments, but the experiments were restricted to the established reactor configurations. The objectives of research performed at the Valduc criticality laboratory, mainly on aqueous fissile media, using the CRAC and SILENE facilities, by multidisciplinary teams of physicists, dosimetry specialists, and radio-biologists, were to model criticality accident physics, estimate irradiation risks and radioactive releases, detect excursions, and organize emergency response. The results of the Valduc experiments have contributed toward improved understanding of criticality accident phenomenology and better evaluation of the risks associated with such accidents. (authors)

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

    International Nuclear Information System (INIS)

    Nakajima, Ken

    2003-01-01

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

  9. Nuclear power plant Severe Accident Research Plan

    International Nuclear Information System (INIS)

    Larkins, J.T.; Cunningham, M.A.

    1983-01-01

    The Severe Accident Research Plan (SARP) will provide technical information necessary to support regulatory decisions in the severe accident area for existing or planned nuclear power plants, and covers research for the time period of January 1982 through January 1986. SARP will develop generic bases to determine how safe the plants are and where and how their level of safety ought to be improved. The analysis to address these issues will be performed using improved probabilistic risk assessment methodology, as benchmarked to more exact data and analysis. There are thirteen program elements in the plan and the work is phased in two parts, with the first phase being completed in early 1984, at which time an assessment will be made whether or not any major changes will be recommended to the Commission for operating plants to handle severe accidents. Additionally at this time, all of the thirteen program elements in Chapter 5 will be reviewed and assessed in terms of how much additional work is necessary and where major impacts in probabilistic risk assessment might be achieved. Confirmatory research will be carried out in phase II to provide additional assurance on the appropriateness of phase I decisions. Most of this work will be concluded by early 1986

  10. Bayesian methods for chromosome dosimetry following a criticality accident

    International Nuclear Information System (INIS)

    Brame, R.S.; Groer, P.G.

    2003-01-01

    Radiation doses received during a criticality accident will be from a combination of fission spectrum neutrons and gamma rays. It is desirable to estimate the total dose, as well as the neutron and gamma doses. Present methods for dose estimation with chromosome aberrations after a criticality accident use point estimates of the neutron to gamma dose ratio obtained from personnel dosemeters and/or accident reconstruction calculations. In this paper a Bayesian approach to dose estimation with chromosome aberrations is developed that allows the uncertainty of the dose ratio to be considered. Posterior probability densities for the total and the neutron and gamma doses were derived. (author)

  11. Detection of criticality accidents. The Intertechnique EDAC II system

    International Nuclear Information System (INIS)

    Prigent, R.

    1991-01-01

    The chief aim of the new generation of EDAC II criticality accidents detection system is to reduce the risks associated to the handling of fissile material by providing a swift and safe warning of the development of any criticality accident. To this function already devolving on the EDAC system of the previous generation, the EDAC II adds the possibility of storing in memory the characteristics of the accident, providing a daily follow-up of the striking events in the system through the print-out of a log book and providing assistance to the operators during the periodical tests. (Author)

  12. Process criticality accident likelihoods, magnitudes and emergency planning. A focus on solution accidents

    International Nuclear Information System (INIS)

    McLaughlin, Thomas P.

    2003-01-01

    This paper presents analyses and applications of data from reactor and critical experiment research on the dynamics of nuclear excursions in solution media. Available criticality accident information is also discussed and shown to provide strong evidence of the overwhelming likelihood of accidents in liquid media over other forms and to support the measured data. These analyses are shown to provide valuable insights into key parameters important to understanding solution excursion dynamics in general and in evaluating practical upper bounds on criticality accident magnitudes. This understanding and these upper bounds are directly applicable to the evaluation of the consequences of postulated criticality accidents. These bounds are also essential in order to comply with national and international consensus standards and regulatory requirements for emergency planning. (author)

  13. Accident at the Three Mile Island Nuclear Power Plant

    International Nuclear Information System (INIS)

    Bajusz, J.; Vamos, G.

    1979-01-01

    A short description of the TMI power plant is given. The course of events leading to the reactor accident and that of the first two weeks is described. The effect on the environment is estimated. The reasons and consequences of the accident are analysed. The probability of such an accident at the Paks Nuclear Power Plant is estimated. (R.J.)

  14. Neutron personal dosimetry in criticality accidents

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  15. Severe accident considerations for modern KWU-PWR plants

    International Nuclear Information System (INIS)

    Eyink, J.

    1987-01-01

    In assumption of severe accident on modern KWU-PWR plants the author discusses on the: selection of core meltdown sequences, course of the accident, containment behaviour and source terms for fission products release to the environment

  16. Updated tool for nuclear criticality accident emergency response

    International Nuclear Information System (INIS)

    Broadhead, B.L.; Hopper, C.M.

    1995-01-01

    Some 20 yr ago a hand-held slide rule was developed at the Oak Ridge Y-12 Plant to aid in the response to several postulated nuclear criticality accidents. These assumed accidents involved highly enriched uranium in either a bare metal or a uranyl nitrate system. The slide rule consisted of a sliding scale based on the total fission yield and four corresponding dose indicators: (1) a prompt radiation dose relationship as a function of distance; (2) a delayed fission product gamma dose rate relationship as a function of time and distance; (3) the total dose relationship with time and distance; and (4) the I-min integrated dose relationship with time and distance. The original slide rule was generated assuming very simplistic numerical procedures such as the inverse-square relationship of dose with distance and the Way-Wigner relationship to express the time dependence of the dose. The simple prescriptions were tied to actual dose measurements from similar systems to yield a meaningful, yet simple approach to emergency planning and response needs. This paper describes the application of an advanced procedure to the updating of the original slide rule for five critical systems. These five systems include (a) an unreflected sphere of 93.2 wt% enriched uranium metal, (b) an unreflected sphere of 93.2 wt% enriched uranyl nitrate solution with a H/ 235 U ratio of 500, (c) an unreflected sphere of damp 93.2 wt% enriched uranium oxide with a H/ 235 U ratio of 10, (d) an unreflected sphere of 4.95 wt% enriched uranyl fluoride solution having a H/ 235 U ratio of 410, and (e) an unreflected sphere of damp 5 wt% enriched uranium dioxide having a H/ 235 U ratio of 200

  17. Cognitive systems engineering analysis of the JCO criticality accident

    International Nuclear Information System (INIS)

    Tanabe, Fumiya; Yamaguchi, Yukichi

    2000-01-01

    The JCO Criticality Accident is analyzed with a framework based on cognitive systems engineering. With the framework, analysis is conducted integrally both from the system viewpoint and actors viewpoint. The occupational chemical risk was important as safety constraint for the actors as well as the nuclear risk, which is due to criticality accident, to the public and to actors. The inappropriate actor's mental model of the work system played a critical role and several factors (e.g. poor training and education, lack of information on criticality safety control in the procedures and instructions, and lack of warning signs at workplace) contributed to form and shape the mental model. Based on the analysis, several countermeasures, such as warning signs, information system for supporting actors and improved training and education, are derived to prevent such an accident. (author)

  18. Development of the criticality accident analysis code, AGNES

    International Nuclear Information System (INIS)

    Nakajima, Ken

    1989-01-01

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

  19. Development of likelihood estimation method for criticality accidents of mixed oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Tamaki, Hitoshi; Yoshida, Kazuo; Kimoto, Tatsuya; Hamaguchi, Yoshikane

    2010-01-01

    A criticality accident in a MOX fuel fabrication facility may occur depending on several parameters, such as mass inventory and plutonium enrichment. MOX handling units in the facility are designed and operated based on the double contingency principle to prevent criticality accidents. Control failures of at least two parameters are needed for the occurrence of criticality accident. To evaluate the probability of such control failures, the criticality conditions of each parameter for a specific handling unit are necessary for accident scenario analysis to be clarified quantitatively with a criticality analysis computer code. In addition to this issue, a computer-based control system for mass inventory is planned to be installed into MOX handling equipment in a commercial MOX fuel fabrication plant. The reliability analysis is another important issue in evaluating the likelihood of control failure caused by software malfunction. A likelihood estimation method for criticality accident has been developed with these issues been taken into consideration. In this paper, an example of analysis with the proposed method and the applicability of the method are also shown through a trial application to a model MOX fabrication facility. (author)

  20. Program for rapid dose assessment in criticality accident, RADAPAS

    International Nuclear Information System (INIS)

    Takahashi, Fumiaki

    2006-09-01

    In a criticality accident, a person near fissile material can receive extremely high dose which can cause acute health effect. For such a case, medical treatment should be carried out for the exposed person, according to severity of the exposure. Then, radiation dose should be rapidly assessed soon after an outbreak of an accident. Dose assessment based upon the quantity of induced 24 Na in human body through neutron exposure is expected as one of useful dosimetry techniques in a criticality accident. A dose assessment program, called RADAPAS (RApid Dose Assessment Program from Activated Sodium in Criticality Accidents), was therefore developed to assess rapidly radiation dose to exposed persons from activity of induced 24 Na. RADAPAS consists of two parts; one is a database part and the other is a part for execution of dose calculation. The database contains data compendiums of energy spectra and dose conversion coefficients from specific activity of 24 Na induced in human body, which had been derived in a previous analysis using Monte Carlo calculation code. Information for criticality configuration or characteristics of radiation in the accident field is to be interactively given with interface displays in the dose calculation. RADAPAS can rapidly derive radiation dose to the exposed person from the given information and measured 24 Na specific activity by using the conversion coefficient in database. This report describes data for dose conversions and dose calculation in RADAPAS and explains how to use the program. (author)

  1. Risks of potential accidents of nuclear power plants in Europe

    NARCIS (Netherlands)

    Slaper H; Eggink GJ; Blaauboer RO

    1993-01-01

    Over 200 nuclear power plants for commercial electricity production are presently operational in Europe. The 1986 accident with the nuclear power plant in Chernobyl has shown that severe accidents with a nuclear power plant can lead to a large scale contamination of Europe. This report is focussed

  2. Dose evaluation on the basis of {sup 24}Na activity in the human body for the criticality accident at JCO Tokai nuclear fuel processing plant

    Energy Technology Data Exchange (ETDEWEB)

    Momose, T.; Tsujimura, N.; Tasaki, T.; Kanai, K.; Hayashi, N.; Shinohara, K. [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2000-07-01

    Sodium-24({sup 24}Na) generated in human body due to neutron activation was measured by whole body counter (WBC) in JNC Tokai works. Total 148 persons (JCO employees and contractor, public member, fire fighters, etc.) were measured and {sup 24}Na was detected in the 62 persons. Neutron energy spectrum around the facility was calculated using ANISN and MCNP code and estimated mean capture probability {xi} of neutron for human body at this accident was around 0.25-0.28 at any distance from the center of the precipitation tank. Effective dose equivalent for the 62 persons were estimated based on the calculated conversion factors from {sup 24}Na specific activity to neutron dose. Maximum {sup 24}Na activity was 7.7 kBq (83 Bq({sup 24}Na)/g({sup 23}Na)) in total body and the evaluated effective dose equivalent was 47 mSv. (author)

  3. PNNL Results from 2010 CALIBAN Criticality Accident Dosimeter Intercomparison Exercise

    International Nuclear Information System (INIS)

    Hill, Robin L.; Conrady, Matthew M.

    2011-01-01

    This document reports the results of the Hanford personnel nuclear accident dosimeter (PNAD) and fixed nuclear accident dosimeter (FNAD) during a criticality accident dosimeter intercomparison exercise at the CEA Valduc Center on September 20-23, 2010. Pacific Northwest National Laboratory (PNNL) participated in a criticality accident dosimeter intercomparison exercise at the Commissariat a Energie Atomique (CEA) Valduc Center near Dijon, France on September 20-23, 2010. The intercomparison exercise was funded by the U.S. Department of Energy, Nuclear Criticality Safety Program, with Lawrence Livermore National Laboratory as the lead Laboratory. PNNL was one of six invited DOE Laboratory participants. The other participating Laboratories were: Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Savannah River Site (SRS), the Y-12 National Security Complex at Oak Ridge, and Sandia National Laboratory (SNL). The goals of PNNL's participation in the intercomparison exercise were to test and validate the procedures and algorithm currently used for the Hanford personnel nuclear accident dosimeters (PNADs) on the metallic reactor, CALIBAN, to test exposures to PNADs from the side and from behind a phantom, and to test PNADs that were taken from a historical batch of Hanford PNADs that had varying degrees of degradation of the bare indium foil. Similar testing of the PNADs was done on the Valduc SILENE test reactor in 2009 (Hill and Conrady, 2010). The CALIBAN results are reported here.

  4. A second simulated criticality accident dosimetry experiment

    CERN Document Server

    Adams, N

    1973-01-01

    This experiment was undertaken to facilitate training in criticality dose assessment by UKAEA and BNFL establishments with potential criticality hazards. Personal dosemeters, coins, samples of hair, etc. supplied by the seven participating establishments were attached to a man-phantom filled with a solution of sodium nitrate (simulating 'body-sodium'), and exposed to a burst of radiation from the AWRE pulsed reactor VIPER. The neutron and photon doses were each several hundred rads. Participants made two sets of dose assessments. The first, made solely from the evidence of their routine dosemeters the activation of body-sodium and standard monitoring data, simulated the initial dose assessment that would be made before the circumstances of a real incident were established. The second was made when the position and orientation of the phantom relative to the reactor and the shielding (20 cm of copper) between the reactor core and the phantom were disclosed. Neutron and photon dose assessments for comparison wit...

  5. Criticality accident studies and research performed in the Valduc criticality laboratory, France

    International Nuclear Information System (INIS)

    Barbry, F.; Fouillaud, P.

    2001-01-01

    In 1967, the IPSN (Institut de Protection et de Surete Nucleaire - Nuclear Protection and Safety Institute) started studies and research in France on criticality accidents, with the objective of improving knowledge and modelling of accidents in order to limit consequences to the public, the environment and installations. The criticality accident is accompanied by an intense emission of neutronic and gamma radiation and releases of radioactive products in the form of gas and aerosols, generating irradiation and contamination risks. The main objectives of the studies carried out, particularly using the CRAC installation and the SILENE reactor at Valduc (France), were to model the physics of criticality accidents, to estimate the risks of irradiation and radioactive releases, to elaborate an accident detection system and to provide information for intervention plans. This document summarizes the state of knowledge in the various fields mentioned above. The results of experiments carried out in the Valduc criticality laboratory are used internationally as reference data for the qualification of calculation codes and the assessment of the consequences of a criticality accident. The SILENE installation, that reproduces the various conditions encountered during a criticality accident, is also a unique international research tool for studies and training on those matters. (author)

  6. Study Of Severe Accident Phenomena In Nuclear Power Plant

    International Nuclear Information System (INIS)

    Sugiyanto; Antariksawan; Anhar, R.; Arifal

    2001-01-01

    Several phenomena that occurred in the light water reactor type of nuclear power plant during severe accident were studied. The study was carried out based on the results of severe accident researches in various countries. In general, severe accident phenomena can be classified into in-vessel phenomena, retention in the reactor coolant system, and ex-vessel phenomena. In-vessel retention has been recommended as a severe accident management strategy

  7. Comparison of the dose evaluation methods for criticality accident

    International Nuclear Information System (INIS)

    Shimizu, Yoshio; Oka, Tsutomu

    2004-01-01

    The improvement of the dose evaluation method for criticality accidents is important to rationalize design of the nuclear fuel cycle facilities. The source spectrums of neutron and gamma ray of a criticality accident depend on the condition of the source, its materials, moderation, density and so on. The comparison of the dose evaluation methods for a criticality accident is made. Some methods, which are combination of criticality calculation and shielding calculation, are proposed. Prompt neutron and gamma ray doses from nuclear criticality of some uranium systems have been evaluated as the Nuclear Criticality Slide Rule. The uranium metal source (unmoderated system) and the uranyl nitrate solution source (moderated system) in the rule are evaluated by some calculation methods, which are combinations of code and cross section library, as follows: (a) SAS1X (ENDF/B-IV), (b) MCNP4C (ENDF/B-VI)-ANISN (DLC23E or JSD120), (c) MCNP4C-MCNP4C (ENDF/B-VI). They have consisted of criticality calculation and shielding calculation. These calculation methods are compared about the tissue absorbed dose and the spectrums at 2 m from the source. (author)

  8. Case examples of chemical plant accidents. What we learn from them?

    International Nuclear Information System (INIS)

    Nakamura, Masayoshi

    2009-01-01

    Lessons learned from the JCO Nuclear Criticality Accident of 30 September 1999 in a uranium conversion test plant in Tokai-mura, Japan, are reviewed by referring some pertinent matters from the official report of this accident to remind of the universal characteristics among possible accidents of chemical plants. The paper discusses the responsibility of the establishment or institution to the demand alternation or request change from the client, how to respond to the proposal arising from the factory floor, and the safety control system of every-day maintenance of the factory which are important to prevent accidents in chemical plants. After explaining a background leading to the JCO accident, the author summarizes the lessons as follows: (1) changeable control system, (2) perfect provision of the manual considering the actual condition, and (3) clarification of the roles each played by the managers and the workers are most necessary and important. (S. Ohno)

  9. Recalibration of indium foil for personnel screening in criticality accidents.

    Science.gov (United States)

    Takada, C; Tsujimura, N; Mikami, S

    2011-03-01

    At the Nuclear Fuel Cycle Engineering Laboratories of the Japan Atomic Energy Agency (JAEA), small pieces of indium foil incorporated into personal dosemeters have been used for personnel screening in criticality accidents. Irradiation tests of the badges were performed using the SILENE reactor to verify the calibration of the indium activation that had been made in the 1980s and to recalibrate them for simulated criticalities that would be the most likely to occur in the solution process line. In addition, Monte Carlo calculations of the indium activation using the badge model were also made to complement the spectral dependence. The results lead to a screening level of 15 kcpm being determined that corresponds to a total dose of 0.25 Gy, which is also applicable in posterior-anterior exposure. The recalibration based on the latest study will provide a sounder basis for the screening procedure in the event of a criticality accident.

  10. Modeling of criticality accidents and their environmental consequences

    International Nuclear Information System (INIS)

    Thomas, W.; Gmal, B.

    1987-01-01

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

  11. ANS-8.23: Criticality accident emergency planning and response

    International Nuclear Information System (INIS)

    Pruvost, N.L.

    1991-01-01

    A study group has been formed under the auspices of ANS-8 to examine the need for a standard on nuclear criticality accident emergency planning and response. This standard would be ANS-8.23. ANSI/ANS-8.19-1984, Administrative Practices for Nuclear Criticality Safety, provides some guidance on the subject in Section 10 titled -- Planned Response to Nuclear Criticality Accidents. However, the study group has formed a consensus that Section 10 is inadequate in that technical guidance in addition to administrative guidance is needed. The group believes that a new standard which specifically addresses emergency planning and response to a perceived criticality accident is needed. Plans for underway to request the study group be designated a writing group to create a draft of such a new standard. The proposed standard will divide responsibility between management and technical staff. Generally, management will be charged with providing the necessary elements of emergency planning such as a criticality detection and alarm system, training, safe evacuation routes and assembly areas, a system for timely accountability of personnel, and an effective emergency response organization. The technical staff, on the other hand, will be made responsible for establishing specific items such as safe and clearly posted evacuation evacuation routes and dose criteria for personnel assembly areas. The key to the question of responsibilities is that management must provide the resources for the technical staff to establish the elements of an emergency response effort

  12. Seismic isolation of plants at risk of a severe accident

    International Nuclear Information System (INIS)

    Forni, Massimo

    2015-01-01

    More and more devastating earthquakes struck every year our planet. Many of these, though occurring in areas considered at high risk of earthquakes, far exceed the levels required by law. The industrial plants subjected to risk of severe accident, in particular petrochemical and nuclear power plants, are particularly exposed to this risk because of the number and the complexity of the structures and critical components of which they are composed. For this type of structures, anti-seismic techniques able to provide complete protection, even in case of unforeseen events, are needed. Seismic isolation is certainly the most promising technology of modern antiseismic as it allows not only to significantly reduce the dynamic load acting on the structures in case of seismic attack, but to provide safety margins against violent earthquakes, exceeding the assumed maximum design limit. [it

  13. Simple estimate of fission rate during JCO criticality accident

    Energy Technology Data Exchange (ETDEWEB)

    Oyamatsu, Kazuhiro [Faculty of Studies on Contemporary Society, Aichi Shukutoku Univ., Nagakute, Aichi (Japan)

    2000-03-01

    The fission rate during JCO criticality accident is estimated from fission-product (FP) radioactivities in a uranium solution sample taken from the preparation basin 20 days after the accident. The FP radioactivity data are taken from a report by JAERI released in the Accident Investigation Committee. The total fission number is found quite dependent on the FP radioactivities and estimated to be about 4x10{sup 16} per liter, or 2x10{sup 18} per 16 kgU (assuming uranium concentration 278.9 g/liter). On the contrary, the time dependence of the fission rate is rather insensitive to the FP radioactivities. Hence, it is difficult to determine the fission number in the initial burst from the radioactivity data. (author)

  14. Simple estimate of fission rate during JCO criticality accident

    International Nuclear Information System (INIS)

    Oyamatsu, Kazuhiro

    2000-01-01

    The fission rate during JCO criticality accident is estimated from fission-product (FP) radioactivities in a uranium solution sample taken from the preparation basin 20 days after the accident. The FP radioactivity data are taken from a report by JAERI released in the Accident Investigation Committee. The total fission number is found quite dependent on the FP radioactivities and estimated to be about 4x10 16 per liter, or 2x10 18 per 16 kgU (assuming uranium concentration 278.9 g/liter). On the contrary, the time dependence of the fission rate is rather insensitive to the FP radioactivities. Hence, it is difficult to determine the fission number in the initial burst from the radioactivity data. (author)

  15. CARNAC, Neutron Flux and Neutron Spectra in Criticality Accident

    International Nuclear Information System (INIS)

    Bessis, J.

    1976-01-01

    Nature of physical problem solved: Calculation of flux and neutron spectra in the case of a criticality accident. The method is unsophisticated but fast. The program is divided into two parts: (1) The code CRITIC is based on the Fermi age equation and evaluates the neutron number per fission emitted from a moderate critical system and its energy spectrum. (2) The code NARCISSE uses concrete current albedo, evaluates the product of neutron reflection on walls of the source containment and calculates the resulting flux at any point, and its energy distribution into 21 groups. The results obtained seem satisfactory, if compared with a Monte Carlo program

  16. Evaluation of coverage of enriched UF6 cylinder storage lots by existing criticality accident alarms

    International Nuclear Information System (INIS)

    Lee, B.L. Jr.; Dobelbower, M.C.; Woollard, J.E.; Sutherland, P.J.; Tayloe, R.W. Jr.

    1995-03-01

    The Portsmouth Gaseous Diffusion Plant (PORTS) is leased from the US Department of Energy (DOE) by the United States Enrichment Corporation (USEC), a government corporation formed in 1993. PORTS is in transition from regulation by DOE to regulation by the Nuclear Regulatory Commission (NRC). One regulation is 10 CFR Part 76.89, which requires that criticality alarm systems be provided for the site. PORTS originally installed criticality accident alarm systems in all building for which nuclear criticality accidents were credible. Currently, however, alarm systems are not installed in the enriched uranium hexafluoride (UF 6 ) cylinder storage lots. This report analyzes and documents the extent to which enriched UF 6 cylinder storage lots at PORTS are covered by criticality detectors and alarms currently installed in adjacent buildings. Monte Carlo calculations are performed on simplified models of the cylinder storage lots and adjacent buildings. The storage lots modelled are X-745B, X-745C, X745D, X-745E, and X-745F. The criticality detectors modelled are located in building X-343, the building X-344A/X-342A complex, and portions of building X-330 (see Figures 1 and 2). These criticality detectors are those located closest to the cylinder storage lots. Results of this analysis indicate that the existing criticality detectors currently installed at PORTS are largely ineffective in detecting neutron radiation from criticality accidents in most of the cylinder storage lots at PORTS, except sometimes along portions of their peripheries

  17. Severe accident management program at Cofrentes Nuclear Power Plant

    International Nuclear Information System (INIS)

    Borondo, L.; Serrano, C.; Fiol, M.J.; Sanchez, A.

    2000-01-01

    Cofrentes Nuclear Power Plant (GE BWR/6) has implemented its specific Severe Accident Management Program within this year 2000. New organization and guides have been developed to successfully undertake the management of a severe accident. In particular, the Technical Support Center will count on a new ''Severe Accident Management Team'' (SAMT) which will be in charge of the Severe Accident Guides (SAG) when Control Room Crew reaches the Emergency Operation Procedures (EOP) step that requires containment flooding. Specific tools and training have also been developed to help the SAMT to mitigate the accident. (author)

  18. Aerosols released in accidents in reprocessing plants

    International Nuclear Information System (INIS)

    Ballinger, M.Y.; Owczarski, P.C.; Hashimoto, K.; Nishio, G.; Jordan, S.; Lindner, W.

    1987-01-01

    For analyzing the thermodynamic and radiological consequences of solvent fire accidents in reprocessing plants, intensive investigations on burning contaminated condensible liquids were performed at Kernforschungszentrum Karlsruhe (KfK), Pacific Northwest Laboratory (PNL), and Japan Atomic Energy Research Institute (JAERI). In small- and large-scale tests, KfK studied the behavior of kerosene, tributyl phosphate, HNO 3 mixture fires in open air and closed containments. The particle release from uranium-contaminated pool fires was investigated. Different filter devices were tested. For analyzing fires, PNL has developed the FIRIN computer code and has generated small-scale fire data in support of that code. The results of the experiments in which contaminated combustible liquids were burned demonstrate the use of the FIRIN code in simulating a solvent fire in a nuclear reprocessing plant. To demonstrate the safety evaluation of a postulated solvent fire in an extraction process of a reprocessing pant, JAERI conducted large-scale fire tests. Behavior of solvent fires in a cell and the integrity of high-efficiency particulate air (HEPA) filters due to smoke plugging were investigated. To evaluate confinement of radioactive materials released from the solvent fire, the ventilation systems with HEPA filters were tested under postulated fire conditions

  19. Severe accidents: in nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    A ''severe'' nuclear accident refers to a reactor accident that could exceed reactor design specifications to such a degree as to prevent cooling of the reactor's core by normal means. This report summarizes the work of a NEA Senior Group of Experts who have studied the potential response of existing light-water reactors to severe accidents and have found that current designs of reactors are far more capable of coping with severe accidents than design specifications would suggest. The report emphasises the specific knowledge and means that can be used for diagnosing a severe accident and for managing its progression in order to prevent or mitigate its consequences

  20. JCO criticality accident as POST-LOCA: Poor structure induced loss of organizational control accident

    International Nuclear Information System (INIS)

    Furuhama, Yutaka

    2000-01-01

    Some problems in operation and business management of JCO (Japan Nuclear Fuel Conversion Co.) have been studied as background factors of the criticality accident. Open information about business conditions of JCO suggests that the cause of the accident is not so simple as to be attributed only to economic pressure, but includes immanent problems in JCO. We investigate the problems from five viewpoints, organization of safety management, system of operation management, activities for business improvement, risk awareness, and restructuring of business, and discuss the effects and causality of background factors as well as remedies for them. (author)

  1. Development of INCTAC code for analyzing criticality accident phenomena

    International Nuclear Information System (INIS)

    Mitake, Susumu; Hayashi, Yamato; Sakurai, Shungo

    2003-01-01

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

  2. Nuclear accidents and safety measures of domestic nuclear power plants

    International Nuclear Information System (INIS)

    Song Zurong; Che Shuwei; Pan Xiang

    2012-01-01

    Based on the design standards for the safety of nuclear and radiation in nuclear power plants, the three accidents in the history of nuclear power are analyzed. And the main factors for these accidents are found out, that is, human factors and unpredicted natural calamity. By combining the design and operation parameters of domestic nuclear plants, the same accidents are studied and some necessary preventive schemes are put forward. In the security operation technology of domestic nuclear power plants nowadays, accidents caused by human factors can by prevented completely. But the safety standards have to be reconsidered for the unpredicted neutral disasters. How to reduce the hazard of nuclear radiation and leakage to the level that can be accepted by the government and public when accidents occur under extreme conditions during construction and operation of nuclear power plants must be considered adequately. (authors)

  3. The need to study of bounding accident in reprocessing plant

    International Nuclear Information System (INIS)

    Segawa, Satoshi; Fujita, Kunio

    2013-01-01

    There is a clear consensus that the severe accident corresponds to the core damage accident for power reactors. On the other hand, for FCFs, there is no clear consensus on what is the accident to assess the safety in the region of beyond design basis, or what is the accident which has very low probability but large consequence. The need to examine a bounding consequence of each type of accident is explained to advance the rationality of safety management and regulation and, as a result, to reinforce the safety of a reprocessing plant. The likelihood of occurrence of an accident causing a bounding consequence should correspond to that of a severe accident at a nuclear power plant. The bounding consequence will be derived using the deterministic method and sound engineering judgment supplemented by the probabilistic method. Once an agreement on such a concept is reached among regulators, operators and related experts it will help to provide a solid basis to ensure the safety of a reprocessing plant independent of that of a nuclear power plant. In this paper, we show a preliminary risk profile of RRP calculated by QSA (Quantitative Safety Assessment) which JNFL developed. The profile shows that bounding consequences of various accidents in a range of occurrence frequency corresponding to a severe accident at a nuclear power plant. And we find that the bounding consequence of high-level liquid waste boiling is the largest among all in this range. Therefore, the risk of this event is shown in this paper as an example. To build a common consensus about bounding accidents among concerned parties will encourage regulatory body to introduce such an idea for more effective regulation with scientific rationality. Additionally the study of bounding accidents can contribute to substantial development for accident management strategy as reprocessing operators. (authors)

  4. Energy and angular responses of the criticality accident detector using a plastic scintillator

    International Nuclear Information System (INIS)

    Tsujimura, Norio; Yoshida, Tadayoshi

    2006-01-01

    The Japan Atomic Energy Agency (JAEA), Nuclear Fuel Cycle Engineering Laboratories, operates a spent fuel reprocessing plant and MOX (Plutonium-Uranium Mixed Oxide) fuel fabrication plants. Criticality accident detectors have been installed in these facilities. The detector, the Toshiba RD120, is composed of a plastic scintillator coupled to a photomultiplier tube, and an operational amplifier. The alarm triggering point is set to 1.0-3.6 mGy·h -1 in photon dose rate to detect the minimum accident of concern. However, a plastic scintillator is principally sensitive not only to primary photons but also to neutrons by secondary photons and heavy charged particles produced in the detector itself. The authors calculated energy and angular responses of the RD120 criticality accident detector to photons and neutrons using Monte Carlo computer codes. The response to primary photons was evaluated with the MCNP-4B and EGS4 calculations, and photon and X-ray irradiation experiments. The response to neutrons that produce secondary photons and heavy charged particles from neutron interactions was computed using the MCNP-4B and SCINFUL, respectively. As a result, reliable response functions were obtained. These results will be a great help in reassessing the coverage area and in determining the appropriate triggering dose rate level in criticality accidents. (author)

  5. The Tokai-mura JCO criticality accident and the activities of the accident countermeasure support team of Electric Power Companies, Japan

    International Nuclear Information System (INIS)

    Ogawa, Junko

    2000-01-01

    A criticality accident occurred at the JCO Tokai-mura nuclear fuel processing plant on September 30, 1999. This accident brought the damages which were unrivaled in the history of atomic energy development in Japan, seriously influencing the citizen life to such an extent as requesting for 320,000 inhabitants within 10 kilometers radius to stay indoors for as long as 18 hours. However, it could be said that though three workers suffered fatal injuries, no substantial hazards were made upon the regional inhabitants due to little release of radioactive substances. This video recorded the activities of the Accident Countermeasure Support Team of the Electric Power Companies immediately after the accident occurred, showing the chronological overview of the particulars of the accident. (author)

  6. Crisis, criticism, change: Regulatory reform in the wake of nuclear accidents

    International Nuclear Information System (INIS)

    Sexton, Kimberly A.; )

    2015-01-01

    Accidents are a forcing function for change in the nuclear industry. While these events can shed light on needed technical safety reforms, they can also shine a light on needed regulatory system reforms. The TEPCO Fukushima Daiichi nuclear power plant (NPP) accident in Japan is the most recent example of this phenomenon, but it is not the only one. In the wake of the three major accidents that have occurred in the nuclear power industry - Three Mile Island (TMI) in the United States; Chernobyl in Ukraine, in the former Soviet Union; and the Fukushima Daiichi NPP accident in Japan - a commission or committee of experts issued a report (or reports) with harsh criticism of the countries' regulatory system. And each of these accidents prompted changes in the respective regulatory systems. In looking at these responses, however, one must ask if this crisis, criticism, change approach is working and whether regulatory bodies around the world should instead undertake their own systematic reviews, un-prompted by crisis, to better ensure safety. This article will attempt to analyse the issue of regulatory reform in the wake of nuclear accidents by first providing a background in nuclear regulatory systems, looking to international and national legal frameworks. Next, the article will detail a cross-section of current regulatory systems around the world. Following that, the article will analyse the before and after of the regulatory systems in the United States, the Soviet Union and Japan in relation to the TMI, Chernobyl and Fukushima accidents. Finally, taking all this together, the article will address some of the international and national efforts to define exactly what makes a good regulator and provide conclusions on regulatory reform in the wake of nuclear accidents. (author)

  7. Triage and medical management of criticality accident victims

    International Nuclear Information System (INIS)

    Lebaron-Jacobs, L.; Flury-Herard, A.; Cavadore, D.

    2002-01-01

    The criticality accident is the result of an uncontrolled chain fission reaction initiated when the quantities of nuclear materials (uranium or plutonium)present accidentally exceed a given limit called the c ritical mass . As soon as the critical state is exceeded, the chain reaction increases exponentially. The result is a fast increase in the number of fission events which occur within the fissile medium. This phenomenon results in a release of energy mainly in the form of heat, accompanied by the intense emission of neutron and gamma radiation and the release of fission gases (Barby, 1983)

  8. Criticality analysis in uranium enrichment plant

    International Nuclear Information System (INIS)

    Okamoto, Tsuyoshi; Kiyose, Ryohei

    1977-01-01

    In a large scale uranium enrichment plant, uranium inventory in cascade rooms is not very large in quantity, but the facilities dealing with the largest quantity of uranium in that process are the UF 6 gas supply system and the blending system for controlling the product concentration. When UF 6 spills out of these systems, the enriched uranium is accumulated, and the danger of criticality accident is feared. If a NaF trap is placed at the forestage of waste gas treatment system, plenty of UF 6 and HF are adsorbed together in the NaF trap. Thus, here is the necessity of checking the safety against criticality. Various assumptions were made to perform the computation surveying the criticality of the system composed of UF 6 and HF adsorbed on NaF traps with WIMS code (transport analysis). The minimum critical radius resulted in about 53 cm in case of 3.5% enriched fuel for light water reactors. The optimum volume ratio of fissile material in the double salt UF 6 .2NaF and NaF.HF is about 40 vol. %. While, criticality survey computation was also made for the annular NaF trap having the central cooling tube, and it was found that the effect of cooling tube radius did not decrease the multiplication factor up to the cooling tube radius of about 5 cm. (Wakatsuki, Y.)

  9. Severe accident management: radiation dose control, Fukushima Daiichi and TMI-2 nuclear plant accidents

    International Nuclear Information System (INIS)

    Shaw, Roger

    2014-01-01

    This presentation presents valuable dose information related to the Fukushima Daiichi and Three Mile Island Unit 2 (TMI-2) Nuclear Plant accidents. Dose information is provided for what is well known for TMI-2, and what is available for Fukushima Daiichi. Particular emphasis is placed on the difference between the type of reactors involved, overarching plant damage issues, and radiation worker dose outcomes. For TMI-2, more in depth dose data is available for the accident and the subsequent recovery efforts. The comparisons demonstrate the need to understand the wide variation in potential dose management measures and outcomes for severe reactor accidents. (author)

  10. Severe accident management guidance for third Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Su Changsong

    2010-01-01

    The paper describes the background, document structure and the summaries of Severe Accident Management Guidance (SAMG) for Third Qinshan Nuclear Power Plant (TQNPP), and also introduces briefly some design features and their impacts on SAMG. (authors)

  11. Developing and assessing accident management plans for nuclear power plants

    International Nuclear Information System (INIS)

    Hanson, D.J.; Johnson, S.P.; Blackman, H.S.; Stewart, M.A.

    1992-07-01

    This document is the second of a two-volume NUREG/CR that discusses development of accident management plans for nuclear power plants. The first volume (a) describes a four-phase approach for developing criteria that could be used for assessing the adequacy of accident management plans, (b) identifies the general attributes of accident management plans (Phase 1), (c) presents a prototype process for developing and implementing severe accident management plans (Phase 2), and (d) presents criteria that can be used to assess the adequacy of accident management plans. This volume (a) describes results from an evaluation of the capabilities of the prototype process to produce an accident management plan (Phase 3) and (b), based on these results and preliminary criteria included in NUREG/CR-5543, presents modifications to the criteria where appropriate

  12. Plant specific severe accident management - the implementation phase

    International Nuclear Information System (INIS)

    Prior, R.

    1999-01-01

    Many plants are in the process of developing on-site guidance for technical staff to respond to a severe accident situation severe accident management guidance (SAMG). Once the guidance is developed, the SAMG must be implemented at the plant site, and this involves addressing a number of additional aspects. In this paper, approaches to this implementation phase are reviewed, including review and verification of plant specific SAMG, organizational aspects and integration with the emergency plan, training of SAMG users, validation and self-assessment and SAMG maintenance. Examples draw on experience from assisting numerous plants to implement symptom based severe accident management guidelines based on the Westinghouse Owners Group approach, in Westinghouse, non-Westinghouse and VVER plant types. It is hoped that it will be of use to those plant operators about to perform these activities.(author)

  13. A structured approach to individual plant evaluation and accident management

    International Nuclear Information System (INIS)

    Klopp, G.T.

    1991-01-01

    The current requirements for the performance of individual plant evaluations (IPE's) include the derivation of accident management insights as and if they occur in the course of finalizing an IPE. The development of formal, structured accident management programs is, however, explicitly excluded from current IPE requirements. The Nuclear Regulatory Commission is following the Nuclear Management and Resources Council (NUMARC) efforts to establish the framework(s) for accident management program development and plants to issue requirements on such development at a later date. The Commonwealth Edison program consists of comprehensive level 2 PRA's which address the requirements for IPE's and which go beyond those requirements. From the start of the IPE efforts, it was firmly held, within Edison, that the best way to fully and economically extract a viable accident management program from an IPE was to integrate the two efforts from the start and include the accident management program development as a required IPE product

  14. Lessons Learned after Nuclear Power Plants and Hydropower Plants Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Moskalenko, A., E-mail: gce@gce.ru [GCE Group, Saint Petersburg (Russian Federation)

    2014-10-15

    Full text: The World is becoming more open and free for communication. However, the experience (positive or negative) is still badly cross over sectorial borders. I would like to illustrate the point with the examples, even with several unexpected ones. I would like to start with a few words regarding the Sayano – Shushenskaya Hydro Power Plant accident and the factors that caused it. Sayano – Shushenskaya Hydro Power Plant is a unique Hydro Power Plant with efficiency factor of 96 %. Nevertheless, the efficiency factor, in particular, caused a series of restrictions: hydro-electric units vibration amplitude must not exceed 4 micron!!! (Slide 1: Vibration amplitude dependence on output capacity) As it is clearly seen, there is a so called “prohibited area”, which the hydro-electric unit must pass over. Operations in the area are prohibited in accordance with the regulatory documents. However, due to the changes that occurred in Russian power supply industry, the hydro-electric unit passed through the prohibited area more than 12 times, if we take into account only the day of the accident. The bolts keeping the turbine cover, keeping water apart from the machinery hall, were too much released. The mentioned above reasons led to the hydro-electric unit disruption and the machinery hall flooding. Water inflow was possible to stop by putting down the regulating valves. However, the regulating valves control console was in the flooded machinery hall. There was standby emergency control console, but it was in the machinery hall, as well. The machinery hall was flooded, consequently, main and standby systems were destroyed. Moreover, the machinery hall, where all the units were disposed, was a huge hall without dividing walls, etc. (Photo) Take a look at the next slide. (Photo – Chernobyl Nuclear Power Plant machinery hall). Take note of Fukushima–1 Nuclear Power Plant: standby power supply source was situated in the same place and destroyed by water. All the

  15. Severe accident management at South Africa's Koeberg plant

    International Nuclear Information System (INIS)

    Prior, R.P.; Wolvaardt, F.P.; Holderbaum, D.F.; Lutz, R.J.; Taylor, J.J.; Hodgson, C.D.

    1997-01-01

    Between the middle of 1993 and the end of 1995, Westinghouse and Eskom implemented plant specific Severe Accident Management Guidelines (SAMGs) at the Koeberg Nuclear Power Plant in South Africa. Prior to this project, Koeberg, like many plants, had emergency operating procedures which contain guidance for plant personnel to perform preventive accident management measures in event of an accident. There was, however, no structured guidance on recovery from an event which progresses past core damage -mitigative accident management. The SAMGs meet this need. In this paper, the Westinghouse approach to severe accident management is outlined, and the Koeberg implementation project described. A few key issues which arose during implementation are discussed, including plant instrumentation, flooding of the reactor pit, organisation and training of the Technical Support Centre staff, and impact of SAMG on risk. The means by which both generic and plant-specific SAMG have been validated is also summarised. In the next few years, many LWR owners will be implementing SAMG. In the U.S. all plants are in the process of developing SAMG. The Koeberg project is believed to be the first plant specific implementation of the WOG SAMG worldwide, and this paper has hopefully provided insights into some of the implementation issues for those about to undertake similar projects. (author)

  16. Accident analysis in nuclear power plants

    International Nuclear Information System (INIS)

    Silva, D.E. da

    1981-01-01

    The way the philosophy of Safety in Depth can be verified through the analysis of simulated accidents is shown. This can be achieved by verifying that the integrity of the protection barriers against the release of radioactivity to the environment is preserved even during accident conditions. The simulation of LOCA is focalized as an example, including a study about the associated environmental radiological consequences. (Author) [pt

  17. Locations of criticality alarms and nuclear accident dosimeters at Hanford

    International Nuclear Information System (INIS)

    1992-08-01

    Hanford facilities that contain fissionable materials capable of achieving critical mass are monitored with nuclear accident dosimeters (NADS) in compliance with the requirements of DOE Order 5480.11, Chapter XI, Section 4.c. (DOE 1988). The US Department of Energy (DOE) Richland Field Office (RL) has assigned the responsibility for maintaining and evaluating the Hanford NAD system to the Instrumentation and External Dosimetry (I ampersand ED) Section of Pacific Northwest Laboratory's (PNL's) Health Physics Department. This manual provides a description of the Hanford NAD, criteria and instructions for proper NAD placement, and the locations of these dosimeters onsite

  18. Consequences of potential accidents in heavy water plants

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  19. Effects of the Chernobyl accident on public perceptions of nuclear plant accident risks

    International Nuclear Information System (INIS)

    Lindell, M.K.; Perry, R.W.

    1990-01-01

    Assessments of public perceptions of the characteristics of a nuclear power plant accident and affective responses to its likelihood were conducted 5 months before and 1 month after the Chernobyl accident. Analyses of data from 69 residents of southwestern Washington showed significant test-retest correlations for only 10 of 18 variables--accident likelihood, three measures of impact characteristics, three measures of affective reactions, and hazard knowledge by governmental sources. Of these variables, only two had significant changes in mean ratings; frequency of thought and frequency of discussion about a nearby nuclear power plant both increased. While there were significant changes only for two personal consequences (expectations of cancer and genetic effects), both of these decreased. The results of this study indicate that more attention should be given to assessing the stability of risk perceptions over time. Moreover, the data demonstrate that experience with a major accident can actually decrease rather than increase perceptions of threat

  20. Accident analysis device for nuclear power plants

    International Nuclear Information System (INIS)

    Ito, Masayuki.

    1982-01-01

    Purpose: To enable rapid recognition of and countermeasure required for accidents upon scram, by identifying the first contact point of causes for resulting the scram and displaying the contact point of causes. Constitution: When a scram signal is inputted by way of process input device, the time of the input is determined by a timer and the contact point of causes generated just before is taken as the point whose changes occurred prior to but most closely to the generation of the signal while referring to the data memory section for the time of change of the contact point of the cause, and sent to the accident analyzing display. The accident analyzing display extracts, based on the contact point of cause, a list for the forecast accidents corresponding thereto from the data memory section and also extracts the list for the corresponding confirmation items of the accident detection and displays them together with the system from which the scram signal has been generated, the time of generation, the name of the contact point of causes operated at first, and the value of the state quantity contained in the data memory section for the store of contact point of cause at the change. (Kawakami, Y.)

  1. An assessment of criticality safety at the Department of Energy Rocky Flats Plant, Golden, Colorado, July--September 1989

    Energy Technology Data Exchange (ETDEWEB)

    Mattson, Roger J.

    1989-09-01

    This is a report on the 1989 independent Criticality Safety Assessment of the Rocky Flats Plant, primarily in response to public concerns that nuclear criticality accidents involving plutonium may have occurred at this nuclear weapon component fabrication and processing plant. The report evaluates environmental issues, fissile material storage practices, ventilation system problem areas, and criticality safety practices. While no evidence of a criticality accident was found, several recommendations are made for criticality safety improvements. 9 tabs.

  2. A Review of Accident Modelling Approaches for Complex Critical Sociotechnical Systems

    National Research Council Canada - National Science Library

    Qureshi, Zahid H

    2008-01-01

    .... This report provides a review of key traditional accident modelling approaches and their limitations, and describes new system-theoretic approaches to the modelling and analysis of accidents in safety-critical systems...

  3. Consequence of potential accidents in heavy water plants

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  5. Environmental radiation exposure in case of power plant accidents

    International Nuclear Information System (INIS)

    Eder, K.

    1977-01-01

    The paper tries to overcome prejudices concerning radiation effects due to power plant accidents as well as to show the radiation exposure that may be expected near the the patient and to indicate ways and means to avoid or reduce this radiation exposure and to avoid contamination. It is a contribution to better information on radiation accidents and radiolesions in nuclear power plants with the aim of close cooperation between power plants, physicians, and hospitals and of helping to overcome erroneous popular assumptions. (orig./HP) [de

  6. Accident sequences simulated at the Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1998-01-01

    Different hypothetical accident sequences have been simulated at Unit 1 of the Juragua nuclear power plant in Cuba, a plant with two VVER-440 V213 units under construction. The computer code MELCOR was employed for these simulations. The sequences simulated are: (1) a design-basis accident (DBA) large loss of coolant accident (LOCA) with the emergency core coolant system (ECCS) on, (2) a station blackout (SBO), (3) a small LOCA (SLOCA) concurrent with SBO, (4) a large LOCA (LLOCA) concurrent with SBO, and (5) a LLOCA concurrent with SBO and with the containment breached at time zero. Timings of important events and source term releases have been calculated for the different sequences analyzed. Under certain weather conditions, the fission products released from the severe accident sequences may travel to southern Florida

  7. Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

    International Nuclear Information System (INIS)

    Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed

    2013-01-01

    The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed

  8. Solutions to criticality problems in a plutonium extraction plant

    International Nuclear Information System (INIS)

    Jouannaud, C.; Rodier, J.; Fruchard, Y.; Peyresblanques, H.; Papault, C.; Tabardel-Brian, R.

    1968-08-01

    There are two aspects to nuclear criticality safety: prevention of criticality and protection against the consequences of a possible accident: this report considers these two aspects in the case of the Marcoule Plutonium Extraction Plant. After briefly recalling the various techniques used for avoiding criticality (mass, geometry, concentration, poisoning), the authors describe their application in the plant and show in particular that, a rational use of a favorable geometry is a factor both for security and from an economic point of view. The authors then describe the inside organisation which makes it possible to obtain the necessary intrinsic safety standard right from the advance project stage, and to control the workshop safety during the operation of the plant. The second part of the report deals with the system of protection against the consequences of a possible accident: definition of a typical accident, fixing of the boundaries of a critical zone, safety alarm device, individual and collective dosimetry, evacuation plan and safety instructions. (authors) [fr

  9. Applying Functional Modeling for Accident Management of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Morten; Zhang Xinxin [Harbin Engineering University, Harbin (China)

    2014-08-15

    The paper investigate applications of functional modeling for accident management in complex industrial plant with special reference to nuclear power production. Main applications for information sharing among decision makers and decision support are identified. An overview of Multilevel Flow Modeling is given and a detailed presentation of the foundational means-end concepts is presented and the conditions for proper use in modelling accidents are identified. It is shown that Multilevel Flow Modeling can be used for modelling and reasoning about design basis accidents. Its possible role for information sharing and decision support in accidents beyond design basis is also indicated. A modelling example demonstrating the application of Multilevel Flow Modelling and reasoning for a PWR LOCA is presented.

  10. Hazards to nuclear plants from surface traffic accidents

    International Nuclear Information System (INIS)

    Hornyik, K.

    1975-01-01

    Analytic models have been developed for evaluating hazards to nuclear plants from hazardous-materials accidents in the vicinity of the plant. In particular, these models permit the evaluation of hazards from such accidents occurring on surface traffic routes near the plant. The analysis uses statistical information on accident rates, traffic frequency, and cargo-size distribution along with parameters describing properties of the hazardous cargo, plant design, and atmospheric conditions, to arrive at a conservative estimate of the annual probability of a catastrophic event. Two of the major effects associated with hazardous-materials accidents, explosion and release of toxic vapors, are treated by a common formalism which can be readily applied to any given case by means of a graphic procedure. As an example, for a typical case it is found that railroad shipments of chlorine in 55-ton tank cars constitute a greater hazard to a nearby nuclear plant than equally frequent rail shipments of explosives in amounts of 10 tons. 11 references. (U.S.)

  11. Update of the Nuclear Criticality Slide Rule for the Emergency Response to a Nuclear Criticality Accident

    Science.gov (United States)

    Duluc, Matthieu; Bardelay, Aurélie; Celik, Cihangir; Heinrichs, Dave; Hopper, Calvin; Jones, Richard; Kim, Soon; Miller, Thomas; Troisne, Marc; Wilson, Chris

    2017-09-01

    AWE (UK), IRSN (France), LLNL (USA) and ORNL (USA) began a long term collaboration effort in 2015 to update the nuclear criticality Slide Rule for the emergency response to a nuclear criticality accident. This document, published almost 20 years ago, gives order of magnitude estimates of key parameters, such as number of fissions and doses (neutron and gamma), useful for emergency response teams and public authorities. This paper will present, firstly the motivation and the long term objectives for this update, then the overview of the initial configurations for updated calculations and preliminary results obtained with modern 3D codes.

  12. Update of the Nuclear Criticality Slide Rule for the Emergency Response to a Nuclear Criticality Accident

    Directory of Open Access Journals (Sweden)

    Duluc Matthieu

    2017-01-01

    Full Text Available AWE (UK, IRSN (France, LLNL (USA and ORNL (USA began a long term collaboration effort in 2015 to update the nuclear criticality Slide Rule for the emergency response to a nuclear criticality accident. This document, published almost 20 years ago, gives order of magnitude estimates of key parameters, such as number of fissions and doses (neutron and gamma, useful for emergency response teams and public authorities. This paper will present, firstly the motivation and the long term objectives for this update, then the overview of the initial configurations for updated calculations and preliminary results obtained with modern 3D codes.

  13. Development of Northeast Asia Nuclear Power Plant Accident Simulator.

    Science.gov (United States)

    Kim, Juyub; Kim, Juyoul; Po, Li-Chi Cliff

    2017-06-15

    A conclusion from the lessons learned after the March 2011 Fukushima Daiichi accident was that Korea needs a tool to estimate consequences from a major accident that could occur at a nuclear power plant located in a neighboring country. This paper describes a suite of computer-based codes to be used by Korea's nuclear emergency response staff for training and potentially operational support in Korea's national emergency preparedness and response program. The systems of codes, Northeast Asia Nuclear Accident Simulator (NANAS), consist of three modules: source-term estimation, atmospheric dispersion prediction and dose assessment. To quickly assess potential doses to the public in Korea, NANAS includes specific reactor data from the nuclear power plants in China, Japan and Taiwan. The completed simulator is demonstrated using data for a hypothetical release. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  14. Concerning the structure of occupational accidents involving construction workers in the erection of nuclear power plants

    International Nuclear Information System (INIS)

    Nowak, B.; Roebenack, K.D.

    1991-01-01

    An investigation of 561 occupational accidents involving construction workers which took place during the construction of nuclear power plants failed to show any significant deviation in comparison with general construction as regards process classification, classification of accidents according to occupation and situation, and accidents severity. Occupational accidents which are typial for nuclear power plant construction are a rare exception. (orig.) [de

  15. Biological dosimetry following exposure to neutrons in a criticality accident

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, C. (Radiation and Nuclear Safety Authority, STUK (Finland)); Wojcik, A. (Stockholm Univ. (SU), Stockholm (Sweden)); Jaworska, A. (Norwegian Radiation Protection Authority (NRPA) (Norway))

    2011-01-15

    The aim of the BIONCA project was to implement cytogenetic techniques for biodosimetry purposes in the Nordic countries. The previous NKS-funded biodosimetry activities (BIODOS and BIOPEX) concentrated on experiments using gamma-irradiation and on developing the PCC ring assay for biodosimetry. Experiments conducted during the present BIONCA project has broadened the biodosimetry capacity of the Nordic countries to include dose estimation of exposure to neutrons for both PCC ring and dicentric chromosome techniques. In 2009, experiments were conducted for establishing both PCC ring and dicentric dose calibration curves. Neutron irradiation of human whole blood obtained from two volunteers was conducted in the Netherlands at the Petten reactor. Cell cultures and analysis of whole blood exposed to eight doses between 0 and 10 Gy were performed for both techniques. For the dicentric assay, excellent uniformity in dose calibration for data from both SU and STUK was observed. For PCC rings, the SU and STUK curves were not equally congruent, probably due to the less uniform scoring criteria. However, both curves displayed strong linearity throughout the dose range. In 2010, an exercise was conducted to simulate a criticality accident and to test the validity of the established dose calibration curves. For accident simulation, 16 blood samples were irradiated in Norway at the Kjeller reactor and analysed for dose estimation with both assays. The results showed that, despite a different com-position of the radiation beams in Petten and Kjeller, good dose estimates were obtained. The activity has provided good experience on collaboration required in radiation emergency situations where the biodosimetry capacity and resources of one laboratory may be inadequate. In this respect, the project has strengthened the informal network between the Nordic countries: STUK, the Finnish Radiation and Nuclear Safety Authority, NRPA, the Norwegian Radiation Protection Authority and SU

  16. The accident at the Three Mile Island nuclear power plant

    International Nuclear Information System (INIS)

    Butragueno, J.L.

    1980-01-01

    The sequence of events in the Three Mile Island, Unit 2, accident on the March 28, 1979 is analyzed. In this plant a loss of feed-water transient became a small LOCA that caused a serious core damage. A general emergency situation was declared after uncontrolled radioactive releases were detectec. (author)

  17. Children's reactions to the threat of nuclear plant accidents

    International Nuclear Information System (INIS)

    Schwebel, M.; Schwebel, B.

    1981-01-01

    In the wake of Three Mile Island nuclear plant accident, questionnaire and interview responses of children in elementary and secondary schools revealed their perceptions of the dangers entailed in the continued use of nuclear reactors. Results are compared with a parallel study conducted close to 20 years ago, and implications for mental health are examined

  18. A criticism of ANSI/ANS-8.3-1986: Criticality accident alarm system

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1991-01-01

    The American National Standard on criticality accident alarm systems has given rise to confusion in interpretation and implementation of the requirements. In addition, some of the standards have recently been incorporated into US Department of Energy (DOE) orders, and others have been paraphrased in the DOE orders. Some of the DOE orders referencing these standards are being incorporated into law by means of the Code of Federal Regulations. As such, the intent of the authors of the standards to recommend a code of good practice is now being codified into law with attendant civil and criminal penalties for failure to comply. It is suggested that ANSI/ANS-8.3-1986, Critically Accident Alarm System, be carefully reviewed to alleviate the confusion that has been experienced in practice, to clarify the minimum accident of concern, to further define the dose (or dose rate) criteria for activation, and to stress the fact that a prime consideration in any safety system is the overall reduction of risk

  19. The medical implications of nuclear power plant accidents

    International Nuclear Information System (INIS)

    Tyror, J.G.; Pearson, G.W.

    1989-11-01

    This paper examines the UK position regarding the potential for an accident at a nuclear power plant, the safeguards in place to prevent such an accident occurring and the emergency procedures designed to cope with the consequences should one occur. It focuses on the role of the medical services and examines previous accidents to suggest the nature and likely scale of response that may need to be provided. It is apparent that designs of UK nuclear power stations are robust and that the likelihood of a significant accident occurring is extremely remote. Emergency arrangements are, however, in place to deal with the eventuality should it arise and these incorporate sufficient flexibility to accommodate a wide range of accidents. Analysis of previous nuclear accidents at Windscale, Three Mile Island and Chernobyl provide a limited but valuable insight into the diversity and potential scale of response that may be required. It is concluded that above all, the response must be flexible to enable medical services to deal with the wide range of effects that may arise. (author)

  20. Discussion on several issues of the accidents management of nuclear power plants in operation

    International Nuclear Information System (INIS)

    Cao Xuewu; Wang Zhe; Zhang Yingzhen

    2009-01-01

    This article discusses several issues of the accident management of nuclear power plants in operation, for example: the necessity, implementation principle of accident management and accident management program etc. For conducting accident management for beyond design basis accidents, this article thinks that the accident management program should be developed and implemented to ensure that the plant and its personnel with responsibilities for accident management are adequately prepared to take effective on-site actions to prevent or mitigate the consequences of severe accident. (authors)

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

    International Nuclear Information System (INIS)

    Barbry, F.

    1984-01-01

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

  2. Mitigation of severe accidents in Swedish nuclear power plants

    International Nuclear Information System (INIS)

    Soederman, E.

    1987-01-01

    Sweden is the first country to build filtered venting systems, the first one became operable at Barsebaeck nuclear power plant in 1985. In new concepts, now being installed in Sweden, an enhanced containment spray system is the basic element and the filtered venting is only the secondary mitigating system. The filter is a new design, a submerged multi venturi scrubber. The Swedish strategy has been built on three basics: improved knowledge through research; containment integrity through mitigating systems; and accident management to prevent severe accidents. 2 figs

  3. Integrated color face graphs for plant accident display

    International Nuclear Information System (INIS)

    Hara, Fumio

    1987-01-01

    This paper presents an integrated man-machine interface that uses cartoon-like colored graphs in the form of faces, that, through different facial expressions, display a plant condition. This is done by drawing the face on a CRT by nonlinearly transforming 31 variables and coloring the face. This integrated color graphics technique is applied to display the progess of events in the Three Mile Island nuclear power plant accident. Human visual perceptive characteristics are investigated in relation to the perception of the plant accident process, the naturality in face color change, and the consistency between facial expressions and colors. This paper concludes that colors used in an integrated color face graphs must be completely consistent with emotional feelings perceived from the colors. (author)

  4. Nuclear power plant safety - the risk of accidents

    International Nuclear Information System (INIS)

    Higson, D.; Crancher, D.W.

    1975-08-01

    Although it is physically impossible for any nuclear plant to explode like an atom bomb, an accidental release of radioactive material into the environment is conceivable. Three factors reduce the probability of such releases, in dangerous quantities, to an extremely low level. Firstly, there are many safety features built into the plant including a leaktight containment building to prevent the escape of such material. Secondly, the quality of engineering and standards used are far more demanding than in conventional power engineering. Thirdly, strict government licensing and regulatory control is enforced at all phases from design through construction to operation. No member of the general public is known to have been injured or died as a result of any accident to a commercial nuclear power plant. Ten workers have died as a result of over-exposure to radiation from experimental reactors and laboratory work connected with the development of nuclear plant since 1945. Because of this excellent safety record the risk of serious accidents can only be estimated. On the basis of such estimates, the chance of an accident in a nuclear power reactor which could cause a detectable increase in the incidence of radiation-induced illnesses would be less than one chance in a million per year. In a typical highly industrialised society, such as the USA, the estimated risk of an individual being killed by such accidents, from one hundred operating reactors, is no greater than one chance in sixteen million per year. There are undoubtedly risks from reactor accidents but estimates of these risks show that they are considerably less than from other activities which are accepted by society. (author)

  5. TL detectors for gamma ray dose measurements in criticality accidents

    International Nuclear Information System (INIS)

    Miljanic, S.; Zorko, B.; Gregori, B.; Knezevic, Z.

    2007-01-01

    Determination of gamma ray dose in mixed neutron + gamma ray fields is still a demanding task. Dosemeters used for gamma ray dosimetry are usually in some extent sensitive to neutrons and their response variations depend on neutron energy i.e., on neutron spectra. Besides, it is necessary to take into account the energy dependence of dosemeter responses to gamma rays. In this work, several types of thermoluminescent detectors (TLD) placed in different holders used for gamma ray dose determination in the mixed fields were examined. Dosemeters were from three different institutions: Ruder Boskovic Inst. (RBI), Croatia, Jozef Stefan Inst. (JSI), Slovenia and Autoridad Regulatoria Nuclear (ARN), Argentina. All dosemeters were irradiated during the International Intercomparison of Criticality Accident Dosimetry Systems at the SILENE Reactor, Valduc, June 2002. Three accidental scenarios were reproduced and in each irradiation the dosemeters were exposed placed on the front of phantom and 'free in air'. Following types of TLDs were used: 7 LiF (TLD-700), CaF 2 :Mn and Al2 O3 :Mg,Y - all from RBI; CaF 2 :Mn from JSI and 7 LiF (TLD-700) from ARN. Reported doses were compared with the reference values as well as with the values obtained from the results of all participants. The results show satisfactory agreement with other dosimetry systems used in the Intercomparison. The influence of different types of holders and applied corrections of dosemeters' readings are discussed. (authors)

  6. TL detectors for gamma ray dose measurements in criticality accidents.

    Science.gov (United States)

    Miljanić, Saveta; Zorko, Benjamin; Gregori, Beatriz; Knezević, Zeljka

    2007-01-01

    Determination of gamma ray dose in mixed neutron+gamma ray fields is still a demanding task. Dosemeters used for gamma ray dosimetry are usually in some extent sensitive to neutrons and their response variations depend on neutron energy i.e., on neutron spectra. Besides, it is necessary to take into account the energy dependence of dosemeter responses to gamma rays. In this work, several types of thermoluminescent detectors (TLD) placed in different holders used for gamma ray dose determination in the mixed fields were examined. Dosemeters were from three different institutions: Ruder Bosković Institute (RBI), Croatia, JoZef Stefan Institute (JSI), Slovenia and Autoridad Regulatoria Nuclear (ARN), Argentina. All dosemeters were irradiated during the International Intercomparison of Criticality Accident Dosimetry Systems at the SILENE Reactor, Valduc, June 2002. Three accidental scenarios were reproduced and in each irradiation the dosemeters were exposed placed on the front of phantom and 'free in air'. Following types of TLDs were used: 7LiF (TLD-700), CaF2:Mn and Al2O3:Mg,Y-all from RBI; CaF2:Mn from JSI and 7LiF (TLD-700) from ARN. Reported doses were compared with the reference values as well as with the values obtained from the results of all participants. The results show satisfactory agreement with other dosimetry systems used in the Intercomparison. The influence of different types of holders and applied corrections of dosemeters' readings are discussed.

  7. Research on the management of the wastes from plant accidents

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The accident in Fukushima Daiichi Nuclear Power Plant released large amount of radio-nuclides and contaminated wide areas within and out of the site. The decontamination, storage, treatment and disposal of generated wastes are now under planning. Though the regulations for radioactive wastes discharged from normal operation and decommissioning of nuclear facilities have been prepared, it is necessary to make amendments of those regulations to deal with wastes from the severe accidents which may have much different features on nuclides contents, or possibility to accompany hazardous chemical materials. Characteristics, treatment and disposal of wastes from accidents were surveyed by literature and the radionuclide migration from the assumed temporally storage yards of the disaster debris was analyzed for consideration of future regulation. (author)

  8. Investigation of the management of the wastes from plant accident

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-08-15

    The accident in Fukushima Daiichi Nuclear Power Plant discharged large amount of radio-nuclides and contaminated wide areas in and out of the site. The decontamination, storage, treatment and disposal of generated wastes are now under planning. Though regulations for the radioactive wastes arisen from normal operation and decommissioning of nuclear facilities have been prepared, it is necessary to make amendment of those regulations to deal with wastes from the severe accident which may have much different features on nuclides contents, or possible accompanying hazardous chemical materials. Characteristics of wastes from accidents in foreign nuclear installations, and the treatment and the disposal of those wastes were surveyed by literature and radionuclide migration from the assumed temporally storage yards of the disaster debris was analyzed for consideration of future regulation. (author)

  9. Chemical Plant Accidents in a Nuclear Hydrogen Generation Scheme

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Revankar, Shripad T.

    2011-01-01

    A high temperature nuclear reactor (HTR) could be used to drive a steam reformation plant, a coal gasification facility, an electrolysis plant, or a thermochemical hydrogen production cycle. Most thermochemical cycles are purely thermodynamic, and thus achieve high thermodynamic efficiency. HTRs produce large amounts of heat at high temperature (1100 K). Helium-cooled HTRs have many passive, or inherent, safety characteristics. This inherent safety is due to the high design basis limit of the maximum fuel temperature. Due to the severity of a potential release, containment of fission products is the single most important safety issue in any nuclear reactor facility. A HTR coupled to a chemical plant presents a complex system, due primarily to the interactive nature of both plants. Since the chemical plant acts as the heat sink for the nuclear reactor, it important to understand the interaction and feedback between the two systems. Process heat plants and HTRs are generally very different. Some of the major differences include: time constants of plants, safety standards, failure probability, and transient response. While both the chemical plant and the HTR are at advanced stages of testing individually, no serious effort has been made to understand the operation of the integrated system, especially during accident events that are initiated in the chemical plant. There is a significant lack of knowledge base regarding scaling and system integration for large scale process heat plants coupled to HTRs. Consideration of feedback between the two plants during time-dependent scenarios is absent from literature. Additionally, no conceptual studies of the accidents that could occur in either plant and impact the entire coupled system are present in literature

  10. Analysis and evaluation of the nuclear criticality accident in JCO CO. LTD in Japan

    International Nuclear Information System (INIS)

    Liu Hua; Liu Xinhua; Li Bing

    2001-01-01

    The author describes JCO criticality accident situation including the background, process chronology and emergency countermeasures taken of the accident and its radiation consequence. The analysis about the direct and root causes of the accident and some conclusions are also showed. The direct cause of the accident is the use of geometrically unsafe process equipment and personnel violation. However, the root cause is lack of efficient technical management. Therefore, it is necessary to emphasize the criticality safety in nuclear fuel cycle installations and enhance safety culture of regulatory and operational personnel

  11. Millstone Unit 1 plant vulnerabilities during postulated severe nuclear accidents

    International Nuclear Information System (INIS)

    Khalil, Y.F.

    1993-01-01

    Generic Letter 88-20, Supplement No. 1 (Ref. 1), issued by the Nuclear Regulatory Commission (NRC) requested all licensees holding operating licenses and construction permits for nuclear power reactor facilities to perform Individual Plant Examinations (IPE) of their plant(s) for severe accident vulnerabilities and to submit the results to the Commission. This paper summarizes the major Front-End (Level-1 PRA) and Back-End (Level-2 PRA) insights gained from the Millstone Unit 1 (MP-1) IPE study. No major plant vulnerabilities have been identified from a Front-End perspective. The Back-End analysis, however, has identified two potential containment vulnerabilities during postulated events that progress beyond the Design Basis Accidents (DBAs), namely, (1) MP-1 is dominated by early source term releases that would occur within a six-hour time frame from time of accident initiation, or reactor trip, and (2) MP-1 containment is somewhat vulnerable to leak-type failure through the drywell head. As a result of the second finding, a recommendation currently under evaluation, has been made to increase the drywell head bolt's preload from 54 Kips to resist the containment design pressure value (62 psig)

  12. Radiation Exposure and Thyroid Cancer Risk After the Fukushima Nuclear Power Plant Accident in Comparison with the Chernobyl Accident.

    Science.gov (United States)

    Yamashita, S; Takamura, N; Ohtsuru, A; Suzuki, S

    2016-09-01

    The actual implementation of the epidemiological study on human health risk from low dose and low-dose rate radiation exposure and the comprehensive long-term radiation health effects survey are important especially after radiological and nuclear accidents because of public fear and concern about the long-term health effects of low-dose radiation exposure have increased considerably. Since the Great East Japan earthquake and the Fukushima Daiichi Nuclear Power Plant accident in Japan, Fukushima Prefecture has started the Fukushima Health Management Survey Project for the purpose of long-term health care administration and medical early diagnosis/treatment for the prefectural residents. Especially on a basis of the lessons learned from the Chernobyl accident, both thyroid examination and mental health care are critically important irrespective of the level of radiation exposure. There are considerable differences between Chernobyl and Fukushima regarding radiation dose to the public, and it is very difficult to estimate retrospectively internal exposure dose from the short-lived radioactive iodines. Therefore, the necessity of thyroid ultrasound examination in Fukushima and the intermediate results of this survey targeting children will be reviewed and discussed in order to avoid any misunderstanding or misinterpretation of the high detection rate of childhood thyroid cancer. © World Health Organisation 2016. All rights reserved. The World Health Organization has granted Oxford University Press permission for the reproduction of this article.

  13. Radiation Exposure and Thyroid Cancer Risk After the Fukushima Nuclear Power Plant Accident in Comparison with the Chernobyl Accident

    International Nuclear Information System (INIS)

    Yamashita, S.; Takamura, N.; Ohtsuru, A.; Suzuki, S.

    2016-01-01

    The actual implementation of the epidemiological study on human health risk from low dose and low-dose rate radiation exposure and the comprehensive long-term radiation health effects survey are important especially after radiological and nuclear accidents because of public fear and concern about the long-term health effects of low-dose radiation exposure have increased considerably. Since the Great East Japan earthquake and the Fukushima Daiichi Nuclear Power Plant accident in Japan, Fukushima Prefecture has started the Fukushima Health Management Survey Project for the purpose of long-term health care administration and medical early diagnosis/treatment for the prefectural residents. Especially on a basis of the lessons learned from the Chernobyl accident, both thyroid examination and mental health care are critically important irrespective of the level of radiation exposure. There are considerable differences between Chernobyl and Fukushima regarding radiation dose to the public, and it is very difficult to estimate retrospectively internal exposure dose from the short-lived radioactive iodines. Therefore, the necessity of thyroid ultrasound examination in Fukushima and the intermediate results of this survey targeting children will be reviewed and discussed in order to avoid any misunderstanding or misinterpretation of the high detection rate of childhood thyroid cancer. (authors)

  14. Criticality safety evaluation in Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    Shirai, Nobutoshi; Nakajima, Masayoshi; Takaya, Akikazu; Ohnuma, Hideyuki; Shirouzu, Hidetomo; Hayashi, Shinichiro; Yoshikawa, Koji; Suto, Toshiyuki

    2000-04-01

    Criticality limits for equipments in Tokai Reprocessing Plant which handle fissile material solution and are under shape and dimension control were reevaluated based on the guideline No.10 'Criticality safety of single unit' in the regulatory guide for reprocessing plant safety. This report presents criticality safety evaluation of each equipment as single unit. Criticality safety of multiple units in a cell or a room was also evaluated. The evaluated equipments were ones in dissolution, separation, purification, denitration, Pu product storage, and Pu conversion processes. As a result, it was reconfirmed that the equipments were safe enough from a view point of criticality safety of single unit and multiple units. (author)

  15. Proposed chemical plant initiated accident scenarios in a sulphur-iodine cycle plant coupled to a pebble bed modular reactor

    International Nuclear Information System (INIS)

    Brown, N.R.; Revankar, S.T.; Seker, V.; Downar, Th.J.

    2010-01-01

    In the sulphur-iodine (S-I) cycle nuclear hydrogen generation scheme the chemical plant acts as the heat sink for the very high temperature nuclear reactor (VHTR). Thus, any accident which occurs in the chemical plant must feedback to the nuclear reactor. There are many different types of accidents which can occur in a chemical plant. These accidents include intra-reactor piping failure, inter-reactor piping failure, reaction chamber failure and heat exchanger failure. Since the chemical plant acts as the heat sink for the nuclear reactor, any of these accidents induce a loss-of-heat-sink accident in the nuclear reactor. In this paper, several chemical plant initiated accident scenarios are presented. The following accident scenarios are proposed: i) failure of the Bunsen chemical reactor; ii) product flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iii) reactant flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iv) rupture of a reaction chamber. Qualitative analysis of these accident scenarios indicates that each result in either partial or total loss of heat sink accidents for the nuclear reactor. These scenarios are reduced to two types: i) discharge rate limited accidents; ii) discontinuous reaction chamber accidents. A discharge rate limited rupture of the SO 3 decomposition section of the SI cycle is proposed and modelled. Since SO 3 decomposition occurs in the gaseous phase, critical flow out of the rupture is calculated assuming ideal gas behaviour. The accident scenario is modelled using a fully transient control volume model of the S-I cycle coupled to a THERMIX model of a 268 MW pebble bed modular reactor (PBMR-268) and a point kinetics model. The Bird, Stewart and Lightfoot source model for choked gas flows from a pressurised chamber was utilised as a discharge rate model. A discharge coefficient of 0.62 was assumed. Feedback due to the rupture is observed in the nuclear

  16. Assessment of accident risks from german nuclear plants

    International Nuclear Information System (INIS)

    Heuser, F.W.

    1979-01-01

    The German risk study are presented. The main objectives can be summed up as follows: (a) An assessment of the societal risk due to accidents in nuclear power plants with reference to German conditions; (b) To get experience in the field of risk analysis and to provide a basis for estimation of uncertainties; (c) To provide guidance for future activities in the German Reactor Safety Research Program. Finally several conclusions reached by this study are discussed. (author)

  17. Japanese authorities inform IAEA about accident at nuclear plant

    International Nuclear Information System (INIS)

    2004-01-01

    Full text: The IAEA today received information from Japanese nuclear regulatory authorities about an accident in the steam generator turbine circuit of the Mihama Nuclear Power Plant (unit 3). According to the Japanese nuclear authorities this is a non-radioactive part of the plant. The regulatory body has reported that four contract employees died and 7 were injured, and stated that there was no release of radioactivity. The IAEA continues to be in contact with Japanese authorities and expects to receive updates on a continuous basis. No request for IAEA assistance has been received at this time. (IAEA)

  18. Bibliography for nuclear criticality accident experience, alarm systems, and emergency management

    International Nuclear Information System (INIS)

    Putman, V.L.

    1995-09-01

    The characteristics, detection, and emergency management of nuclear criticality accidents outside reactors has been an important component of criticality safety for as long as the need for this specialized safety discipline has been recognized. The general interest and importance of such topics receives special emphasis because of the potentially lethal, albeit highly localized, effects of criticality accidents and because of heightened public and regulatory concerns for any undesirable event in nuclear and radiological fields. This bibliography lists references which are potentially applicable to or interesting for criticality alarm, detection, and warning systems; criticality accident emergency management; and their associated programs. The lists are annotated to assist bibliography users in identifying applicable: industry and regulatory guidance and requirements, with historical development information and comments; criticality accident characteristics, consequences, experiences, and responses; hazard-, risk-, or safety-analysis criteria; CAS design and qualification criteria; CAS calibration, maintenance, repair, and testing criteria; experiences of CAS designers and maintainers; criticality accident emergency management (planning, preparedness, response, and recovery) requirements and guidance; criticality accident emergency management experience, plans, and techniques; methods and tools for analysis; and additional bibliographies

  19. Implementation of accident management programmes in nuclear power plants

    International Nuclear Information System (INIS)

    2004-01-01

    According to the generally established defence in depth concept in nuclear safety, consideration in plant operation is also given to highly improbable severe plant conditions that were not explicitly addressed in the original design of currently operating nuclear power plants (NPPs). Defence in depth is achieved primarily by means of four successive barriers which prevent the release of radioactive material (fuel matrix, cladding, primary coolant boundary and containment), and these barriers are primarily protected by three levels of design measures: prevention of abnormal operation and failures (level 1), control of abnormal operation and detection of failures (level 2) and control of accidents within the design basis (level 3). If these first three levels fail to ensure the structural integrity of the core, e.g. due to beyond the design basis multiple failures, or due to extremely unlikely initiating events, additional efforts are made at level 4 to further reduce the risks. The objective at the fourth level is to ensure that both the likelihood of an accident entailing significant core damage (severe accident) and the magnitude of radioactive releases following a severe accident are kept as low as reasonably achievable. Finally, level 5 includes off-site emergency response measures, with the objective of mitigating the radiological consequences of significant releases of radioactive material. The implementation of the emergency response is usually dependent upon the type and magnitude of the accident. Good co-ordination between the operator and the responding organizations is needed to ensure the appropriate response. Accident management is one of the key components of effective defence in depth. In accordance with defence in depth, each design level should be protected individually, independently of other levels. This report focuses on the fourth level of defence in depth, including the transitions from the third level and into the fifth level. It describes

  20. Operators' arrangement for handling nuclear accidents at power plants

    International Nuclear Information System (INIS)

    Bertron, L.; Meclot, B.

    1986-01-01

    Given the preventive measures adopted by Electricite de France (EDF), the probability of a nuclear accident occurring in a power plant is extremely low but cannot, even so, be considered to be zero. The operator must therefore be prepared for this possibility. Apart from dealing with the consequences of the accident, the organization he sets up must fulfil the double objective of preventing any worsening of the accident and ensuring that the social, political and economic effects remain in proportion to the seriousness of the accident. The paper describes the organization set up by EDF in co-operation with the public authorities, indicating the concepts on which it is based and the logistical resources brought into play, in particular for telecommunications. Reports on the TMI incident showed that public telecommunications services can well be saturated in the event of an emergency. EDF, relying on the combined advantages of all transmission systems which the French Postal and Telecommunications Office can place at its disposal, as well as private networks with a concession from the Government, has taken the necessary precautions to deal with this problem. The organization is also designed to respond to the requirements of the media and the population at large for correct information. These systems are naturally all tested during training exercises which ensure that the organization as a whole can cope, in terms both of manpower and equipment, with a very improbable event. (author)

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

    International Nuclear Information System (INIS)

    Fruchard, Y.; Lavie, J.M.

    1966-01-01

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

  2. Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Gilberto Espinosa-Paredes

    2012-01-01

    Full Text Available The loss-of-coolant accident (LOCA simulation in the boiling water reactor (BWR of Laguna Verde Nuclear Power Plant (LVNPP at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

  3. Automations influence on nuclear power plants: a look at three accidents and how automation played a role.

    Science.gov (United States)

    Schmitt, Kara

    2012-01-01

    Nuclear power is one of the ways that we can design an efficient sustainable future. Automation is the primary system used to assist operators in the task of monitoring and controlling nuclear power plants (NPP). Automation performs tasks such as assessing the status of the plant's operations as well as making real time life critical situational specific decisions. While the advantages and disadvantages of automation are well studied in variety of domains, accidents remind us that there is still vulnerability to unknown variables. This paper will look at the effects of automation within three NPP accidents and incidents and will consider why automation failed in preventing these accidents from occurring. It will also review the accidents at the Three Mile Island, Chernobyl, and Fukushima Daiichi NPP's in order to determine where better use of automation could have resulted in a more desirable outcome.

  4. Evaluation of dose equivalent rate distribution in JCO critical accident by radiation transport calculation

    CERN Document Server

    Sakamoto, Y

    2002-01-01

    In the prevention of nuclear disaster, there needs the information on the dose equivalent rate distribution inside and outside the site, and energy spectra. The three dimensional radiation transport calculation code is a useful tool for the site specific detailed analysis with the consideration of facility structures. It is important in the prediction of individual doses in the future countermeasure that the reliability of the evaluation methods of dose equivalent rate distribution and energy spectra by using of Monte Carlo radiation transport calculation code, and the factors which influence the dose equivalent rate distribution outside the site are confirmed. The reliability of radiation transport calculation code and the influence factors of dose equivalent rate distribution were examined through the analyses of critical accident at JCO's uranium processing plant occurred on September 30, 1999. The radiation transport calculations including the burn-up calculations were done by using of the structural info...

  5. Study of source term evaluation from fuel solution under simulated nuclear criticality accident in TRACY

    International Nuclear Information System (INIS)

    Abe, Hitoshi; Tashiro, Shinsuke; Nagai, Hitoshi; Koike, Tadao; Okagawa, Seigo; Murata, Mikio

    1999-01-01

    In a accident at the dissolver in a reprocessing plant, various fission products and radiolysis gases will be produced in the fuel solution and volatile radioactive nuclides and radiolysis gases and nitrogen oxide will be released into vent-gas spontaneously. Moreover other on-volatile nuclide will be releases as radioactive aerosol (mist) with bursting bubbles at surface of the solution. Therefore quantitative estimation of release and transport behavior of the radioactive material from solution as source term is very important. TRACY is a transient criticality experimental facility for studying the transient criticality characteristics of low enriched uranium. In this paper, experiment methods and results about the release behavior of the hydrogen, radioactive aerosol and iodine species from the fuel solutions are reported. As the results of the experiments, release patterns of H 2 , 140 Ba and 131 I could be grasped. Concentrations of H 2 in the vent-gas and 140 Ba in the gas phase in the core tank attained to the peak just after the transient criticality and decreased exponentially with time. On the other hand, concentrations of 131 I in the gas phase of the tank began to increase with a time lag of several minutes from the transient criticality and attained approximately constant values. (J.P.N.)

  6. Lessons of the accident at Three Mile Island nuclear power plant

    International Nuclear Information System (INIS)

    Veksler, L.M.

    1983-01-01

    Measures taken in the USA for improving safety of NPPs after the accident at ''Three Mile Island'' nuclear power plant are considered. Activities, related to elimination of accident consequences are analyzed. Perspectives of resuming the NPP operation are discussed

  7. Nuclear criticality safety: general. 3. Tokaimura Criticality Accident: Point Model Stochastic Neutronic Interpretation

    International Nuclear Information System (INIS)

    Mechitoua, Boukhmes

    2001-01-01

    This paper shows what can be the stochastic neutronic contribution for the interpretation of criticality accidents. Stochastic neutronic comprehensive texts may be found in refs.1 through 4. We limit our study to the use of initiation probability, which is an important stochastic neutronic tally. Initiation probability P may be defined as the probability for one neutron to initiate an infinite neutron fission chain. The complement probability of P is the extinction probability Q. The probability that the neutron fission chain produced by one neutron will quench is equal to the multiplication of the probability of production of i neutrons g i by the probability of extinction of these i neutrons. We can estimate P by a Newton or by a dichotomic method. We suppose that P S (t) is the probability that an infinite neutron fission chain has been initiated before time t by a neutron produced by the source S(t). P S (t + dt) is the sum of two probabilities: 1. the probability that an infinite neutron fission chain has been initiated before time t by a neutron produced by the source S(t): P S (t); 2. The second probability is a multiplication of two probabilities: the probability that there was no initiation before t that is 1-P S (t), and the probability that a neutron emitted by the source with the probability S dt initiates an infinite neutron fission chain with the probability P(t). This last relation gives the link between P and the source density. The aim of this paper is to show how one can apply the foregoing derivations. We have simplified the Tokaimura criticality accident for this application. We have mono-energetic neutrons with infinite and homogeneous media; we have two reactions: capture and fission. In this section, we show how one can estimate the initiation probability with a source density as a function of time. This estimation makes use of three steps: 1. Reactivity insertion: Estimation of the multiplication coefficient as a function of time K(t). This

  8. Consideration of severe accident issues for the general electric BWR standard plant a status report

    International Nuclear Information System (INIS)

    Holtzclaw, K.W.

    1983-01-01

    In early 1982 the U.S. NRC proposed a policy to address severe accident rulemaking on future plants by utilizing standard plant licensing documentation. This paper, GE's submission, discusses the features of the design that prevent severe accidents from leading to core damage or that mitigate the effects of severe accidents should core damage occur. The quantification of the accident prevention and mitigation features, including those incorporated in the design since the accident at TMI, is provided by means of a comprehensive probabilistic risk assessment, which provides an analysis of the probability and consequences of postulated severe accidents

  9. Use of decision trees for evaluating severe accident management strategies in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclerar Engineering; Lee, Yongjin; Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of). School of Energy Systems Engineering

    2016-07-15

    Accident management strategies are defined to innovative actions taken by plant operators to prevent core damage or to maintain the sound containment integrity. Such actions minimize the chance of offsite radioactive substance leaks that lead to and intensify core damage under power plant accident conditions. Accident management extends the concept of Defense in Depth against core meltdown accidents. In pressurized water reactors, emergency operating procedures are performed to extend the core cooling time. The effectiveness of Severe Accident Management Guidance (SAMG) became an important issue. Severe accident management strategies are evaluated with a methodology utilizing the decision tree technique.

  10. Verification of criticality Safety for ETRR-2 Fuel Manufacturing pilot Plant (FMPP) at Inshas

    International Nuclear Information System (INIS)

    Aziz, M.; Gadalla, A.A.; Orabi, G.

    2006-01-01

    The criticality safety of the fuel manufacturing pilot plant (FMPP) at inshas is studied and analyzed during normal and abnormal operation conditions. the multiplication factor during all stages of the manufacturing processes is determined. several accident scenarios were simulated and the criticality of these accidents were investigated. two codes are used in the analysis : MCNP 4 B code, based on monte Carlo method, and CITATION code , based on diffusion theory. the results are compared with the designer calculations and satisfactory agreement were found. the results of the study indicated that the safety of the fuel manufacturing pilot plant is confirmed

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  12. Plant state identification using fuzzy logic in the framework of computerized accident management support (CAMS)

    International Nuclear Information System (INIS)

    Van Dyck, Claude

    1997-05-01

    CAMS (computerized accident management support) is a system that will provide assistance in case of accident in a nuclear power plant. In order to support the user in evaluating the plant state, it contains a state identification module. The state identification module provides high-level, qualitative information about the status of critical safety functions, about the availability of safety systems and about the occurrence of initiating events. This information is sent to the man-machine interface and to other CAMS modules. The state identification module is developed using a specific tool: GPS (Goal Processing System) which is based on the Goal Tree - Success Tree formalism. GPS is a tool designed to manage ''process related'' knowledge and aimed at process supervision via real-time acquisition of process variables. Fuzzy logic has been introduced in GPS in order to have smoother transitions between different states of critical safety functions and systems changes and to have a truth value associated to each piece of information provided to the user. The whole system has been tested, integrated with the rest of CAMS, on several accident scenarios. The test results are satisfactory. A brief comparison is made between the present work and previous related work at the HRP. (author)

  13. Consequences of the nuclear power plant accident at Chernobyl

    International Nuclear Information System (INIS)

    Ginzburg, H.M.; Reis, E.

    1991-01-01

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

  14. Tracing nuclear elements released by Fukushima Nuclear Power Plant accident

    Science.gov (United States)

    Tsujimura, M.; Onda, Y.; Abe, Y.; Hada, M.; Pun, I.

    2011-12-01

    Radioactive contamination has been detected in Fukushima and the neighboring regions due to the nuclear accident at Fukushima Daiichi Nuclear Power Plant (NPP) following the earthquake and tsunami occurred on 11th March 2011. The small experimental catchments have been established in Yamakiya district, Kawamata Town, Fukushima Prefecture, located approximately 35 km west from the Fukushima NPP. The tritium (3H) concentration and stable isotopic compositions of deuterium and oxygen-18 have been determined on the water samples of precipitation, soil water at the depths of 10 to 30 cm, groundwater at the depths of 5 m to 50 m, spring water and stream water taken at the watersheds in the recharge and discharge zones from the view point of the groundwater flow system. The tritium concentration of the rain water fell just a few days after the earthquake showed a value of approximately 17 Tritium Unit (T.U.), whereas the average concentration of the tritium in the precipitation was less than 5 T.U. before the Fukushima accident. The spring water in the recharge zone showed a relatively high tritium concentration of approximately 12 T.U., whereas that of the discharge zone showed less than 5 T.U. Thus, the artificial tritium was apparently injected in the groundwater flow system due to the Fukushima NPP accident, whereas that has not reached at the discharge zone yet. The monitoring of the nuclear elements is now on going from the view points of the hydrological cycles and the drinking water security.

  15. To improve nuclear plant safety by learning from accident's experience

    International Nuclear Information System (INIS)

    Matsumoto, Hidezo; Kida, Masanori; Kato, Hiroyuki; Hara, Shin-ichi

    1994-01-01

    The ultimate goal of this study is to produce an expert system that enables the experience (records and information) gained from accidents to be put to use towards improving nuclear plant safety. A number of examples have been investigated, both domestic and overseas, in which experience gained from accidents was utilized by utilities in managing and operating their nuclear power stations to improve safety. The result of investigation has been used to create a general 'basic flow' to make the best use of experience. The ultimate goal is achieved by carrying out this 'basic flow' with artificial intelligence (AI). To do this, it is necessary (1) to apply language analysis to process the source information (primary data base; domestic and overseas accident's reports) into the secondary data base, and (2) to establish an expert system for selecting (screening) significant events from the secondary data base. In the processing described in item (1), a multi-lingual thesaurus for nuclear-related terms become necessary because the source information (primary data bases) itself is multi-lingual. In the work described in item (2), the utilization of probabilistic safety assessment (PSA), for example, is a candidate method for judging the significance of events. Achieving the goal thus requires developing various new techniques. As the first step of the above long-term study project, this report proposes the 'basic flow' and presents the concept of how the nuclear-related AI can be used to carry out this 'basic flow'. (author)

  16. The prediction of the LWR plant accident based on the measured plant data

    International Nuclear Information System (INIS)

    Miettinen, J.; Schmuck, P.

    2005-01-01

    In case of accident affecting a nuclear reactor, it is essential to anticipate the possible development of the situation to efficiently succeed in emergency response actions, i.e. firstly to be early warned, to get sufficient information on the plant: and as far as possible. The ASTRID (Assessment of Source Term for Emergency Response based on Installation Data) project consists in developing a methodology: of expertise to; structure the work of technical teams and to facilitate cross competence communications among EP players and a qualified computer tool that could be commonly used by the European countries to reliably predict source term in case of an accident in a light water reactor, using the information available on the plant. In many accident conditions the team of analysts may be located far away from the plant experiencing the accident and their decision making is based on the on-line plant data transmitted into the crisis centre in an interval of 30 - 600 seconds. The plant condition has to be diagnosed based on this information, In the ASTRID project the plant status diagnostics has been studied for the European reactor types including BWR, PWR and VVER plants. The directly measured plant data may be used for estimations of the break size from the primary system and its locations. The break size prediction may be based on the pressurizer level, reactor vessel level, primary pressure and steam generator level in the case of the steam generator tube rupture. In the ASTRID project the break predictions concept was developed and its validity for different plant types and is presented in the paper, when the plant data has been created with the plant specific thermohydraulic simulation model. The tracking simulator attempts to follow the plant behavior on-line based on the measured plant data for the main process parameters and most important boundary conditions. When the plant state tracking fails, the plant may be experiencing an accident, and the tracking

  17. Framatome-ANP France UO2 fuel fabrication. Criticality safety analysis in the light of the JCO accident

    International Nuclear Information System (INIS)

    Doucet, M.; Zheng, S.; Mouton, J.; Porte, R.

    2003-01-01

    In France the 1999' Tokai Mura criticality accident in Japan had a big impact on the nuclear fuel manufacturing facility community. Moreover this accident led to a large public discussion about all the nuclear facilities. The French Safety Authorities made strong requirements to the industrials to revisit completely their safety analysis files mainly those concerning nuclear fuels treatments. The FRAMATOME-ANP production of its French low enriched (5 w/o) UO2 fuel fabrication plant (FBFC/Romans) exceeds 1000 metric tons a year. Special attention was given to the emergency evacuation plan that should be followed in case of a criticality accident. If a criticality accident happens, site internal and external radioprotection requirements need to have an emergency evacuation plan showing the different routes where the absorbed doses will be as low as possible for people. The French Safety Authorities require also an update of the old based neutron source term accounting for state of the art methodology. UO2 blenders units contain a large amount of dry powder strictly controlled by moderation; a hypothetical water leakage inside one of these apparatus is simulated by increasing the water content of the powder. The resulted reactivity insertion is performed by several static calculations. The French IRSN/CEA CRISTAL codes are used to perform these static calculations. The kinetic criticality code POWDER simulates the power excursion versus time and determines the consequent total energy source term. MNCP4B performs the source term propagation (including neutrons and gamma) used to determine the isodose curves needed to define the emergency evacuation plant. This paper deals with the approach FRAMATOME-ANP has taken to assess Safety Authorities demands using the more up to date calculation tools and methodology. (author)

  18. Analysis of articles in weekly magazines on scientific issues related to Fukushima nuclear power plant accident

    International Nuclear Information System (INIS)

    Sano, Kazumi; Kikuchi, Macoto

    2012-01-01

    The large area was polluted by the radioactive fallout released after the nuclear fuel meltdown of Fukushima first nuclear plant of Tokyo electric power company. The news media that reported the accident were required to have scientific knowledge on the structure of the nuclear reactor and the physics and health issues of the radioactivity. In this paper, we focus on how the weekly magazines reported this critical accident. The weekly magazines are not regarded as a neutral news media. Rather, their articles in general strongly reflect the editorial opinions. In this sense, the weekly magazines are 'biased media'. So, there are many points to discuss from the view point of the science communication. We analyze the articles appeared in the seven major weekly magazines published during the first half year after the earthquake. We found that the differences in the scientific literacy between magazines are reflected, for example, in selection of the experts who made comments in articles. (author)

  19. Characteristics of Hydrogen Monitoring Systems for Severe Accident Management at a Nuclear Power Plant

    Science.gov (United States)

    Petrosyan, V. G.; Yeghoyan, E. A.; Grigoryan, A. D.; Petrosyan, A. P.; Movsisyan, M. R.

    2018-02-01

    One of the main objectives of severe accident management at a nuclear power plant is to protect the integrity of the containment, for which the most serious threat is possible ignition of the generated hydrogen. There should be a monitoring system providing information support of NPP personnel, ensuring data on the current state of a containment gaseous environment and trends in its composition changes. Monitoring systems' requisite characteristics definition issues are considered by the example of a particular power unit. Major characteristics important for proper information support are discussed. Some features of progression of severe accident scenarios at considered power unit are described and a possible influence of the hydrogen concentration monitoring system performance on the information support reliability in a severe accident is analyzed. The analysis results show that the following technical characteristics of the combustible gas monitoring systems are important for the proper information support of NPP personnel in the event of a severe accident at a nuclear power plant: measured parameters, measuring ranges and errors, update rate, minimum detectable concentration of combustible gas, monitoring reference points, environmental qualification parameters of the system components. For NPP power units with WWER-440/270 (230) type reactors, which have a relatively small containment volume, the update period for measurement results is a critical characteristic of the containment combustible gas monitoring system, and the choice of monitoring reference points should be focused not so much on the definition of places of possible hydrogen pockets but rather on the definition of places of a possible combustible mixture formation. It may be necessary for the above-mentioned power units to include in the emergency operating procedures measures aimed at a timely heat removal reduction from the containment environment if there are signs of a severe accident phase

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

    International Nuclear Information System (INIS)

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

    1993-05-01

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

  1. Effects of the criticality accident at Tokai-mura on the public's attitude to nuclear power generation

    Energy Technology Data Exchange (ETDEWEB)

    Kitada, Atsuko [Institute of Social Research, Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan); Hayashi, Chikio [The Institute of Statistical Mathematics, Tokyo (Japan)

    2000-09-01

    The objective of our study was to clarify the effects on the public's attitude of nuclear power and the criticality accident that occurred at the JCO plant in Tokai-mura, Ibaraki Prefecture. For this purpose, we conducted an awareness survey in the Kansai and Kanto areas two months after the accident. Analysis was made on the basis of the comparison of the survey results with the data that the Institute of Nuclear Safety System had accumulated through continuous awareness surveys on nuclear power generation (regular surveys) since 1993. The public's reactions were twofold. On one hand, there were emotional reactions about accidents in nuclear facilities and a reduction in the sense of security. On the other hand, there were reactions concerning the image of nuclear power plant workers and demand on electricity utilities for enhanced employee education and training. The latter reactions correspond to the problems pointed out after the JCO accident. Regarding the utilization of nuclear power generation, the opinion that 'the utilization of nuclear power generation is unavoidable' accounts for 60% of those surveyed. With the opinion that 'nuclear power generation should be utilized' added, 70% of those surveyed take an affirmative attitude to nuclear power utilization. This situation has remained about the same since 1998, the year before the JCO accident. Using the quantification method III to analyze a number of questionnaires about nuclear power generation such as the anxiety about it, we determined overall attitude indexes regarding nuclear power to perform a time sequence comparison. The comparison shows that the attitude after the JCO accident tended to be more negative than in 1998. However, no significant difference in the overall indexes is seen between 1993 and 1998. Judging the comparison results on the basis of the time span starting in 1993 allows us to conclude that the JCO accident has not greatly contributed to worsening

  2. Effects of the criticality accident at Tokai-mura on the public's attitude to nuclear power generation

    International Nuclear Information System (INIS)

    Kitada, Atsuko; Hayashi, Chikio

    2000-01-01

    The objective of our study was to clarify the effects on the public's attitude of nuclear power and the criticality accident that occurred at the JCO plant in Tokai-mura, Ibaraki Prefecture. For this purpose, we conducted an awareness survey in the Kansai and Kanto areas two months after the accident. Analysis was made on the basis of the comparison of the survey results with the data that the Institute of Nuclear Safety System had accumulated through continuous awareness surveys on nuclear power generation (regular surveys) since 1993. The public's reactions were twofold. On one hand, there were emotional reactions about accidents in nuclear facilities and a reduction in the sense of security. On the other hand, there were reactions concerning the image of nuclear power plant workers and demand on electricity utilities for enhanced employee education and training. The latter reactions correspond to the problems pointed out after the JCO accident. Regarding the utilization of nuclear power generation, the opinion that 'the utilization of nuclear power generation is unavoidable' accounts for 60% of those surveyed. With the opinion that 'nuclear power generation should be utilized' added, 70% of those surveyed take an affirmative attitude to nuclear power utilization. This situation has remained about the same since 1998, the year before the JCO accident. Using the quantification method III to analyze a number of questionnaires about nuclear power generation such as the anxiety about it, we determined overall attitude indexes regarding nuclear power to perform a time sequence comparison. The comparison shows that the attitude after the JCO accident tended to be more negative than in 1998. However, no significant difference in the overall indexes is seen between 1993 and 1998. Judging the comparison results on the basis of the time span starting in 1993 allows us to conclude that the JCO accident has not greatly contributed to worsening the attitude towards nuclear

  3. Development of an accident diagnosis system using a dynamic neural network for nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Seung Jun; Kim, Jong Hyun; Seong, Poong Hyun

    2004-01-01

    In this work, an accident diagnosis system using the dynamic neural network is developed. In order to help the plant operators to quickly identify the problem, perform diagnosis and initiate recovery actions ensuring the safety of the plant, many operator support system and accident diagnosis systems have been developed. Neural networks have been recognized as a good method to implement an accident diagnosis system. However, conventional accident diagnosis systems that used neural networks did not consider a time factor sufficiently. If the neural network could be trained according to time, it is possible to perform more efficient and detailed accidents analysis. Therefore, this work suggests a dynamic neural network which has different features from existing dynamic neural networks. And a simple accident diagnosis system is implemented in order to validate the dynamic neural network. After training of the prototype, several accident diagnoses were performed. The results show that the prototype can detect the accidents correctly with good performances

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

    International Nuclear Information System (INIS)

    Abe, Kiyoharu.

    1995-05-01

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

  5. Consideration of severe accident issues for the General Electric BWR standard plant: Chapter 10

    International Nuclear Information System (INIS)

    Holtzclaw, K.W.

    1983-01-01

    In early 1982, the U.S. Nuclear Regulatory Commission (NRC) proposed a policy to address severe accident rulemaking on future plants by utilizing standard plant licensing documentation. GE provided appendices to the licensing documentation of its standard plant design, GESSAR II, which address severe accidents for the GE BWR/6 Mark III 238 nuclear island design. The GE submittals discuss the features of the design that prevent severe accidents from leading to core damage or that mitigate the effects of severe accidents should core damage occur. The quantification of the accident prevention and mitigation features, including those incorporated in the design since the accident at Three Mile Island (TMI), is provided by means of a comprehensive probabilistic risk assessment, which provides an analysis of the probability and consequences of postulated severe accidents

  6. Initial medical management of criticality accident victim; Conduite a tenir aux victimes d'un accident de criticite

    Energy Technology Data Exchange (ETDEWEB)

    Miele, A; Bebaron-Jacobs, L

    2005-07-01

    The extremely severe criticality accidents known to this day, and the subsequent deaths recorded (Sarov 1997 and Tokai Mura 1999), demonstrate the need for sustained surveillance and constant adapted training for the teams in charge of irradiated and/or contaminated victims. The aim of this work group, composed of occupational health services and associated medical biology laboratories, is to present, in leaflet format, the essential data on the documentation and the conduct to be held when facing the victims of a criticality accident. The studies of this work group confirm the difficulties involved in managing this type of accident, both from the dosimetric evaluation point of view and from the therapeutic management point of view. That is why several research themes and perspectives are developed. During the different phases of victim triage, the recommendations given on these leaflets describe the operational conducts to be held. This work will have to be updated according to the evolution in knowledge and means: short and long term effects of exposure to neutrons, multi-competence hospital cooperation, expertise networks related to dosimetric reconstitution. (authors)

  7. Initial medical management of criticality accident victim; Conduite a tenir aux victimes d'un accident de criticite

    Energy Technology Data Exchange (ETDEWEB)

    Miele, A.; Bebaron-Jacobs, L

    2005-07-01

    The extremely severe criticality accidents known to this day, and the subsequent deaths recorded (Sarov 1997 and Tokai Mura 1999), demonstrate the need for sustained surveillance and constant adapted training for the teams in charge of irradiated and/or contaminated victims. The aim of this work group, composed of occupational health services and associated medical biology laboratories, is to present, in leaflet format, the essential data on the documentation and the conduct to be held when facing the victims of a criticality accident. The studies of this work group confirm the difficulties involved in managing this type of accident, both from the dosimetric evaluation point of view and from the therapeutic management point of view. That is why several research themes and perspectives are developed. During the different phases of victim triage, the recommendations given on these leaflets describe the operational conducts to be held. This work will have to be updated according to the evolution in knowledge and means: short and long term effects of exposure to neutrons, multi-competence hospital cooperation, expertise networks related to dosimetric reconstitution. (authors)

  8. State of reaction on news media for JCO criticality accident on abroad

    International Nuclear Information System (INIS)

    Itoh, Takeshi

    1999-01-01

    The criticality accident, which occurred in JCO Tokai on September 30th 1999, was the first accident accompanied with serious radiation exposure to persons at Japanese nuclear facilities. As an evacuation order for local residents was issued, it caused uneasiness to the public. It also gave great impact to the foreign countries. In this report we have investigated the reactions in such countries, as U.S., France, Germany and U.K. by means of news media like TV, newspapers and magazines. Finding are as follows: They were all surprised to know the cause of the accident, which was by improper procedure of JCO workers. Because they couldn't imagine that such an accident might happen in such a high-tech country as Japan. The Japanese regulator was criticized for their insufficient criticality facility surveillance. There arose some questions for Japanese nuclear reliabilities. Because of the delayed announcement of the accident by Japanese public sector, anti-nuclear groups, like Greenpeace, NCI, etc., have a chance to carry on their campaign. The information from Japanese public sector was not enough to satisfy the foreign news media. We concluded that it is also necessary to develop effective information dissemination to overseas in case of a nuclear accident. (author)

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

    International Nuclear Information System (INIS)

    Fraix, J.

    1989-09-01

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

  10. Analysis to the criticality the storage and containers to the Juragua Nuclear Power Plant Fuel

    International Nuclear Information System (INIS)

    Guerra Valdes, R.

    1998-01-01

    Presently analysis the criticality the warehouses and containers the nuclear fuels in Juragua nuclear power plant the property multiplicity determined in these system and it is verified that for the geometry and operation conditions defined in the design as well as in accidents situations, the arrangement the fuel stays subcritical with an appropriate margin

  11. Compendium on neutron spectra in criticality accident dosimetry

    International Nuclear Information System (INIS)

    Ing, H.

    1978-01-01

    Graphical and tabulated neutron spectra are presented: from selected critical assemblies; from critical solutions; of fission neutrons through shielding; of H 2 O-moderated fission neutrons through shielding; of D 2 O-moderated fission neutrons through shielding; of fission neutrons reflected from various materials; from the D(T, 4 He)n reaction (''14 MeV'' neutrons) through shielding and of ''14 MeV'' neutrons reflected from various materials

  12. Consideration of Command and Control Performance during Accident Management Process at the Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Nisrene M. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Sok Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    The accident at the Fukushima Daiichi nuclear power plants shifted the nuclear safety paradigm from risk management to on-site management capability during a severe accident. The kernel of on-site management capability during an accident at a nuclear power plant is situation awareness and agility of command and control. However, little consideration has been given to accident management. After the events of September 11, 2001 and the catastrophic Fukushima nuclear disaster, agility of command and control has emerged as a significant element for effective and efficient accident management, with many studies emphasizing accident management strategies, particularly man-machine interface, which is considered a key role in ensuring nuclear power plant safety during severe accident conditions. This paper proposes a conceptual model for evaluating command and control performance during the accident management process at a nuclear power plant. Communication and information processing while responding to an accident is one of the key issues needed to mitigate the accident. This model will give guidelines for accurate and fast communication response during accident conditions.

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

    International Nuclear Information System (INIS)

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

    1993-10-01

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

  14. Application of Whole Body Counter to Neutron Dose Assessment in Criticality Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kurihara, O.; Tsujimura, N.; Takasaki, K.; Momose, T.; Maruo, Y. [Japan Nuclear Cycle Development Institute, Tokai (Japan)

    2001-09-15

    Neutron dose assessment in criticality accidents using Whole Body Counter (WBC) was proved to be an effective method as rapid neutron dose estimation at the JCO criticality accident in Tokai-mura. The 1.36MeV gamma-ray of {sup 24}Na in a body can be detected easily by a germanium detector. The Minimum Detectable Activity (MDA) of {sup 24}Na is approximately 50Bq for 10minute measurement by the germanium-type whole body counter at JNC Tokai Works. Neutron energy spectra at the typical shielding conditions in criticality accidents were calculated and the conversion factor, whole body activity-to-organ mass weighted neutron absorbed dose, corresponding to each condition were determined. The conversion factor for uncollied fission spectrum is 7.7 [(Bq{sup 24}Na/g{sup 23}Na)/mGy].

  15. JAEA's activities relating the Fukushima Nuclear Plant accident

    International Nuclear Information System (INIS)

    Tagawa, Akihiro

    2012-01-01

    JAEA started the activities relating to the Fukushima nuclear plant accident immediately after the Great East Japan Earthquake. The Office of Fukushima Partnership Operations for Environmental Remediation was opened and the JAEA staff was stationed as the base of cooperation with other organizations. It is conducting environmental radiation monitoring, environmental radioactivity analyses, resident public consulting, and demonstration of decontamination technology. Experts of JAEA are providing technical advice and supports to the Nuclear Safety Commission of Japan and the Ministry of Education, Culture and Sports. Furthermore, the water radiolysis leading to hydrogen gas evolution by Cs 137 adsorbed zeolite and the technique for radioactive waste process and its disposal of fuel debris are being studied. JAEA's Nuclear Emergency Assistance and Training Center (NEAT) is acting as a center of these supporting activities of JAEA. (S. Ohno)

  16. The link between off-site-emergency planning and plant-internal accident management

    Energy Technology Data Exchange (ETDEWEB)

    Braun, H.; Goertz, R.

    1995-02-01

    A variety of accident management measures has been developed and implemented in the German nuclear power plants. They constitute a fourth level of safety in the defence-in-depth concept. The containment venting system is an important example. A functioning link with well defined lines of communication between plant-internal accident management and off-site disaster emergency planning has been established.

  17. Instructions on the nuclear critical accident and how to correspond to future

    International Nuclear Information System (INIS)

    Kimura, Itsuro

    2000-01-01

    The nuclear accident occurred at Tokai Works of the JCO Co., Ltd. is a simple but disallowable one formed by neglecting possibility of nuclear fission chain reaction, one of three dangers specific to nuclear power facilities and by feeding middle concentrated uranium solution with more than critical mass into a precipitation tank. As a man consumed most of his life to nuclear power, it is to occur a critical accident forming about 50 previous examples in the world and about 10 victims at Tokai-mura the most earnestly promoting its experiment and analysis and to generate new victims, what was thought to be the most regrettable in this accident. How the previous experiences and results in Tokai-mura could be transmitted to the JCO Co., Ltd. ? This was a large alarm-bell for persons engaging to R and D on nuclear power. As this accident was much deplorable and apological for the common public, it must be carried out to thoroughly analyze its causes, to establish its future responses, and to promote its essential countermeasures. As it is important to open informations on its contents, it is hopeful not to over-exaggerate and over-differentiate the accident, to calmly and scientifically analyze the risk as well as in the other accidents, and to construct actually effective countermeasures. (G.K.)

  18. Evaluation of Neutron Response of Criticality Accident Alarm System Detector to Quasi-Monoenergetic 24 keV Neutrons

    Science.gov (United States)

    Tsujimura, Norio; Yoshida, Tadayoshi; Yashima, Hiroshi

    The criticality accident alarm system (CAAS), which was recently developed and installed at the Japan Atomic Energy Agency's Tokai Reprocessing Plant, consists of a plastic scintillator combined with a cadmium-lined polyethylene moderator and thereby responds to both neutrons and gamma rays. To evaluate the neutron absorbed dose rate response of the CAAS detector, a 24 keV quasi-monoenergetic neutron irradiation experiment was performed at the B-1 facility of the Kyoto University Research Reactor. The detector's evaluated neutron response was confirmed to agree reasonably well with prior computer-predicted responses.

  19. Evaluation of neutron response of criticality accident alarm system detector to quasi-monoenergetic 24 keV neutrons

    International Nuclear Information System (INIS)

    Tsujimura, Norio; Yoshida, Tadayoshi; Yashima, Hiroshi

    2016-01-01

    The criticality accident alarm system (CAAS), which was recently developed and installed at the Japan Atomic Energy Agency's Tokai Reprocessing Plant, consists of a plastic scintillator combined with a cadmium-lined polyethylene moderator and thereby responds to both neutrons and gamma rays. To evaluate the neutron absorbed dose rate response of the CAAS detector, a 24 keV quasi-monoenergetic neutron irradiation experiment was performed at the B-1 facility of the Kyoto University Research Reactor. The detector's evaluated neutron response was confirmed to agree reasonably well with prior computer-predicted responses. (author)

  20. Organizational factors and reoccurrence protection on the JCO nuclear critical accident

    International Nuclear Information System (INIS)

    Takano, Kenichi

    2000-01-01

    A nuclear critical accident formed at a nuclear fuel conversion factory in Tokai-mura on September, 1999 became gradually clear not to be a simple human error formed at a level of workmen but to be an organizational error or accident relating to various organizational factors. As a nuclear power facility adopts a depth protection system fundamentally, a large accident with serious danger would not form only by a single trouble and a human error and unless some factors overlaps. By reviewing recent serious accidents and troubles, all of them seem to have a keyword of 'organizational factor'. In the JCO accident, there are some organizational factors such as a climate deviating from a manual, insufficient and loose check against change of procedure, reduction of operators from a reason of profit priority, attitude on priority of working efficiency, and so forth, which are partially common to the Chernobyl accident. Recently, accidents and troubles impossible to make them a cause of simple human error by a person but to have to say an organizational error, have increased. This trend seems to depend upon not only complication and scale-up of technology system but also graduate change of social and management systems operating them. Therefore, it seems to be necessary to introduce a concept of depth protection (multiple protection) in order to keep its reliability and safety when complicating and scaling-up of system. (G.K.)

  1. Questionnaire survey report about the criticality accident at a nuclear fuel processing facility

    International Nuclear Information System (INIS)

    2000-01-01

    The Radiation Protection Section of the Japanese Society of Radiological Technology conducted a questionnaire survey on the criticality accident at the nuclear fuel processing facility in Tokai village on September 30, 1999 in order to identify factors related to the accident and consider countermeasures to deal with such accidents. The questionnaire was distributed to 347 members (122 facilities) of the Japanese Society of Radiological Technology who were working or living in Ibaraki Prefecture, and replies were obtained from 104 members (75 facilities). Questions to elicit the opinions of individuals were as following: method of obtaining information about the accident, knowledge about radiation, opinions about the accident, and requests directed to the Society. Questions regarding facilities concerned the following: communication after the accident, requests for dispatch to the accident site, and possession of radiometry devices. In regard to acquisition of information, 91 of the 104 members (87.5%) answered 'television or radios' followed by newspapers. Forty-five of 101 members were questioned about radiation exposure and radiation effects by the public. There were many opinions that accurate news should be provided rapidly, by the mass media. Many members (75%) felt that they lacked knowledge about radiation, reconfirming the importance of education and instruction concerning radiation. Dispatch was requested of 36 of the 75 facilities (48%), and 44 of 83 facilities (53%) owned radiometry instruments. (K.H.)

  2. Critical examination of emergency plans for nuclear accidents

    International Nuclear Information System (INIS)

    Catsaros, Nicolas.

    1986-08-01

    An analysis of emergency plans of various countries for nuclear installations on- and off-site emergency preparedness is presented. The analysis is focused on the off-site organization and countermeasures to protect public health and safety. A critical examination of the different approaches is performed and recommendations for effectiveness improvement and optimization are formulated. (author)

  3. Rules of thumb from plant data to estimate the public consequences of a nuclear power plant accident

    International Nuclear Information System (INIS)

    Baggenstos, M.; Uboldi, P.; Schulz, R.

    2001-01-01

    In an accident situation with core degradation there is typically a pre-phase in which the radioactivity is inside the primary system and the containment before a release to the environment. The assessment of the possible risk to the public in this situation must be based on the situation inside the plant (violation of safety parameters) and a forecast of the containment behaviour to be expected. To obtain a first quick estimate of the source term and, therefore, the off-site dose rules of thumb were established which should fulfil the following purposes: assessment of the danger, estimated from plant data, estimation of the dose at the critical point outside based on the dose rate inside the containment. The rules of thumb which were introduced in Switzerland, are explained. The assessment is based on roughly 30 plant parameters which are transmitted in case of an accident in real-time. The rules were designed in such a way that they rely on simply determined parameters such core exit temperature or dose rate in the containment. (orig.) [de

  4. Severe accident risks: An assessment for five US nuclear power plants: Appendices A, B, and C

    International Nuclear Information System (INIS)

    1990-12-01

    This report summarizes an assessment of the risks from severe accidents in five commercial nuclear power plants in the United States. These risks are measured in a number of ways, including: the estimated frequencies of core damage accidents from internally initiated accidents and externally initiated accidents for two or the plants; the performance of containment structures under severe accident loadings; the potential magnitude of radionuclide release and offsite consequences of such accidents; and the overall risk (the product of accident frequencies and consequences). Supporting this summary report are a large number of reports written under contract to NRC that provide the detailed discussion of the methods used and results obtained in these risk studies. Volume 2 of this report contains three appendices, providing greater detail on the methods used, an example risk calculation, and more detailed discussion of particular technical issues found important in the risk studies

  5. Severe accident risks: An assessment for five US nuclear power plants

    International Nuclear Information System (INIS)

    1991-01-01

    This report summarizes an assessment of the risks from severe accidents in five commercial nuclear power plants in the United State. These risks are measured in a number of ways, including: the estimated frequencies of core damage accidents from internally initiated accidents and externally initiated accidents for two of the plants; the performance of containment structures under severe accident loadings; the potential magnitude of radionuclide releases and offsite consequences of such accidents; and the overall risk (the product of accident frequencies and consequences). Supporting this summary report are a large number of reports written under contract to NRC that provide the detailed discussion of the methods used and results obtained in these risk studies. This report, Volume 3, contains two appendices. Appendix D summarizes comments received, and staff responses, on the first (February 1987) draft of NUREG-1150. Appendix E provides a similar summary of comments and responses, but for the second (June 1989) version of the report

  6. Using modular neural networks to monitor accident conditions in nuclear power plants

    International Nuclear Information System (INIS)

    Guo, Z.

    1992-01-01

    Nuclear power plants are very complex systems. The diagnoses of transients or accident conditions is very difficult because a large amount of information, which is often noisy, or intermittent, or even incomplete, need to be processed in real time. To demonstrate their potential application to nuclear power plants, neural networks axe used to monitor the accident scenarios simulated by the training simulator of TVA's Watts Bar Nuclear Power Plant. A self-organization network is used to compress original data to reduce the total number of training patterns. Different accident scenarios are closely related to different key parameters which distinguish one accident scenario from another. Therefore, the accident scenarios can be monitored by a set of small size neural networks, called modular networks, each one of which monitors only one assigned accident scenario, to obtain fast training and recall. Sensitivity analysis is applied to select proper input variables for modular networks

  7. Hazard Identification, Risk Assessment and Risk Control (HIRARC Accidents at Power Plant

    Directory of Open Access Journals (Sweden)

    Ahmad Asmalia Che

    2016-01-01

    Full Text Available Power plant had a reputation of being one of the most hazardous workplace environments. Workers in the power plant face many safety risks due to the nature of the job. Although power plants are safer nowadays since the industry has urged the employer to improve their employees’ safety, the employees still stumble upon many hazards thus accidents at workplace. The aim of the present study is to investigate work related accidents at power plants based on HIRARC (Hazard Identification, Risk Assessment and Risk Control process. The data were collected at two coal-fired power plant located in Malaysia. The finding of the study identified hazards and assess risk relate to accidents occurred at the power plants. The finding of the study suggested the possible control measures and corrective actions to reduce or eliminate the risk that can be used by power plant in preventing accidents from occurred

  8. The critical safety functions and plant operation

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Church, J.F.; Porter, N.J.; Cross, M.T.; Guinn, W.M.

    1981-01-01

    The paper outlines the operator's role in nuclear safety and introduces the concept of ''safety functions''. Safety functions are a group of actions that prevent core melt or minimize radiation releases to the general public. They can be used to provide a hierarchy of practical plant protection that an operator should use. ''An accident identical to that at Three Mile Island is not going to happen again'', said the Rogovin investigators. The concepts put forward in this paper are intended to help the operator avoid serious consequence from the next unexpected threat. On the basis of the safety evaluation, the operator has three roles in assuring that the consequences of an event will be no worse than the predicted acceptable results. These three operator roles are: first, maintain plant setup in readiness to properly respond; second, operate the plant in a manner such that fewer, milder events minimize the frequency and the severity of adverse events; third, the operator needs to monitor the plant to verify that the safety functions are accomplished. The operator needs a systematic approach to mitigating the consequences of an event. The concept of ''safety function'' introduces that systematic approach and prevents a hierarchy of protection. If the operator has difficulty in identifying an event for any reason, the systematic safety function approach allows ones to accomplish the overall path of mitigating consequences. There are ten identified functions designed to protect against core melt, preserve containment integrity, prevent indirect release of radioactivity, and maintain vital auxiliaries needed to support the other safety functions. The paper describes in detail the operator's role and the safety functions, and provides many examples of the use of alternative success paths to accomplish the safety function

  9. Heat and fluid flow in accident of Fukushima Daiichi Nuclear Power Plant, Unit 2. Accident scenario based on thermodynamic model

    International Nuclear Information System (INIS)

    Maruyama, Shigenao

    2012-01-01

    An accident scenario of Fukushima Daiichi Nuclear Power Plant, Unit 2 is analyzed from the data open to the public. Phase equilibrium process model was introduced that the vapor and water are at saturation point in the vessels. Proposed accident scenario agrees very well with the data of the plant parameters obtained just after the accident. The estimation describes that the rupture time of the reactor pressure vessel (RPV) was at 22:50 14/3/2011. The estimation shows that the rupture time of the pressure containment vessel (RCP) was at 7:40 15/3/2011. These estimations are different from the ones by TEPCO, however; many measured evidences show good accordance with the present scenario. (author)

  10. Accident at Three Mile Island nuclear power plant and lessons learned

    International Nuclear Information System (INIS)

    Ashrafi, A.; Farnoudi, F.; Tochai, M.T.M.; Mirhabibi, N.

    1986-01-01

    On March 28, 1979, the TMI, unit 2 nuclear power plant experienced a loss of coolant accident (LOCA) which has had a major impact among the others, upon the safety of nuclear power plants. Although a small part of the reactor core melted in this accident, but due to well performance of the vital safety equipment, there was no serious radioactivity release to the environment, and the accident has had no impact on the basic safety goals. A brief scenario of the accident, its consequences and the lessons learned are discussed

  11. Calculation code used in criticality analyses for the accident of JCO precipitation tank

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori

    2000-01-01

    In order to evaluate nuclear features on criticality accident formed at the nuclear fuel processing facility in Tokai Works of the JCO, Ltd. (JCO), in Tokai-mura, Ibaraki prefecture, dynamic analyses to calculate output change after occurring the accident as well as criticality analyses to calculate reactivity added to precipitation tank, were carried out according to scenario on accident formation. For the criticality analyses, a continuous energy Monte Carlo code MCNP was used to carry out calculation of reactivity fed into the precipitation tank as correctly as possible. And, SRAC code system was used for calculation on temperature and void reactivity coefficients, effective delayed neutron ratio beta eff , and instantaneous neutron generation time required for parameters controlling transition features at criticality accident. In addition, for the dynamic analyses, because of necessity of considering on volume expansion of solution fuels used as exothermic body and radiation decomposition gas forming into solution, output behavior, numbers of nuclear fission, and so forth at initial burst portion were calculated by using TRACE and quasi-regular code, at a center of AGNES-2 promoting on its development in JAERI. Here were reported on outlines and an analysis example on calculation code using for the nuclear features evaluation. (G.K.)

  12. Hypothetical accidents of light-water moderated nuclear power plants in the framework of emergency planning

    International Nuclear Information System (INIS)

    1979-07-01

    Hypothetical accidents in nuclear power plants are events which by definition can have a devastating impact on the surroundings of the plant. Apart from an adequate plant design, the protection of the population in case of an accident is covered by the emergency planning. Of major importance are the measures for the short-term emergency protection. The decision on whether these measures are applied has to be based on appropriate measurements within the plant. The aim and achieved result of this investigation is to specify accident types. They serve as operational decision making criteria to determine the necessary measurements for analysing the accident in the accident situation, and to provide indications for choosing the suitable strategy for the protection measures. (orig.) [de

  13. Incorporation of severe accidents in the licensing of nuclear power plants

    International Nuclear Information System (INIS)

    Alvarenga, Marco Antonio Bayout; Rabello, Sidney Luiz

    2011-01-01

    Severe accidents are the result of multiple faults that occur in nuclear power plants as a consequence from the combination of latent failures and active faults, such as equipment, procedures and operator failures, which leads to partial or total melting of the reactor core. Regardless of active and latent failures related to the plant management and maintenance, aspects of the latent failures related to the plant design still remain. The lessons learned from the TMI accident in the U.S.A., Chernobyl in the former Soviet Union and, more recently, in Fukushima, Japan, suggest that severe accidents must necessarily be part of design-basis of nuclear power plants. This paper reviews the normative basis of the licensing of nuclear power plants concerning to severe accidents in countries having nuclear power plants under construction or in operation. It was addressed not only the new designs of nuclear power plants in the world, but also the design changes in plants that are in operation for decades. Included in this list are the Brazilian nuclear power plants, Angra-1, Angra-2, and Angra-3. This paper also reviews the current status of licensing in Brazil and Brazilian standards related to severe accidents. It also discusses the impact of severe accidents in the emergency plans of nuclear power plants. (author)

  14. Incorporation of severe accidents in the licensing of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Alvarenga, Marco Antonio Bayout; Rabello, Sidney Luiz, E-mail: bayout@cnen.gov.b, E-mail: sidney@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN) Rio de Janeiro, RJ (Brazil)

    2011-07-01

    Severe accidents are the result of multiple faults that occur in nuclear power plants as a consequence from the combination of latent failures and active faults, such as equipment, procedures and operator failures, which leads to partial or total melting of the reactor core. Regardless of active and latent failures related to the plant management and maintenance, aspects of the latent failures related to the plant design still remain. The lessons learned from the TMI accident in the U.S.A., Chernobyl in the former Soviet Union and, more recently, in Fukushima, Japan, suggest that severe accidents must necessarily be part of design-basis of nuclear power plants. This paper reviews the normative basis of the licensing of nuclear power plants concerning to severe accidents in countries having nuclear power plants under construction or in operation. It was addressed not only the new designs of nuclear power plants in the world, but also the design changes in plants that are in operation for decades. Included in this list are the Brazilian nuclear power plants, Angra-1, Angra-2, and Angra-3. This paper also reviews the current status of licensing in Brazil and Brazilian standards related to severe accidents. It also discusses the impact of severe accidents in the emergency plans of nuclear power plants. (author)

  15. Mitigation of Hydrogen Hazards in Severe Accidents in Nuclear Power Plants

    International Nuclear Information System (INIS)

    2011-07-01

    Consideration of severe accidents in nuclear power plants is an essential component of the defence in depth approach in nuclear safety. Severe accidents have very low probabilities of occurring, but may have significant consequences resulting from the degradation of nuclear fuel. The generation of hydrogen and the risk of hydrogen combustion, as well as other phenomena leading to overpressurization of the reactor containment in case of severe accidents, represent complex safety issues in relation to accident management. The combustion of hydrogen, produced primarily as a result of heated zirconium metal reacting with steam, can create short term overpressure or detonation forces that may exceed the strength of the containment structure. An understanding of these phenomena is crucial for planning and implementing effective accident management measures. Analysis of all the issues relating to hydrogen risk is an important step for any measure that is aimed at the prevention or mitigation of hydrogen combustion in reactor containments. The main objective of this publication is to contribute to the implementation of IAEA Safety Standards, in particular, two IAEA Safety Requirements: Safety of Nuclear Power Plants: Design and Safety of Nuclear Power Plants: Operation. These Requirements publications discuss computational analysis of severe accidents and accident management programmes in nuclear power plants. Specifically with regard to the risk posed by hydrogen in nuclear power reactors, computational analysis of severe accidents considers hydrogen sources, hydrogen distribution, hydrogen combustion and control and mitigation measures for hydrogen, while accident management programmes are aimed at mitigating hydrogen hazards in reactor containments.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  17. Containment pressure monitoring method after severe accident in nuclear power plant

    International Nuclear Information System (INIS)

    Luo Chuanjie; Zhang Shishui

    2011-01-01

    The containment atmosphere monitoring system in nuclear power plant was designed on the basis of design accident. But containment pressure will increase greatly in a severe accident, and pressure instrument in the containment can't satisfy the monitoring requirement. A new method to monitor the pressure change in the containment after a severe accident was considered, through which accident soften methods can be adopted. Under present technical condition, adding a pressure monitoring channel out of containment for post-severe accident is a considerable method. Daya Bay Nuclear Power Plant implemented this modification, by which the containment release time can be delayed during severe accident, and nuclear safety can be increased. After analysis, this method is safe and feasible. (authors)

  18. Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, Dwight A [ORNL; Poore III, Willis P [ORNL

    2015-01-01

    The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Mark I plant for those instrumentation systems considered most important for accident management purposes.

  19. Applying Functional Modeling for Accident Management of Nuclear Power Plant

    DEFF Research Database (Denmark)

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    Modeling is given and a detailed presentation of the foundational means-end concepts is presented and the conditions for proper use in modelling accidents are identified. It is shown that Multilevel Flow Modeling can be used for modelling and reasoning about design basis accidents. Its possible role...... for information sharing and decision support in accidents beyond design basis is also indicated. A modelling example demonstrating the application of Multilevel Flow Modelling and reasoning for a PWR LOCA is presented...

  20. Applying Functional Modeling for Accident Management of Nucler Power Plant

    DEFF Research Database (Denmark)

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    Modeling is given and a detailed presentation of the foundational means-end concepts is presented and the conditions for proper use in modelling accidents are identified. It is shown that Multilevel Flow Modeling can be used for modelling and reasoning about design basis accidents. Its possible role...... for information sharing and decision support in accidents beyond design basis is also indicated. A modelling example demonstrating the application of Multilevel Flow Modelling and reasoning for a PWR LOCA is presented....

  1. Main lessons based on the Chernobyl nuclear power plant accident liquidation experience

    International Nuclear Information System (INIS)

    Vasil'chenko, V.N.; Nosovskij, A.V.

    2006-01-01

    The authors review the main lessons of the Chernobyl nuclear power plant accident and the liquidation of its consequences in the area of the nuclear reactors safety operation, any major accident management, liquidation accident consequences criteria, emergency procedures, preventative measures and treatment irradiated victims, the monitoring methods etc. The special emphasis is put on the questions of the emergency response and the antiaccidental measures planning in frame of international cooperation program

  2. Human factors review for nuclear power plant severe accident sequence analysis

    International Nuclear Information System (INIS)

    Krois, P.A.; Haas, P.M.

    1985-01-01

    The paper discusses work conducted to: (1) support the severe accident sequence analysis of a nuclear power plant transient based on an assessment of operator actions, and (2) develop a descriptive model of operator severe accident management. Operator actions during the transient are assessed using qualitative and quantitative methods. A function-oriented accident management model provides a structure for developing technical operator guidance on mitigating core damage preventing radiological release

  3. Off-gas and air cleaning systems for accident conditions in nuclear power plants

    International Nuclear Information System (INIS)

    1993-01-01

    This report surveys the design principles and strategies for mitigating the consequences of abnormal events in nuclear power plants by the use of air cleaning systems. Equipment intended for use in design basis accident and severe accident conditions is reviewed, with reference to designs used in IAEA Member States. 93 refs, 48 figs, 23 tabs

  4. Radiological dose assessment for bounding accident scenarios at the Critical Experiment Facility, TA-18, Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    1991-09-01

    A computer modeling code, CRIT8, was written to allow prediction of the radiological doses to workers and members of the public resulting from these postulated maximum-effect accidents. The code accounts for the relationships of the initial parent radionuclide inventory at the time of the accident to the growth of radioactive daughter products, and considers the atmospheric conditions at time of release. The code then calculates a dose at chosen receptor locations for the sum of radionuclides produced as a result of the accident. Both criticality and non-criticality accidents are examined

  5. Review of design criteria for Criticality Accident Alarm System (CAAS) used in Fuel Reprocessing Facility

    International Nuclear Information System (INIS)

    Chandrasekaran, S.; Basu, Pew; Sivasubramaniyan, K.; Venkatraman, B.

    2016-01-01

    Though fuel cycle facilities handling fissile materials are designed with careful criticality safety analysis, the criticality accident cannot be ruled out completely. Criticality Accident Alarm System (CAAS) is being installed as part of criticality safety management in fuel cycle facilities. CAAS system being used in India, is ECIL make, ionization chamber based gamma detector, which houses three identical detectors and works on 2/3 logic. As per ISO 7753 and ANSI/ANS-8.3, the CAAS must be designed to be capable of detecting any minimum accident occurs which could be of concern. Based on this, alarm limit used in CAAS is: 4 R/h (fast transient excursion) and 3 mR in 0.5 sec (slow excursion). In case of reprocessing facilities wherein process tanks located in heavy shielding, identification of CAAS installation locations require detailed radiation transport calculations. A study has been taken to estimate the gamma dose rate from thick concrete hot cells in order to determine the locations of CAAS to meet the present design criteria of alarm limit

  6. Evaluation of the 17 June 1997 Criticality Accident at Arzamas-16

    International Nuclear Information System (INIS)

    Morris Klein

    1999-01-01

    On June 17, 1997, a critically accident occurred at Arzamas-16, which resulted in the death (within three days) of A. N. Zakharov, a Russian scientist with 20 years' experience conducting multiassembly experiments. In this case, the multiplying assembly was a fast metal system consisting of a 235 U (90% enriched) core and a copper reflector. According to the Russian press, ''Zakharov misjudged the degree of criticality of the breeding system and committed several gross violations of regulations.'' As we see it, there were three major causes of this accident. First, the experiment was flawed by Zakharov's misreading of the appropriate size of the assembly, which he took from a notebook that described the old experiment he was attempting to repeat. Second, he disregarded the appropriate procedures and safety regulations. Third, these two mistakes were compounded by an improperly set audible alarm system and Zakharov's unsafe use of the table. We also discuss our reconstruction of the accident based on information given by the Russians to US scientists and information culled from Russian newspaper and magazine articles. We also describe our thoughts on the behavior of the assembly following the accident and the radiation dose level Zakharov may have received. These levels match values we have lately obtained from translations of Russian news articles. This accident clearly points out the penalty for weak administrative control of work with multiplying systems. Criticality experimentation requires formality of operation. The experimenter, his peers, and a trained safety person need to document that they understand the experiment and how it will be conducted. Knowing that the experiment was successfully run several decades ago does not justify bypassing a safety evaluation

  7. ANSI/ANS-8.23-1997: nuclear criticality accident emergency planning and response

    International Nuclear Information System (INIS)

    Baker, J.S.

    2004-01-01

    American National Standard ANSUANS-8.23 was developed to expand upon the basic emergency response guidance given in American National Standard, 'Administrative Practices for Nuclear Criticality Safety' ANSI/ANS-8.19-1996 (Ref. 1). This standard provides guidance for minimizing risks to personnel during emergency response to a nuclear criticality accident outside reactors. This standard is intended to apply to those facilities for which a criticality accident alarm system, as specified in American National Standard, 'Criticality Accident Alarm System', ANSI/ANS-8.3-1997 (Ref. 2) is in use. The Working Group was established in 1990, with Norman L. Pruvost as chairman. The Working Group had up to twenty-three members representing a broad range of the nuclear industry, and has included members from Canada, Japan and the United Kingdom. The initial edition of ANSI/ANS-8.23 was approved by the American National Standards Institute on December 30, 1997. It provides guidance for the following topics: (1) Management and technical staff responsibilities; (2) Evaluation of a potential criticality accident; (3) Emergency plan provisions; (4) Evacuation; (5) Re-entry, rescue and stabilization; and (6) Classroom training, exercises and evacuation drills. This guidance is not for generic emergency planning issues, but is specific to nuclear criticality accidents. For example, it assumes that an Emergency Plan is already established at facilities that implement the standard. During the development of the initial edition of ANSI/ANS-8.23, each Working Group member evaluated potential use of the standard at a facility with which the member was familiar. This revealed areas where a facility could have difficulty complying with the standard. These reviews helped identify and eliminate many potential problems and ambiguities with the guidance. The Working Group has received very limited feedback from the user community since the first edition of the standard was published. Suggestions

  8. WIPP conceptual design report. Addendum G. Accident analysis for Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Shefelbine, H.C.; Metcalf, J.H.

    1977-06-01

    The types of accidents or risks pertinent to the Waste Isolation Pilot Plant (WIPP) are presented. Design features addressing these risks are discussed. Also discussed are design features that protect the public

  9. Risk informed analysis of training effectiveness for mitigating accidents of nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Chang Ju

    2012-01-01

    A critical area for deriving expected benefits from training and exercise is the measurement of 'training effectiveness'-how well the training inputs are serving the intended purpose. This aspect is often neglected by nuclear organizations, saying that measurement is difficult. However, I believe that a technique in nuclear society has developed sufficiently to measure most important aspects of training by way of human reliability analysis (HRA) used in probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The consequences of errors caused by lack of training can be evaluated in terms of the overall vulnerability to human error of the facility under consideration. This study presents current situation and considerations for measures of robustness on nuclear accidents and HRA technique on the training effectiveness. In view of risk informed approach with this consideration and an example case, I'd like to identify appropriate relationship between risk measures of robustness and training effectiveness

  10. Radiation protection service for a nucleonic control system of continuous casting plant after events of accident

    International Nuclear Information System (INIS)

    Chakrabarti, Santanu; Massand, O.P.

    1998-01-01

    Extensive use of nucleonic control systems like level controllers was observed during radiation protection surveys in industries such as refineries, steel plants etc., located in the eastern region of India. There were two accidents at continuous casting plant in 1995 which affected the nucleonic control system installed in 1992. The authorities contacted Bhabha Atomic Research Centre (BARC) for radiation protection surveys for the involved nucleonic gauges. The present paper describes the radiation protection services rendered by BARC during such accidents. (author)

  11. Behaviour of a pressurized-water reactor nuclear power plant during loss-of-coolant accident

    International Nuclear Information System (INIS)

    Adam, E.; Carl, H.; Kubis, K.

    1979-01-01

    Starting from the foundation of the design basis accident in a PWR-type nuclear power plant - Loss of Coolant Accident -the actual status of the processes to be expected in the reactor are described. Operating behaviour of the heat removal system and efficiency of the safety systems are evaluated. Final considerations are concerned with the overall behaviour of the plant under such conditions. Probable failures, shut down times and possibilities of repair are estimated. (author)

  12. Nuclear power plant severe accident research plan. Revision 1

    International Nuclear Information System (INIS)

    Marino, G.P.

    1986-04-01

    Subsequent to the Three Mile Island Unit 2 accident, recommendations were made by a number of review committees to consider regulatory changes which would provide better protection of the public from severe accidents. Over the past six years a major research effort has been underway by the NRC to develop an improved understanding of severe accidents and to provide a technical basis to support regulatory decisions. The purpose of this report is to describe current plans for the completion and extension of this research in support of ongoing regulatory actions in this area

  13. Study on actions for social acceptance of a nuclear power plant incident/accident

    International Nuclear Information System (INIS)

    Kotani, Fumio; Tsukada, Tetsuya; Hiramoto, Mitsuru; Nishimura, Naoyuki

    1998-01-01

    When an incident/accident has occurred, dealing technically with it in an appropriate way is essential for social acceptance. One of the most important actions that are expected from the plant representative is to provide, without delay, each of the concerned authorities and organizations with full information concerning the incident/accident, while necessary technical measures are being implemented. While the importance of socially dealing with the incident/accident is widely recognized, up to now there have been no attempts to study previous incidents/accidents cases from the social sciences viewpoint. Therefore, in the present study is a case study of the incident/accident that occurred in 1991 at the No.2 Unit of the Mihama Nuclear Plant of Kansai Power Co., Ltd.. The data used in the present study is based on intensive interview of the staff involved in this incident/accident. The purpose of the study was to shed light on the conditions necessary for maintaining and improving the skill of the plant representative when dealing with social response in case of an incident/accident. The results of the present study has led to a fuller recognition of the importance of the following factors: On the personal level: 1) recognition of personal accountability, 2) complete disclosure of information concerning the incident/accident. On the organizational level: 1) acceptance of different approaches and viewpoints, 2) promoting risk-taking behavior, 3) top management's vision and commitment to providing a social response. (author)

  14. Development of criticality accident detector measuring neutrons and gamma-rays

    International Nuclear Information System (INIS)

    Tsujimura, Norio; Yoshida, Tadayoshi; Ishii, Masato

    2005-01-01

    The authors developed a new criticality accident detector measuring neutrons and gamma-rays. The detector is a cylindrical plastic scintillator coupled to a current-mode operated photomultiplier, and is covered by an inner cadmium shell, acting as a neutron to gamma-ray converter, and a 5cm thick outer polyethylene moderator in order to respond to the same threshold triggering dose regardless of whether it was exposed to neutrons, gamma-rays or a mixture of the two radiations. (author)

  15. Source term estimation during incident response to severe nuclear power plant accidents. Draft

    Energy Technology Data Exchange (ETDEWEB)

    McKenna, T J; Giitter, J

    1987-07-01

    The various methods of estimating radionuclide release to the environment (source terms) as a result of an accident at a nuclear power reactor are discussed. The major factors affecting potential radionuclide releases off site (source terms) as a result of nuclear power plant accidents are described. The quantification of these factors based on plant instrumentation also is discussed. A range of accident conditions from those within the design basis to the most severe accidents possible are included in the text. A method of gross estimation of accident source terms and their consequences off site is presented. The goal is to present a method of source term estimation that reflects the current understanding of source term behavior and that can be used during an event. (author)

  16. Source term estimation during incident response to severe nuclear power plant accidents. Draft

    International Nuclear Information System (INIS)

    McKenna, T.J.; Giitter, J.

    1987-01-01

    The various methods of estimating radionuclide release to the environment (source terms) as a result of an accident at a nuclear power reactor are discussed. The major factors affecting potential radionuclide releases off site (source terms) as a result of nuclear power plant accidents are described. The quantification of these factors based on plant instrumentation also is discussed. A range of accident conditions from those within the design basis to the most severe accidents possible are included in the text. A method of gross estimation of accident source terms and their consequences off site is presented. The goal is to present a method of source term estimation that reflects the current understanding of source term behavior and that can be used during an event. (author)

  17. Source term estimation during incident response to severe nuclear power plant accidents

    International Nuclear Information System (INIS)

    McKenna, T.J.; Glitter, J.G.

    1988-10-01

    This document presents a method of source term estimation that reflects the current understanding of source term behavior and that can be used during an event. The various methods of estimating radionuclide release to the environment (source terms) as a result of an accident at a nuclear power reactor are discussed. The major factors affecting potential radionuclide releases off site (source terms) as a result of nuclear power plant accidents are described. The quantification of these factors based on plant instrumentation also is discussed. A range of accident conditions from those within the design basis to the most severe accidents possible are included in the text. A method of gross estimation of accident source terms and their consequences off site is presented. 39 refs., 48 figs., 19 tabs

  18. Activities of JAERI's health physics department for the criticality accident of JCO

    International Nuclear Information System (INIS)

    Yamamoto, Katsumune; Kitano, Kyoshiro; Murakami, Hiroyuki; Yamaguchi, Takenori; Tsunoda, Masahiko

    2000-01-01

    This report describes early health physics activities from September 30 to October 1 taken by the authors' department after the JCO accident. They firstly knew the accident at around 12:20 (about 2 hr after the criticality). The activities involved the planning of schedule for ending the criticality; calculation of scheduled dose for the work to end it; dose measurement around JCO site; loaning out of devices for measuring neutron and of personal dose-meter; collection and radioactivity measurement of dust and soil, and of drinking water; and examination for contamination of people around the site, of their houses inside and of school gardens and equipments. The dose was scheduled to be firstly 20 mSv and then changed to 50 mSv due to the actual measurement at the accident site. The working time was to be 3 min at the site. The work was on either the dose or time. Radiation monitoring outside the JCO site revealed the presence of Na-24 and Cs-138: neutron dose was 10 times as high as γ-ray dose. The time course of dose rate change was found to be in parallel with the progress of works to end the criticality. (K.H.)

  19. Community emergency response to nuclear power plant accidents: A selected and partially annotated bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Youngen, G.

    1988-10-01

    The role of responding to emergencies at nuclear power plants is often considered the responsibility of the personnel onsite. This is true for most, if not all, of the incidents that may happen during the course of the plant`s operating lifetime. There is however, the possibility of a major accident occurring at anytime. Major nuclear accidents at Chernobyl and Three Mile Island have taught their respective countries and communities a significant lesson in local emergency preparedness and response. Through these accidents, the rest of the world can also learn a great deal about planning, preparing and responding to the emergencies unique to nuclear power. This bibliography contains books, journal articles, conference papers and government reports on emergency response to nuclear power plant accidents. It does not contain citations for ``onsite`` response or planning, nor does it cover the areas of radiation releases from transportation accidents. The compiler has attempted to bring together a sampling of the world`s collective written experience on dealing with nuclear reactor accidents on the sate, local and community levels. Since the accidents at Three Mile Island and Chernobyl, that written experience has grown enormously.

  20. Studies of potential severe accidents in Finnish nuclear power plants. Quarterly report 3. quarter 1987

    International Nuclear Information System (INIS)

    Aro, Ilari.

    1989-07-01

    This thesis is based on six publications dealing with severe accident studies in Finnish nuclear power plants. Main emphasis has been put on general technical bases and methodologies applied in severe accident evaluation in Finland. As an example of the use of the analysis and evaluation methods, the analysis of one representative accident sequence, t otal loss of AC power , has been presented for both Finnish power plant types. This accident sequence is required to be analyzed in the Finnish safety guide YVL 2.2 which deals with transient and accident analyses as a basis of technical solutions at nuclear powr plants. Two different analysis methods, MAAP 3.0 and MARCH 3/STCP have been used for receiving as complete a picture as possible of the flow of events and for verifying the models to some extent. Besides the use of the two different models, the method of sensitivity analysis has been used for evaluating the effects of some important technical parameters on the accident flow. Finally, conclusions of the applicability of the two methods for analyzing severe accident sequences in Finnish plants have been discussed

  1. The accident of the Three Mile Island nuclear power plant

    International Nuclear Information System (INIS)

    Llory, M.

    1999-01-01

    This book questions which statement can be made twenty years after the accident of the Three Mile Island reactor (USA) on the performances of complex reactor safety systems and on their evolutions and improvements. It questions also todays limits of reactors security and how such a reactor accident can be possible today. It presents also an analysis of the organizations which propose new perspectives in nuclear safety. (J.S.)

  2. Environmental impacts of an accident with a nuclear power plant

    International Nuclear Information System (INIS)

    1985-10-01

    A maximum credible reactor accident is considered: all safety systems fail and the reactor core is not cooled anymore. This so-called meltdown accident is discussed for two different weather situations. For these cases, the effects on public health and environment is studied (radioactive clouds, inhalation and deposition of radioactive materials). The radiation doses calculated are compared with standard levels. In so doing, an estimation is made of the measures necessary to reduce unfavourable consequences. (G.J.P.)

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

    Science.gov (United States)

    Mete, R; Sabatucci, A

    1989-09-30

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

  4. Criticality accidents in solution (CRAC and SILENE programmes) and complementary studies of accidents; radiation dosimetry in human organism during the CRAC programme

    International Nuclear Information System (INIS)

    Barbry, M.; Dousset, M.

    C.R.A.C. (CRiticality occurring ACcidentally) programme is intended to study experimentally the development of a criticality accident as it could occur when handling solutions of fissile material as well as the radiological consequences of such an accident. The fissile matter solutions have been chosen (a) for practical considerations of use and (b) because the probability of an accident occurring seems greater with this type of environment, as the known accidents have shown. The programme is twofold: study of accident physics: form of the evolution (peak, plateau, oscillations, boil up of solutions) the most probable maximum power, minimal power, flux and radiation spectra emitted, freed energy, associated effects, radiolysis, constraints, etc., study of radiological consequences: area dosimetry, individual dosimetry, radiobiological studies, etc. Additional criticality Accident experiments have been and continue to be made on the SILENE reactor in the following principal domains: determination of the emission rate of gaseous fission products and aerosols, area dosimetry and health dosimetry in the presence of shields around the core to vary the neutron and gamma components of the radiation field. Improvement in the knowledge of certain particular aspects of the power excursion, radiolysis gas and pressure wave, experiments of the ''boiling'' type [fr

  5. Source term analysis for a criticality accident in metal production line glove boxes

    International Nuclear Information System (INIS)

    Nguyen, D.H.

    1991-06-01

    A recent development in criticality accident analysis is the deterministic calculations of the transport of fission products and actinides through the barriers of the physical facility. The knowledge of the redistribution of the materials inside the facility will help determine the reentry and clean-up procedures. The amount of radioactive materials released to the environment is the source term for dispersion calculations. We have used an integrated computer model to determine the release of fission products to the environment from a hypothetical criticality event in a glove box of the metal production line (MPL) at the Lawrence Livermore National Laboratory (LLNL)

  6. Postulated accidents

    International Nuclear Information System (INIS)

    Ullrich, W.

    1980-01-01

    This lecture on 'Postulated Accidents' is the first of a series of lectures on the dynamic and transient behaviour of nuclear power plants, especially pressurized water reactors. The main points covered will be: Reactivity Accidents, Transients (Intact Loop) and Loss of Cooland Accidents (LOCA) including small leak. This lecture will discuss the accident analysis in general, the definition of the various operational phases, the accident classification, and, as an example, an accident sequence analysis on the basis of 'Postulated Accidents'. (orig./RW)

  7. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Executive summary: main report

    International Nuclear Information System (INIS)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks

  8. The estimation economic impacts from severe accidents of a nuclear power plant

    International Nuclear Information System (INIS)

    Jeong, J. T.; Jeong, W. D.

    2001-01-01

    The severe accidents of a nuclear power plant may cause health effects in the exposed population and societal economic impacts or costs. Techniques to assess the consequences of an accident in terms of cost may be applied in studies on the design of plant safety features and in examining countermeasure options as part of emergency planning or in decision making after an accident. In this study, the costs resulting from the severe accidents of a nuclear power plant were estimated for the different combinations of source term release parameters and meteorological data. Also, the costs were estimated for the different scenarios considering seasonal characteristics of Korea. The results can be used as essential inputs in costs/benefit analysis and in developing optimum risk reduction strategies

  9. Measuring Risk Aversion for Nuclear Power Plant Accident: Results of Contingent Valuation Survey in Korea

    International Nuclear Information System (INIS)

    Lee, Sang Hun; Kang, Hyun Gook

    2015-01-01

    Within the evaluation of the external cost of nuclear energy, the estimation of the external cost of nuclear power plant (NPP) severe accident is one of the major topics to be addressed. For the evaluation of the external cost of NPP severe accident, the effect of public risk averse behavior against the group accidents, such as NPP accident, dam failure, must be addressed. Although the equivalent fatalities from a single group accident are not common and its risk is very small compared to other accidents, people perceive the group accident more seriously. In other words, people are more concerned about low probability/high consequence events than about high probability/low consequence events having the same mean damage. One of the representative method to integrate the risk aversion in the external costs of severe nuclear reactor accidents was developed by Eeckoudt et al., and he used the risk aversion coefficient, mainly based on the analysis of financial risks in the stock markets to evaluate the external cost of nuclear severe accident. However, the use of financial risk aversion coefficient to nuclear severe accidents is not appropriate, because financial risk and nuclear severe accident risk are entirely different. In this paper, the individual-level survey was conducted to measure the risk aversion coefficient and estimate the multiplication factor to integrate the risk aversion in the external costs of NPP severe accident. This study propose an integrated framework on estimation of the external cost associated with severe accidents of NPP considering public risk aversion behavior. The theoretical framework to estimate the risk aversion coefficient/multiplication factor and to assess economic damages from a hypothetical NPP accident was constructed. Based on the theoretical framework, the risk aversion coefficient can be analyzed by conducting public survey with a carefully designed lottery questions. Compared to the previous studies on estimation of the

  10. Measuring Risk Aversion for Nuclear Power Plant Accident: Results of Contingent Valuation Survey in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Hun; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Within the evaluation of the external cost of nuclear energy, the estimation of the external cost of nuclear power plant (NPP) severe accident is one of the major topics to be addressed. For the evaluation of the external cost of NPP severe accident, the effect of public risk averse behavior against the group accidents, such as NPP accident, dam failure, must be addressed. Although the equivalent fatalities from a single group accident are not common and its risk is very small compared to other accidents, people perceive the group accident more seriously. In other words, people are more concerned about low probability/high consequence events than about high probability/low consequence events having the same mean damage. One of the representative method to integrate the risk aversion in the external costs of severe nuclear reactor accidents was developed by Eeckoudt et al., and he used the risk aversion coefficient, mainly based on the analysis of financial risks in the stock markets to evaluate the external cost of nuclear severe accident. However, the use of financial risk aversion coefficient to nuclear severe accidents is not appropriate, because financial risk and nuclear severe accident risk are entirely different. In this paper, the individual-level survey was conducted to measure the risk aversion coefficient and estimate the multiplication factor to integrate the risk aversion in the external costs of NPP severe accident. This study propose an integrated framework on estimation of the external cost associated with severe accidents of NPP considering public risk aversion behavior. The theoretical framework to estimate the risk aversion coefficient/multiplication factor and to assess economic damages from a hypothetical NPP accident was constructed. Based on the theoretical framework, the risk aversion coefficient can be analyzed by conducting public survey with a carefully designed lottery questions. Compared to the previous studies on estimation of the

  11. Airway accidents in critical care unit: A 3-year retrospective study in a Public Teaching Hospital of Eastern India

    Science.gov (United States)

    Dasgupta, Sugata; Singh, Shipti Shradha; Chaudhuri, Arunima; Bhattacharya, Dipasri; Choudhury, Sourav Das

    2016-01-01

    Background: Although tracheal tubes are essential devices to control and protect airway in a critical care unit (CCU), they are not free from complications. Aims: To document the incidence and nature of airway accidents in the CCU of a government teaching hospital in Eastern India. Methods: Retrospective analysis of all airway accidents in a 5-bedded (medical and surgical) CCU. The number, types, timing, and severity of airway accidents were analyzed. Results: The total accident rate was 19 in 233 intubated and/or tracheostomized patients over 1657 tube days (TDs) during 3 years. Fourteen occurred in 232 endotracheally intubated patients over 1075 endotracheal tube (ETT) days, and five occurred in 44 tracheostomized patients over 580 tracheostomy TDs. Fifteen accidents were due to blocked tubes. Rest four were unplanned extubations (UEs), all being accidental extubations. All blockages occurred during night shifts and all UEs during day shifts. Five accidents were mild, the rest moderate. No major accident led to cardiorespiratory arrest or death. All blockages occurred after 7th day of intubation. The outcome of accidents were more favorable in tracheostomy group compared to ETT group (P = 0.001). Conclusions: The prevalence of airway accidents was 8.2 accidents per 100 patients. Blockages were the most common accidents followed by UEs. Ten out of the 15 blockages and all 4 UEs were in endotracheally intubated patients. Tracheostomized patients had 5 blockages and no UEs. PMID:27076709

  12. PCTRAN-3: The third generation of personal computer-based plant analyzer for severe accident management

    International Nuclear Information System (INIS)

    Li-Chi Cliff Po; Link, John M.

    2004-01-01

    PCTRAN is a plant analyzer that uses a personal computer to simulate plant response. The plant model is recently expanded to accommodate beyond design-basis severe accidents. In the event of multiple failures of the plant safety systems, the core may experience heatup and extensive failure. Using a high-powered personal computer (PC), PCTRAN-3 is designed to operate at a speed significantly faster than real-time. A convenient, interactive and user-friendly graphics interface allows full control by the operator. The plant analyzer is intended for use in severe accident management. In this paper the code's component models and sample runs ranging from normal operational transients to severe accidents are reviewed. (author)

  13. A cost effective approach for criticality accident analysis of a DOE SNF storage facility

    International Nuclear Information System (INIS)

    Garrett, R.L.; Couture, G.F.; Gough, S.T.

    1997-01-01

    This paper presents the methodologies used to derive criticality accident analyses for a spent nuclear fuel receipt, storage, handling, and shipping facility. Two criticality events are considered: process-induced and Natural Phenomena Hazards (NPH)-induced. The criticality analyses required the development of: (1) the frequency at which each sceanario occurred, (2) the estimated number of fissions for each scenario, and (3) the consequences associated with each criticality scenario. A fault tree analysis was performed to quantify the frequency of criticality due to process-induced events. For the frequency analysis of NPH-induced criticality, a probabilistic approach was employed. To estimate the consequences of a criticality event, the resulting fission yield was determined using a probabilistic approach. For estimating the source term, a 95% amount of overall conservatism was targeted. This methodology applied to the facility criticality scenarios indicated that: (1) the 95th percentile yield levels from the historical yield distributions are approximately 5 x 10 17 fissions and 5 x 10 18 fissions for internal event and NPH-induced criticality event, respectively; and (2) using probabilistic Latin Hypercube Sampling, the downwind 95th percentile dose to a receptor at the US DOE reservation boundary is 2.2 mrem. This estimate is compared to the bounding dose of 1.4 rem. 4 refs., 1 fig

  14. Possible emission of radioactive fission products during off-design accidents at a nuclear power plant with VVER-1000 reactor

    International Nuclear Information System (INIS)

    Dubkov, A.P.; Kozlov, V.F.; Luzanova, L.M.

    1995-01-01

    It is well known that eight nuclear power plants with VVER-1000 reactors have been constructed in Russia, Ukraine, and in the Republic of Belarus and they have been operating successfully without any serious accidents since 1980. These facilities have been analyzed for various accident scenarios, and measures have been incorporated which will prevent core damage during these possible events. However, an off-design accident can occur, and in such a case, the radiological consequences would exceed the worst design accidents. This paper reviews a number of potential off-design accidents in order to develop an accident plan to mitigate the consequences of such an accident

  15. Recent Perspective on the Severe Accident Management Programme for Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kim, Manwoong; Lee, Sukho; Lee, Jungjae; Chung, Kuyoung

    2017-01-01

    Severe Accident Management Guidelines (SAMGs), has been developed to help operators to prevent or mitigate the impacts of accidents at nuclear power plants. Severe accident management was first introduced in the 1990s with the creation of SAMGs following recognition that post-Three Mile Island Emergency Operating Procedures (EOPs) did not adequately address severe core damage conditions. Establishing and maintaining multiple layers of defence against any internal/external hazards is an important measure to reduce radiological risks to the public and environment. This study is intended to suggest future regulatory perspectives to strengthen the prevention and mitigation strategies for severe accidents by review of the current status of revision of IAEA Safety Standard on Severe Accident Management Programmes for Nuclear Power Plants and the combined PWR SAMG. This new IAEA Safety Guide will address guidelines for preparation, development, implementation and review of severe accident management programs during all operating conditions for both reactor and spent fuel pool. This Guide is used by operating organizations of nuclear power plants and their support organizations. It may also be used by national regulatory bodies and technical support organizations as a reference for developing their relevant safety requirements and for conducting reviews and safety assessments for SAMP including SAMG. The Pressurized Water Reactor Owner’s Group (PWROG) is upgrading the original generic Severe Accident Management Guidelines (SAMGs) into single Severe Accident Guidelines (SAGs) for the PWR SAMG aims to consolidate the advantages of each of the separate vendor severe accident (SA) mitigation methods. This new PWROG SAGs changes the SAMG process to be made that can improve SA response. Changes have been made that guidance is available for control room operators when the TSC is not activated thus allowing for timely accident response. Other changes were made to the guidance

  16. Severe Accidents: French Regulatory Practice for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Colin, M.

    1997-01-01

    In the framework of a continuous and iterative process, the French Safety Authority asks the utility EDF to implement equipment and procedure modifications on the operating reactors, in order to cope with the most likely Severe Accident sequences. As a result of Probabilistic Safety Assessments published in 1990, important equipment and procedure modifications are being implemented on the French PWRs to improve the safety in shutdown states. The implementation of another set of modifications against some reactivity accident sequences is also in progress. More recently, the Safety Authority expressed specific Severe Accident requirements in terms of instrumentation, equipment qualification, high pressure core melt accidents and hydrogen risk prevention. In that respect, EDF was asked to implement hydrogen recombiners on its reactors. On the other hand, the French Safety authority is involved with its German counterpart in the assessment process of the European Pressurized Water Reactor Project. In consistency with the common recommendations of the Safety Authorities involved, Severe Accident provisions for this reactor are being taken into account at the design stage

  17. Community emergency response to nuclear power plant accidents: A selected and partially annotated bibliography

    International Nuclear Information System (INIS)

    Youngen, G.

    1988-10-01

    The role of responding to emergencies at nuclear power plants is often considered the responsibility of the personnel onsite. This is true for most, if not all, of the incidents that may happen during the course of the plant's operating lifetime. There is however, the possibility of a major accident occurring at anytime. Major nuclear accidents at Chernobyl and Three Mile Island have taught their respective countries and communities a significant lesson in local emergency preparedness and response. Through these accidents, the rest of the world can also learn a great deal about planning, preparing and responding to the emergencies unique to nuclear power. This bibliography contains books, journal articles, conference papers and government reports on emergency response to nuclear power plant accidents. It does not contain citations for ''onsite'' response or planning, nor does it cover the areas of radiation releases from transportation accidents. The compiler has attempted to bring together a sampling of the world's collective written experience on dealing with nuclear reactor accidents on the sate, local and community levels. Since the accidents at Three Mile Island and Chernobyl, that written experience has grown enormously

  18. National radiological emergency response to the Fukushima Daiichi Nuclear Power Plant accident

    International Nuclear Information System (INIS)

    Dela Rosa, Alumanda M.

    2011-01-01

    The Fukushima nuclear power plant accident occurred on March 11, 2011, when two natural disasters of unprecedented strengths, an earthquake with magnitude 9 followed one hour later by a powerful tsunami struck northeastern Japan and felled the external power supply and the emergency diesel generators of the Fukushima Daiichi nuclear power station, resulting in a loss of coolant accident. There were core meltdowns in three nuclear reactors with the release of radioactivity estimated to be 1/10 of what was released to the environment during the Chernobyl nuclear power plant accident in April 1986. The Fukushima nuclear accident tested the capability of the Philippine Nuclear Research Institute (PNRI) and the National Disaster Risk Reduction and Management Council (NDRRMC) in responding to such radiological emergency as a nuclear power plant accident. The PNRI and NDRRMC activated the RADPLAN for possible radiological emergency. The emergency response was calibrated to the status of the nuclear reactors on site and the environmental monitoring undertaken around the site and off-site, including the marine environment. This orchestrated effort enabled the PNRI and the national agencies concerned to reassure the public that the nuclear accident does not have a significant impact on the Philippines, both on the health and safety of the people and on the safety of the environment. National actions taken during the accident will be presented. The role played by the International Atomic Energy Agency as the central UN agency for nuclear matters will be discussed. (author)

  19. Knowledge data base for severe accident management of nuclear power plants

    International Nuclear Information System (INIS)

    Ogino, Masao; Kawabe, Ryuhei; Nagasaka, Hideo; Sumida, Susumu; Fukasawa, Masanori; Muta, Hitoshi

    2011-01-01

    For the reinforcement of the safety of NPPs, the continuous efforts are very important to take in the up-to-date scientific and technical knowledge positively and to reflect them into the safety regulation. The purpose of this present study is to gather effectively the scientific and technical knowledge about the severe accident (SA) phenomena and the accident management (AM) for prevention and mitigation of severe accident, and to take in the experimental data by participating in the international cooperative experiments regarding the important SA phenomena and the effectiveness of accident management. Based on those data and knowledge, JNES is developing and improving severe accident analysis models to maintain the severe accident analysis codes and the accident management knowledge base for assessment of the NPPs in Japan. The activities in fiscal year 2010 are as follows; Experimental study on OECD/NEA projects such as MCCI, SERENA, SFP and international cooperative PSI-ARTIST project, and analytical study on accident management review of new plant and making regulation for severe accident. (author)

  20. Concept and objectives of accident management in LWR type plants

    International Nuclear Information System (INIS)

    Herttrich, P.M.; Hicken, E.F.

    1990-01-01

    For the sake of putting the previous protection and prevention concept in its proper place, it is shown, first of all, on which basis the prevention against damages required according to the state of the art in science and technology was proved under the licensing practice applied so far. Secondly, the previous practice of dynamic upgrading of safety engineering and risk prevention is explained. The introduction of accident management measures is a consequent continuation of this practice. Concrete approaches and objectives of accident management are outlined; an overview of scientific and technical foundations for the development, assessment and introduction of accident management measures is given, and finally the most important organizational and procedural aspects are dealt with. (orig./DG) [de

  1. Mental health problems after the 2011 Fukushima Dai-ichi nuclear power plant accident

    International Nuclear Information System (INIS)

    Niwa, Shin-ichi

    2012-01-01

    The name of Fukushima has now become well-known worldwide after Hiroshima and Nagasaki as the third place exposed to radiation in Japan. This radiation pollution has severely damaged the chief industries of Fukushima Prefecture, namely agriculture, fishery, and tourist industry. It has also stimulated strong anxious feelings among parents with young children. The accident has caused a critical situation in the psychiatric and mental health services in Fukushima as well. Five hospitals with psychiatric beds within 30 km from the Fukushima Dai-ichi Nuclear Power Plant were ordered to transfer their inpatients to other hospitals outside the designated 30 km-areas and to close down the hospitals immediately after the nuclear plant accident. In total, more than 800 psychiatric beds disappeared in an instant, and 1,228 persons including psychiatric inpatients and residents of elderly people nursing homes were transferred to other facilities far away. Rational explanation that low-level radiation in Fukushima will not do harm to people did not necessarily relieve existing anxiety among people. The terms 'safety' and 'relief' are usually used in combination; however, 'relief' was separated from 'safety' this time in Fukushima. People gradually began to feel 'relieved', when they themselves got involved in the cleaning work of radiation although its effect remained ambiguous. Now we have the following mental health problems after the 2011 Fukushima Dai-ichi nuclear power plant accident; recovery and maintenance of treatment systems for psychiatric patients in the affected areas, efforts for early detection and intervention of depression, severe stress disorder, adaptation disorder, and alcohol abuse which are expected to occur due to the earthquake and radiation pollution, prevention of suicides, relief from anxiety resulting from radiation pollution, adequate treatment of mental problems among children with long-term evacuation, prevention of fall in physical and mental

  2. Estimation of dose distribution and neutron spectra in JCO critical accident by shielding calculations

    International Nuclear Information System (INIS)

    Sakamoto, Yukio

    2001-01-01

    The information about neutrons at the surrounding of JCO site in the critical accident is limited to survey results by neutron Rem counter in the period of accident and activation data very near the test facility measured after the shut down of accident. This caused the big uncertainty in the dose estimation by detailed shielding calculation codes. On the other hand, environmental activity data measured by radiochemical researchers included the information about fast neutrons inside of JCO site and thermal neutrons up to 1 km from test facility. It is important to grasp the actual circumstance and examine the executed evaluation of the critical accident as scientifically as possible. Therefore, it is meaningful for different field researchers to corporate and exchange the information. In the Technical Divisions of Radiation Science and Technology in Atomic Energy Society of Japan, the information about neutron spectra are released from their home page and three groups of JAERI/CRC, Sumitomo Atomic Energy Industry and Nuclear Power Engineering Corp. (NUPEC)/Mitsubishi Research Institute Inc. (MRI), tried the shielding calculation by Monte Carlo Code MCNP-4B. The procedures and main results of shielding calculations were reviewed in this report. The main difference of shielding calculation by three groups was density and water content of autoclaved light-weight concrete (ALC) as the wall and ceiling. From the result by NUPEC/MRI, it was estimated that the water content in ALC was from 0.05 g/cm 3 to 0.10 g/cm 3 . The behavior of dose equivalent attenuation obtained by shielding calculation was very similar with the measured data from 250 m to 1,700 m obtained by survey meter, TLD and monitoring post. For more exact dose estimation, more detail examination of density and water content of ALC will be needed. (author)

  3. Strategies to cope with severe accidents at nuclear power plants

    International Nuclear Information System (INIS)

    Kovacs, Zoltan; Rydzi, Stanislav

    2015-01-01

    The paper focusses, in particular, on SAMG – Severe Accident Management Guidelines, and on SBEOP - Symptom Based Emergency Operating Procedures. It is shown how the concepts are applicable, how they are applied in practice, and in which aspects they need improvements. (orig.)

  4. Pending issues for severe accident management in Wolsong plants

    International Nuclear Information System (INIS)

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

    2015-01-01

    While the fraction of electric power supplied from a PHWR is more than 10% in Korea, the establishment of PHWR safety enhancement based on the SAM (Severe Accident Management) technology is still weak. The final approval on the extended operation and a stress test of Wolsong-1 were made under the condition that SAM is to be enhanced. Under this situation, the current research at KAERI of Korea has a vision to strengthen the unique value of a PHWR by resolving the pending SAM issues devaluating the PHWRs’ original value. Research activities in this area will be presented. This presentation will include: The operating strategy of CFVS (Containment Filtered Vent System) for Wolsong in which vent size and closure pressure are treated because some peak spikes (at failure times of calandria and calandria vault) are difficult to be controlled; Reactor Building failure pressure at which failure probability is treated for different modes such as global and leak failures; the adequacy of DCRV (Degasser Condenser tank Relief Valve) steam relief capacity with severe SGTR source term, and Hydrogen generation and control issue which is specific to CANDU. Furthermore, current SAM guidance has a lack of information on accident diagnostic and prognostic analyses, which is difficult for the TSC (Technical Service Center) emergency staff members to deal with under real accident conditions. Thus, prototypic technologies (such as an accident inferring engine and simulator) together with SAM updates are being developed as key elements to SAM supporting tools called SAMEX-CANDU

  5. A microcomputer-based model for identifying urban and suburban roadways with critical large truck accident rates

    International Nuclear Information System (INIS)

    Brogan, J.D.; Cashwell, J.W.

    1992-01-01

    This paper presents an overview of techniques for merging highway accident record and roadway inventory files and employing the combined data set to identify spots or sections on highway facilities in urban and suburban areas with unusually high large truck accident rates. A statistical technique, the rate/quality control method, is used to calculate a critical rate for each location of interest. This critical rate may then be compared to the location's actual accident rate to identify locations for further study. Model enhancements and modifications are described to enable the technique to be employed in the evaluation of routing alternatives for the transport of radioactive material

  6. Accident precursors, near misses, and warning signs: Critical review and formal definitions within the framework of Discrete Event Systems

    International Nuclear Information System (INIS)

    Saleh, Joseph H.; Saltmarsh, Elizabeth A.; Favarò, Francesca M.; Brevault, Loïc

    2013-01-01

    An important consideration in safety analysis and accident prevention is the identification of and response to accident precursors. These off-nominal events are opportunities to recognize potential accident pathogens, identify overlooked accident sequences, and make technical and organizational decisions to address them before further escalation can occur. When handled properly, the identification of precursors provides an opportunity to interrupt an accident sequence from unfolding; when ignored or missed, precursors may only provide tragic proof after the fact that an accident was preventable. In this work, we first provide a critical review of the concept of precursor, and we highlight important features that ought to be distinguished whenever accident precursors are discussed. We address for example the notion of ex-ante and ex-post precursors, identified for postulated and instantiated (occurred) accident sequences respectively, and we discuss the feature of transferability of precursors. We then develop a formal (mathematical) definition of accident precursors as truncated accident sequences within the modeling framework of Discrete Event Systems. Additionally, we examine the related notions of “accident pathogens” as static or lurking adverse conditions that can contribute to or aggravate an accident, as well as “near misses”, “warning signs” and the novel concept of “accident pathway”. While these terms are within the same linguistic neighborhood as “accident precursors”, we argue that there are subtle but important differences between them and recommend that they not be used interchangeably for the sake of accuracy and clarity of communication within the risk and safety community. We also propose venues for developing quantitative importance measures for accident precursors, similar to component importance measures in reliability engineering. Our objective is to establish a common understanding and clear delineation of these terms, and

  7. A defense in depth approach for nuclear power plant accident management

    Energy Technology Data Exchange (ETDEWEB)

    Chih-Yao Hsieh; Hwai-Pwu Chou [Institute of Nuclear Engineering and Science, National Tsing Hua University, Hsinchu, TW (China)

    2015-07-01

    An initiating event may lead to a severe accident if the plant safety functions have been challenged or operators do not follow the appropriate accident management procedures. Beyond design basis accidents are those corresponding to events of very low occurrence probability but such an accident may lead to significant consequences. The defense in depth approach is important to assure nuclear safety even in a severe accident. Plant Damage States (PDS) can be defined by the combination of the possible values for each of the PDS parameters which are showed on the nuclear power plant simulator. PDS is used to identify what the initiating event is, and can also give the information of safety system's status whether they are bypassed, inoperable or not. Initiating event and safety system's status are used in the construction of Containment Event Tree (CET) to determine containment failure modes by using probabilistic risk assessment (PRA) technique. Different initiating events will correspond to different CETs. With these CETs, the core melt frequency of an initiating event can be found. The use of Plant Damage States (PDS) is a symptom-oriented approach. On the other hand, the use of Containment Event Tree (CET) is an event-oriented approach. In this study, the Taiwan's fourth nuclear power plants, the Lungmen nuclear power station (LNPS), which is an advanced boiling water reactor (ABWR) with fully digitized instrumentation and control (I and C) system is chosen as the target plant. The LNPS full scope engineering simulator is used to generate the testing data for method development. The following common initiating events are considered in this study: loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), loss of offsite power (LOOP), station blackout (SBO). Studies have indicated that the combination of the symptom-oriented approach and the event-oriented approach can be helpful to find mitigation strategies and is useful for the accident

  8. Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Kuruc, J.; Matel, L.

    2007-01-01

    The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)

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

    Directory of Open Access Journals (Sweden)

    G. V. Arkhangelskaya

    2012-01-01

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

  10. Prediction of accident sequence probabilities in a nuclear power plant due to earthquake events

    International Nuclear Information System (INIS)

    Hudson, J.M.; Collins, J.D.

    1980-01-01

    This paper presents a methodology to predict accident probabilities in nuclear power plants subject to earthquakes. The resulting computer program accesses response data to compute component failure probabilities using fragility functions. Using logical failure definitions for systems, and the calculated component failure probabilities, initiating event and safety system failure probabilities are synthesized. The incorporation of accident sequence expressions allows the calculation of terminal event probabilities. Accident sequences, with their occurrence probabilities, are finally coupled to a specific release category. A unique aspect of the methodology is an analytical procedure for calculating top event probabilities based on the correlated failure of primary events

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  12. In-plant considerations for optimal offsite response to reactor accidents

    International Nuclear Information System (INIS)

    Burke, R.P.; Heising, C.D.; Aldrich, D.C.

    1982-11-01

    Offsite response decision-making methods based on in-plant conditions are developed for use during severe reactor-accident situations. Dose projections are used to eliminate all LWR plant systems except the reactor core and the spent-fuel storage pool from consideration for immediate offsite emergency response during accident situations. A simple plant information-management scheme is developed for use in offsite response decision-making. Detailed consequence calculations performed with the CRAC2 model are used to determine the appropriate timing of offsite-response implementation for a range of PWR accidents involving the reactor core. In-plant decision criteria for offsite-response implementation are defined. The definition of decision criteria is based on consideration of core-accident physical processes, in-plant accident monitoring information, and results of consequence calculations performed to determine the effectiveness of various public-protective measures. The benefits and negative aspects of the proposed response-implementation criteria are detailed

  13. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices VII, VIII, IX, and X. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy.

  14. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendices VII, VIII, IX, and X

    International Nuclear Information System (INIS)

    1975-10-01

    Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy

  15. Analysis of accidents and troubles of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Kobayashi, Kunio

    1980-01-01

    In Japan, electric power companies are obliged to report the accidents and troubles occurred in nuclear power stations to the MITI according to the relevant laws, and 166 cases in total have been reported as of the end of March, 1980. These accidents and troubles are all trivial, and do not cause problems from the viewpoint of the safety nuclear power stations. Regarding respective accidents and troubles, the causes have been sought thoroughly, and the sufficient countermeasures have been taken on all occasions. But in order to improve the reliability of nuclear power stations further, it is important to treat the accidents and troubles occurred so far statistically and grasp the general trend. Thereupon, 152 accidents and troubles occurred till September, 1979, were analyzed quantitatively, and the results are reported in this paper. From the results, the prospect hereafter is discussed. The number of the reported cases of accidents and troubles in each nuclear power plant in operation every year is tabulated. The accidents and troubles were relatively frequent in the initial two or three years of operation of respective new reactor types, but decreased thereafter. The systems to which troubled equipments belong and the troubled equipments are shown. Most troubles have occurred in reactor cooling systems and valves. The situations and causes of troubles, the operational conditions at the time of the accidents and troubles and the effects and others are reported. (Kako, I.)

  16. Residents call for greater openness, accountability and involvement: Lessons learned from the JCO criticality accident

    International Nuclear Information System (INIS)

    Taniguchi, Taketoshi; Tsuchiya, Tomoko; Kosugi, Motoko

    2000-01-01

    This paper discusses the JCO (Japan Nuclear Fuel Conversion Co.) criticality accident from social viewpoints based on the detailed examination of the survey data and experience of participation into Tokai village office's surveys. We focus the mechanisms of amplifying anxieties of the local residents and clarify the key factors affected in the social amplification process. And we discuss the importance of communicating and deliberating among the lay people, public officials and professionals about health, safety and environmental risks associated with nuclear energy, referring to the public opinions about what kinds of information and actions are needed. (J.P.N.)

  17. Environment radiological monitoring by JNC related to the JCO criticality accident

    International Nuclear Information System (INIS)

    Watanabe, Hitoshi

    2001-01-01

    Concerning the criticality accident at JCO Co., Ltd. (JCO) which occurred at 10:35 on 30th Sep. 1999, Japan Nuclear Cycle Development Institute (JNC) established ''JNC's task force'' at 12:35 on the same data in conjunction with Head Office and Tokai Works. JNC's task force had collaborated on environmental radiological monitoring with the government of Japan and the local governments. This report compiles the results of the environmental monitoring performed by JNC's task force based on the request from the government of Japan and the local governments. (author)

  18. Study of a criticality accident involving fuel rods and water outside a power reactor

    International Nuclear Information System (INIS)

    Beloeil, L.

    2000-01-01

    It is possible to imagine highly unlikely but numerous accidental situations where fuel rods come into contact with water under conditions close to atmospheric values. This work is devoted to modelling and simulation of first instants of the power excursion that may result from such configurations. We show that void effect is a preponderant feedback for most severe accidents. The formation of a vapour film around the rods is put forward and confirmed with the help of experimental transients using electrical heating. We propose then a vapour/liquid flow model able to reproduce void fraction evolution. The vapour film is treated as a compressible medium. Conservation balance equations are solved on a moving mesh with a two-dimensional scheme and boundary conditions taking notice of interfacial phenomena and axial escape possibility. Movements of the liquid phase are modelled through a non-stationary integral equation and a dissipative term suited to the particular geometry of this flow. The penetration of energy into the liquid is also calculated. Thus, the coupling of aerodynamic and hydrodynamic modules gives results in excellent agreement with experiments. Next, neutronic phenomena into the fuel pellet, their feedback effects and the distribution of power through the rod are numerically translated. For each developed module, validation tests are provided. Then, it is possible to simulate the first seconds of the whole criticality accident. Even if this calculation tool is only a way of study as a first approach, performed simulations are proving coherent with reported data on recorded accidents. (author)

  19. Multiple parameter biodosimetry of exposed workers from the JCO criticality accident in Tokai-mura

    Energy Technology Data Exchange (ETDEWEB)

    Blakely, William F. [Armed Forces Radiobiology Research Institute, Bethesda, MD (United States)

    2002-03-01

    molecular biomarkers for neutron exposures in order to overcome this limitation. Lymphocyte counts were used to estimate radiation doses for the same three severely exposed patients. Haematological responses are an early response biomarker for radiation dose assessment. Goans and colleagues earlier reported lymphocyte depletion kinetic models for dose estimates based on human radiation accident registry data for whole-body acute gamma exposures and more recently for criticality accidents. Their data indicate a neutron RBE for lymphocyte depletion kinetics close to unity. Measurement of lymphocyte depletion kinetics is useful for initial dose estimation for radiation accidents. This biodosimetric approach alone, however, does not gauge the greater effectiveness of neutron radiation in criticality accidents involving mixed neutron and gamma radiation. Sodium-24 counts, with both blood and whole-body (Worker C) counts based on neutron activation of stable {sup 23}Na, were also used to obtain early estimates of absorbed doses for the Tokai-mura victims. The measurement of {sup 32}P in blood and urine samples is a potentially useful alternative for estimation of severe neutron radiation exposures because, during interaction with biological materials, fast neutrons convert {sup 32}S to {sup 32}P and thermal neutrons convert {sup 31}P to {sup 32}P. Nishimura and colleagues report, in the accompanying article in this issue, blood and urine {sup 32}P values obtained from the three severely exposed patients in the JCO criticality accident. These data, combined with dose estimates derived from other approaches, can contribute to the establishment of an in vivo human calibration curve for neutron dose assessment based on urine {sup 32}P kinetics measurements. The current state of the art for dose assessment following radiation accidents involves use of multiple parameter biological dosimetry. Several of the radiation bioassays (i.e., chromosome aberrations, lymphocyte counts, prodromal

  20. A structured approach to individual plant evaluation and accident management

    International Nuclear Information System (INIS)

    Klopp, G.T.

    1992-01-01

    The need for long term development of accident management programs is acknowledged and the key tool for that development is identified as the IPE Program. The Edison commitment to build an integrated program is cited and the effect on the IPE effort is considered. Edison's integrated program is discussed in detail. The key benefits, realism and long term savings, are discussed. Some of the highly visible products such as neural network artificial intelligence systems are cited

  1. Accident prevention ordinance 2.0 Thermal Power Plants

    International Nuclear Information System (INIS)

    Egyptien, H.H.; Fischermann, E.

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

  2. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    International Nuclear Information System (INIS)

    Park, Hae Kyun; Kim, Sang Nyung

    2013-01-01

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was conservatively calculated

  3. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hae Kyun; Kim, Sang Nyung [Kyung Hee Univ., Yongin (Korea, Republic of)

    2013-10-15

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was

  4. Local governments' roles of the compensation for damage by the Tokai JCO criticality accident

    Energy Technology Data Exchange (ETDEWEB)

    Tanabe, Tomoyuki [Central Research Inst. of Electric Power Industry, Tokyo (Japan). Socio-Economic Research Center

    2003-03-01

    The Tokai JCO criticality accident on September 30, 1999 was the first case to which The Law on Compensation for Nuclear Damage was applied. Although the Law on Compensation for Nuclear Damage formulates the outline of the institutional framework for nuclear third party liability together with operator's insurance scheme, details of actual compensation procedure are not specified. By this reason, the compensation procedure in the Tokai accident had been executed without a concrete legal specification and a precedent. In spite of this situation, the compensation procedure with the accident led to an unexpectedly successful result. We observe the several reasons why the compensation procedure was implemented successfully despite the lack of concrete legal specification and a precedent. One of the reasons is that the local governments, Tokai Village and Ibaraki Prefecture, immediately took the leadership in implementing a temporary regime of compensation procedure without wasting time for waiting national government's directives. Upon practicing this compensation procedure, the local governments implemented the following steps. (1) Initial estimation of the amount and scope of damage. (2) Providing the criteria and heads of damage subject to compensation. (3) Unitary compensation procedure at the local levels. (4) Distribution of emergency payments for the victims. (5) Facilitating compensatory negotiation between the victims and JCO as arbitrator. However, some concerns are also pointed out about the fact that the local government directed the whole procedure without sufficient adjustment with the national government for compensation policy. Among all, in the compensation led by the local governments, it was difficult to guarantee fairness of compensation because victims who are influential on the local government such as industrial associations would have unfairly strong negotiation power in the compensatory negotiation, while the operator being

  5. Local governments' roles of the compensation for damage by the Tokai JCO criticality accident

    International Nuclear Information System (INIS)

    Tanabe, Tomoyuki

    2003-01-01

    The Tokai JCO criticality accident on September 30, 1999 was the first case to which The Law on Compensation for Nuclear Damage was applied. Although the Law on Compensation for Nuclear Damage formulates the outline of the institutional framework for nuclear third party liability together with operator's insurance scheme, details of actual compensation procedure are not specified. By this reason, the compensation procedure in the Tokai accident had been executed without a concrete legal specification and a precedent. In spite of this situation, the compensation procedure with the accident led to an unexpectedly successful result. We observe the several reasons why the compensation procedure was implemented successfully despite the lack of concrete legal specification and a precedent. One of the reasons is that the local governments, Tokai Village and Ibaraki Prefecture, immediately took the leadership in implementing a temporary regime of compensation procedure without wasting time for waiting national government's directives. Upon practicing this compensation procedure, the local governments implemented the following steps. (1) Initial estimation of the amount and scope of damage. (2) Providing the criteria and heads of damage subject to compensation. (3) Unitary compensation procedure at the local levels. (4) Distribution of emergency payments for the victims. (5) Facilitating compensatory negotiation between the victims and JCO as arbitrator. However, some concerns are also pointed out about the fact that the local government directed the whole procedure without sufficient adjustment with the national government for compensation policy. Among all, in the compensation led by the local governments, it was difficult to guarantee fairness of compensation because victims who are influential on the local government such as industrial associations would have unfairly strong negotiation power in the compensatory negotiation, while the operator being responsible for the

  6. Including severe accidents in the design basis of nuclear power plants: An organizational factors perspective after the Fukushima accident

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.; Frutuoso e Melo, P.F.

    2015-01-01

    Highlights: • The Fukushima accident was man-made and not caused by natural phenomena. • Vulnerabilities were known by regulator and licensee but measures were not taken. • There was lack of independence and transparency of the regulatory body. • Laws and regulations have not been updated to international standards. • Organizational failures have played an important role in the Fukushima accident. - Abstract: The Fukushima accident was clearly an accident made by humans and not caused by natural phenomena as was initially thought. Vulnerabilities were known by both regulators and operator but they postponed measures. The emergency plan was not effective in protecting the public, because the involved parties were not sufficiently prepared to make the right decisions. The shortcomings and faults mentioned above resulted from the lack of independence and transparency of the regulatory body. Even laws and regulations, and technical standards, have not been upgraded to international standards. Regulators have not defined requirements and left for the operator to decide what would be more appropriate. In this aspect, there was clearly a lack of independence between these bodies and operator’s lobby power. The above situation raised the question of urgent updating of institutions, in particular those responsible for nuclear safety. The above evidences show that several nuclear safety principles were not followed. This paper intends to highlight some existing safety criteria that were developed from the operational experience of the severe accidents that occurred at TMI and Chernobyl that should be incorporated in the design of new nuclear power plants and to provide appropriate design changes (backfittings) for reactors that belong to the previous generation prior to the occurrence of these accidents, through the study of design vulnerabilities. Furthermore, the main criteria that define an effective regulatory agency are also discussed. Although these

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

    International Nuclear Information System (INIS)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor

    2013-01-01

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

  8. Criticality studies: One of the two pillars of criticality safety at the Belgonucleaire MOX plant

    International Nuclear Information System (INIS)

    Lance, B.; Maldague, T.; Evrard, G.; Renard, A.; Kockerols, P.

    2001-01-01

    The present paper focuses on the criticality studies performed by the Engineering Division of Belgonucleaire. These are one of the two pillars of the criticality prevention implemented for the Belgonucleaire MOX producing plant. (author)

  9. External Cost Assessment of Nuclear Power Plant Accident considering Public Risk Aversion Behavior: the Korean Case

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Hun; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    The conventional approach for monetary valuation of NPP accident consequence consists of calculating the expected value of various accident scenarios. However, the main criticism of the conventional approach is that there is a discrepancy between the social acceptability of the risk and the estimated expected value of NPP accident. Therefore, an integrated framework for the estimation of the external cost associated with an NPP accident considering the public risk aversion behavior was proposed in this study based on the constructed theoretical framework for estimating both the value of statistical life (VSL) and the risk aversion coefficient associated with an NPP accident to take account of the accident cost into the unit electricity generation cost of NPP. To estimate both parameters, an individual-level survey was conducted on a sample of 1,364 participants in Korea. Based on the collected survey responses, both parameters were estimated based on the proposed framework and the external cost of NPP accident was estimated based on the consequence analysis and considering the direct cost factors for NPP accident. Internalization of external costs into the comprehensive energy production cost has been considered as a potentially efficient policy instrument for a more sustainable energy supply and use. However, the internalization of externalities, such as public health damage, have raised a number of generic policy issues in a nuclear energy sector, with specific challenges resulting from the distinct characteristics of external cost estimation. Especially, the major challenge remained to address the public safety concerns regarding a nuclear accident, which can be specified as low-probability high-consequence accident, driven by the aspects of public risk aversion.

  10. External Cost Assessment of Nuclear Power Plant Accident considering Public Risk Aversion Behavior: the Korean Case

    International Nuclear Information System (INIS)

    Lee, Sang Hun; Kang, Hyun Gook

    2016-01-01

    The conventional approach for monetary valuation of NPP accident consequence consists of calculating the expected value of various accident scenarios. However, the main criticism of the conventional approach is that there is a discrepancy between the social acceptability of the risk and the estimated expected value of NPP accident. Therefore, an integrated framework for the estimation of the external cost associated with an NPP accident considering the public risk aversion behavior was proposed in this study based on the constructed theoretical framework for estimating both the value of statistical life (VSL) and the risk aversion coefficient associated with an NPP accident to take account of the accident cost into the unit electricity generation cost of NPP. To estimate both parameters, an individual-level survey was conducted on a sample of 1,364 participants in Korea. Based on the collected survey responses, both parameters were estimated based on the proposed framework and the external cost of NPP accident was estimated based on the consequence analysis and considering the direct cost factors for NPP accident. Internalization of external costs into the comprehensive energy production cost has been considered as a potentially efficient policy instrument for a more sustainable energy supply and use. However, the internalization of externalities, such as public health damage, have raised a number of generic policy issues in a nuclear energy sector, with specific challenges resulting from the distinct characteristics of external cost estimation. Especially, the major challenge remained to address the public safety concerns regarding a nuclear accident, which can be specified as low-probability high-consequence accident, driven by the aspects of public risk aversion

  11. The Chernobyl accident and the Spanish nuclear power plants. Technical report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-11-15

    On the morning of April 26, 1986, Unit 4 of the Chernobyl Nuclear Power Plant (Ukraine, USSR) suffered an accident of the greatest magnitude among those which have taken place in nuclear energy installations employed for peaceful uses. The accident reached a degree of severity unknown up to now in nuclear energy generating plants, both with respect to the loss of human lives and the effects caused to the neighboring population (as well as to other nations within a wide radius of radioactivity dispersal), and also with respect to the damage caused in the nuclear plant itself. In the light of the anxiety created internationally, the USSR State Committee for the Utilization of Atomic Energy prepared a report (1), based on the conclusions of the Governmental Commission entrusted to study the causes of the accident, which was presented at the international meeting of experts held at the International Atomic Energy Agency (IAEA) headquarters in Vienna from August 25 to 29, 1986. The present technical report has been prepared by the Spanish nuclear power plants within the framework of UNIDAD ELECTRICA, S.A. (UNESA) - the Association of Spanish electric utilities - in collaboration with EMPRESARIOS AGRUPADOS, S.A. The report reflects the utilities' analyses of the causes and consequences of the accident and, based on similarities and differences with Spanish plants under construction and in operation, intends to: a. Evaluate the possibility of an accident with similar consequences occurring in a Spanish plant b. Identify possible design and operation modifications indicated by the lessons learned from this accident.

  12. The Chernobyl accident and the Spanish nuclear power plants. Technical report

    International Nuclear Information System (INIS)

    1986-11-01

    On the morning of April 26, 1986, Unit 4 of the Chernobyl Nuclear Power Plant (Ukraine, USSR) suffered an accident of the greatest magnitude among those which have taken place in nuclear energy installations employed for peaceful uses. The accident reached a degree of severity unknown up to now in nuclear energy generating plants, both with respect to the loss of human lives and the effects caused to the neighboring population (as well as to other nations within a wide radius of radioactivity dispersal), and also with respect to the damage caused in the nuclear plant itself. In the light of the anxiety created internationally, the USSR State Committee for the Utilization of Atomic Energy prepared a report (1), based on the conclusions of the Governmental Commission entrusted to study the causes of the accident, which was presented at the international meeting of experts held at the International Atomic Energy Agency (IAEA) headquarters in Vienna from August 25 to 29, 1986. The present technical report has been prepared by the Spanish nuclear power plants within the framework of UNIDAD ELECTRICA, S.A. (UNESA) - the Association of Spanish electric utilities - in collaboration with EMPRESARIOS AGRUPADOS, S.A. The report reflects the utilities' analyses of the causes and consequences of the accident and, based on similarities and differences with Spanish plants under construction and in operation, intends to: a. Evaluate the possibility of an accident with similar consequences occurring in a Spanish plant b. Identify possible design and operation modifications indicated by the lessons learned from this accident

  13. 2010 Criticality Accident Alarm System Benchmark Experiments At The CEA Valduc SILENE Facility

    International Nuclear Information System (INIS)

    Miller, Thomas Martin; Dunn, Michael E.; Wagner, John C.; McMahan, Kimberly L.; Authier, Nicolas; Jacquet, Xavier; Rousseau, Guillaume; Wolff, Herve; Piot, Jerome; Savanier, Laurence; Baclet, Nathalie; Lee, Yi-kang; Masse, Veronique; Trama, Jean-Christophe; Gagnier, Emmanuel; Naury, Sylvie; Lenain, Richard; Hunter, Richard; Kim, Soon; Dulik, George Michael; Reynolds, Kevin H.

    2011-01-01

    Several experiments were performed at the CEA Valduc SILENE reactor facility, which are intended to be published as evaluated benchmark experiments in the ICSBEP Handbook. These evaluated benchmarks will be useful for the verification and validation of radiation transport codes and evaluated nuclear data, particularly those that are used in the analysis of CAASs. During these experiments SILENE was operated in pulsed mode in order to be representative of a criticality accident, which is rare among shielding benchmarks. Measurements of the neutron flux were made with neutron activation foils and measurements of photon doses were made with TLDs. Also unique to these experiments was the presence of several detectors used in actual CAASs, which allowed for the observation of their behavior during an actual critical pulse. This paper presents the preliminary measurement data currently available from these experiments. Also presented are comparisons of preliminary computational results with Scale and TRIPOLI-4 to the preliminary measurement data.

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

    International Nuclear Information System (INIS)

    Gieseke, J.A.

    1985-01-01

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

  15. Application of simulation techniques for accident management training in nuclear power plants

    International Nuclear Information System (INIS)

    2003-05-01

    Many IAEA Member States operating nuclear power plants (NPPs) are at present developing accident management programmes (AMPs) for the prevention and mitigation of severe accidents. However, the level of implementation varies significantly between NPPs. The exchange of experience and best practices can considerably contribute to the quality, and facilitate the implementation of AMPs at the plants. Various IAEA activities assist countries in the area of accident management. Several publications have been developed which provide guidance and support in establishing accident management at NPPs. The defence in depth concept in nuclear safety requires that, although highly unlikely, beyond design basis and severe accident conditions should also be considered, in spite of the fact that they were not explicitly addressed in the original design of currently operating nuclear power plants (NPPs). Defence in depth is physically achieved by means of four successive barriers (fuel matrix, cladding, primary coolant boundary, and containment) that prevent the release of radioactive material. These barriers are protected by a set of design measures at three levels, including prevention of abnormal operation and failures (level 1), control of abnormal operation and detection of failures (level 2) and control of accidents within the design basis (level 3). Should these first three levels fail to ensure the structural integrity of the core, additional efforts are made at the fourth level of defence in depth in order to further reduce the risks. The objective at level 4 is to ensure that both the likelihood of an accident entailing significant core damage (severe accident) and the magnitude of radioactive releases following a severe accident are kept as low as reasonably achievable. The term 'accident management' refers to the overall range of capabilities of a NPP and its personnel to both prevent and mitigate accident situations that could lead to severe fuel damage in the reactor

  16. Electrical systems design applications on Japanese PWR plants in light of the Fukushima Daiichi Accident

    International Nuclear Information System (INIS)

    Nomoto, Tsutomu

    2015-01-01

    After the Fukushima Daiichi nuclear power plant (1F-NPP) accident (i.e. Station Blackout), several design enhancements have been incorporated or are under considering to Mitsubishi PWR plants' design of not only operational plants' design but also new plants' design. Especially, there are several important enhancements in the area of the electrical system design. In this presentation, design enhancements related to following electrical systems/equipment are introduced; - Offsite Power System; - Emergency Power Source; - Safety-related Battery; - Alternative AC Power Supply Systems. In addition, relevant design requirements/conditions which are or will be considered in Mitsubishi PWR plants are introduced. (authors)

  17. Problems of probabilistic safety assessment after Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    Sugiyama, Naoki

    2011-01-01

    Probabilistic safety assessment (PSA) methodology to assure nuclear safety is had great expectations of lessons learned from Fukushima Daiichi nuclear power plant (NPP) accident and on the other hand this accident made actualized technical problems of PSA. Effectiveness of current PSA methodology for risk assessment was confirmed by comparing the accident development with accident scenario of PSA and equipment failure rate. From a viewpoint of nuclear safety objective and defense in depth approach of IAEA, technical problems of PSA were (1) extension of PSA for spent fuel pool and waste disposal system as well as level 3PSA for broader environmental contamination and (2) overlapping of accident scenario of plural unit site, balance of high quality plant management and preceding negation, treatment of uncertainty of external events, severe accident measure and human reliability analysis and reflection of disaster prevention capability to level 3PSA. In order to upgrade PSA technology, six proposals were described for nuclear safety and defense in depth, comprehensive evaluation scope and catch-up of latest technology, necessity of strategic preparation of PSA standard, human resources fostering and risk communication. (T. Tanaka)

  18. Root causes and impacts of severe accidents at large nuclear power plants.

    Science.gov (United States)

    Högberg, Lars

    2013-04-01

    The root causes and impacts of three severe accidents at large civilian nuclear power plants are reviewed: the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima Daiichi accident in 2011. Impacts include health effects, evacuation of contaminated areas as well as cost estimates and impacts on energy policies and nuclear safety work in various countries. It is concluded that essential objectives for reactor safety work must be: (1) to prevent accidents from developing into severe core damage, even if they are initiated by very unlikely natural or man-made events, and, recognizing that accidents with severe core damage may nevertheless occur; (2) to prevent large-scale and long-lived ground contamination by limiting releases of radioactive nuclides such as cesium to less than about 100 TBq. To achieve these objectives the importance of maintaining high global standards of safety management and safety culture cannot be emphasized enough. All three severe accidents discussed in this paper had their root causes in system deficiencies indicative of poor safety management and poor safety culture in both the nuclear industry and government authorities.

  19. Root Causes and Impacts of Severe Accidents at Large Nuclear Power Plants

    International Nuclear Information System (INIS)

    Hoegberg, Lars

    2013-01-01

    The root causes and impacts of three severe accidents at large civilian nuclear power plants are reviewed: the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima Daiichi accident in 2011. Impacts include health effects, evacuation of contaminated areas as well as cost estimates and impacts on energy policies and nuclear safety work in various countries. It is concluded that essential objectives for reactor safety work must be: (1) to prevent accidents from developing into severe core damage, even if they are initiated by very unlikely natural or man-made events, and, recognizing that accidents with severe core damage may nevertheless occur; (2) to prevent large-scale and long lived ground contamination by limiting releases of radioactive nuclides such as cesium to less than about 100 TBq. To achieve these objectives the importance of maintaining high global standards of safety management and safety culture cannot be emphasized enough. All three severe accidents discussed in this paper had their root causes in system deficiencies indicative of poor safety management and poor safety culture in both the nuclear industry and government authorities

  20. Root Causes and Impacts of Severe Accidents at Large Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Hoegberg, Lars

    2013-04-15

    The root causes and impacts of three severe accidents at large civilian nuclear power plants are reviewed: the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima Daiichi accident in 2011. Impacts include health effects, evacuation of contaminated areas as well as cost estimates and impacts on energy policies and nuclear safety work in various countries. It is concluded that essential objectives for reactor safety work must be: (1) to prevent accidents from developing into severe core damage, even if they are initiated by very unlikely natural or man-made events, and, recognizing that accidents with severe core damage may nevertheless occur; (2) to prevent large-scale and long lived ground contamination by limiting releases of radioactive nuclides such as cesium to less than about 100 TBq. To achieve these objectives the importance of maintaining high global standards of safety management and safety culture cannot be emphasized enough. All three severe accidents discussed in this paper had their root causes in system deficiencies indicative of poor safety management and poor safety culture in both the nuclear industry and government authorities.

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

    International Nuclear Information System (INIS)

    Jordanov, I.; Khristov, Y.

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

  3. What kind of accidents can happen in a nuclear power plant

    International Nuclear Information System (INIS)

    Debes, M.

    1995-01-01

    The lessons drawn from real reactor accidents are of great value. The safety approach in France relies on defence in depth and takes into account accidents in the plant design, completed by a probabilistic approach and experience feedback. Ultimate procedure are implemented on the basis of severe accidents studies which include core melting or partial containment defect, in order to mitigate their consequences even if they are improbable, and to enable a proper implementation of emergency planning countermeasures. The accident hypothesis and consequences are considered to draw the emergency planning procedures. Off site countermeasures, such as in house-confinement, limited evacuation or iodine distribution, are efficient in limiting the consequences for the public. Experience feedback, in association with a proactive vigilance and prevention policy, is developed in order to detect and correct in a proactive way the root causes of any deviation, even minor, so as to avoid multiple failures and ensure safety. (author). 4 refs., 2 figs., 1 tab

  4. Accident management

    International Nuclear Information System (INIS)

    Lutz, R.J.; Monty, B.S.; Liparulo, N.J.; Desaedeleer, G.

    1989-01-01

    The foundation of the framework for a Severe Accident Management Program is the contained in the Probabilistic Safety Study (PSS) or the Individual Plant Evaluations (IPE) for a specific plant. The development of a Severe Accident Management Program at a plant is based on the use of the information, in conjunction with other applicable information. A Severe Accident Management Program must address both accident prevention and accident mitigation. The overall Severe Accident Management framework must address these two facets, as a living program in terms of gathering the evaluating information, the readiness to respond to an event. Significant international experience in the development of severe accident management programs exist which should provide some direction for the development of Severe Accident Management in the U.S. This paper reports that the two most important elements of a Severe Accident Management Program are the Emergency Consultation process and the standards for measuring the effectiveness of individual Severe Accident Management Programs at utilities

  5. Cancer rates after the Three Mile Island nuclear accident and proximity of residence to the plant.

    Science.gov (United States)

    Hatch, M C; Wallenstein, S; Beyea, J; Nieves, J W; Susser, M

    1991-06-01

    In the light of a possible link between stress and cancer promotion or progression, and of previously reported distress in residents near the Three Mile Island (TMI) nuclear power plant, we attempted to evaluate the impact of the March 1979 accident on community cancer rates. Proximity of residence to the plant, which related to distress in previous studies, was taken as a possible indicator of accident stress; the postaccident pattern in cancer rates was examined in 69 "study tracts" within a 10-mile radius of TMI, in relation to residential proximity. A modest association was found between postaccident cancer rates and proximity (OR = 1.4; 95% CI = 1.3, 1.6). After adjusting for a gradient in cancer risk prior to the accident, the odds ratio contrasting those closest to the plant with those living farther out was 1.2 (95% CI = 1.0, 1.4). A postaccident increase in cancer rates near the Three Mile Island plant was notable in 1982, persisted for another year, and then declined. Radiation emissions, as modeled mathematically, did not account for the observed increase. Interpretation in terms of accident stress is limited by the lack of an individual measure of stress and by uncertainty about whether stress has a biological effect on cancer in humans. An alternative mechanism for the cancer increase near the plant is through changes in care-seeking and diagnostic practice arising from postaccident concern.

  6. Assessment of PASS Effectiveness under Severe Accidents in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Choi, Yu Jung; Lee, Sung Bok; Kim, Hyeong Taek; Lee, Jin Yong

    2008-01-01

    Following the accident at Three Mile Island Unit 2 (TMI-2) on March 28, 1979, the USNRC formed a lessons-learned Task Force to identify and evaluate safety concerns originating with the TMI-2 accident. NUREG-0578 documented the results of the task force effort. One of the recommendations of the task force was for licensees to upgrade the capability to obtain samples from the reactor coolant system and containment atmosphere under high radioactivity conditions and to provide the capability for chemical and spectral analyses of high-level samples on site. NUREG-0737 contained the details of the TMI recommendations that were to be implemented by the licensees. Additional criteria for post accident sampling system(PASS) were issued by Regulatory Guide 1.97. As the results, PASS has been installed on nuclear power plants(NPPs) in Korea as well as United States. However, significant improvements have been achieved since the TMI-2 accident in the areas of understanding risks associated with nuclear plant operations and developing better strategies for managing the response to potential severe accidents at NPPs. Thus, the requirements for PASS have been re-evaluated in some reports. According to the reports, the samples and measurements from PASS do not contribute significantly to emergency management response to severe accidents due to the long analyzing time, 3 hours. Hence, this paper focused on the development of the quantitative analysis methodology to analyze the sequence of the severe accident in Yonggwang nuclear power plants (YGN) and presented the results of the analysis according to the developed methodology

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  9. Results of stress tests of European nuclear power plants after the Fukushima-Daiichi accident

    International Nuclear Information System (INIS)

    Kovacs, Zoltan; Novakova, Helena

    2012-01-01

    In response to the Fukushima-Daiichi accident, the European Council laid down the requirement that a transparent and comprehensive risk assessment exercise ('stress tests') be carried out at each European nuclear power plant. The stress tests concentrated on the nuclear power plants' safety margins in the light of the lessons learned from the accident. The reviews focused on natural external events including earthquake, tsunami and extreme weather, loss of safety functions, and severe accident management. The stress test procedure comprised 3 steps: (i) The nuclear facility operators performed the stress tests and prepared proposals for safety improvements. (ii) The national regulators performed independent reviews of the stress tests and prepared national reports. (iii) The reports submitted by the national regulators were subjected to review at a European level. The article describes the scope of the stress tests and their results, verified at the European level. (orig.)

  10. Securing the Safety of Nuclear Power Plants against Oil Spill Accidents at Sea

    International Nuclear Information System (INIS)

    Hyun, Seung Gyu; Choi, Ho Seon; Kim, Sang Yun

    2008-01-01

    As of 2008, 20 nuclear power plants are under operation and six plants are under construction in Korea. NPPs account for approximately 38% of Korea's electric power production; however, it is expected that the share of power produced by NPPs will be further increased to reduce the level of CO 2 emissions, taking into account the concern over global warming. All of NPPs in Korea are located on the coast to facilitate the supply of cooling water sources. Thus, tar and other floating matters from vessels following oil spill accidents at sea may affect intake systems, and consequently interrupt the supply of cooling water. This study will review cases of response measures taken by NPPs against large-scale crude oil spill accidents that had occurred off the coast of Korea, including such accidents as the Sea Prince (July 23, 1995) and the Hebei Sprit(December 7, 2007), and relevant regulatory requirements at home and abroad

  11. Single parameter controls for nuclear criticality safety at the Oak Ridge Y-12 Plant

    International Nuclear Information System (INIS)

    Baker, J.S.; Peek, W.M.

    1995-01-01

    At the Oak Ridge Y-12 Plant, there are numerous situations in which nuclear criticality safety must be assured and subcriticality demonstrated by some method other than the straightforward use of the double contingency principle. Some cases are cited, and the criticality safety evaluation of contaminated combustible waste collectors is considered in detail. The criticality safety evaluation for combustible collectors is based on applying one very good control to the one controllable parameter. Safety can only be defended when the contingency of excess density is limited to a credible value based on process knowledge. No reasonable single failure is found that will result in a criticality accident. The historically accepted viewpoint is that this meets double contingency, even though there are not two independent controls on the single parameter of interest

  12. Study of time-critical diagnostic method for emergency operation of nuclear power plant

    International Nuclear Information System (INIS)

    Gofuku, A.; Yoshikawa, H.; Itoh, K.; Wakabayashi, J.

    1986-01-01

    In order to support the emergency operation of nuclear power plant, the method of time-critical diagnostic plant analyzer has been investigated. The conception of the emergency operation support center is proposed and two types of plant analyzer may be installed in this center. One analyzer is a real-time tracking simulation code using the observed signals and another is a fast trend-prediction code. A real-time tracking code, TOKRAC, has been developed for analyzing the PWR primary loop thermo-hydraulics at SBLOCA, and the applicability of this code was examined by the numerical experiments for the initial phase transient of both TMI-2 accident and 6% coldleg SBLOCA of a Westinghouse-type PWR plant. The results showed that fairly good tracking was carried out by TOKRAC. The CPU time of TOKRAC was about 12-14 percent of real-time

  13. The Chernobyl Nuclear Power Plant accident: ecotoxicological update

    Science.gov (United States)

    Eisler, R.; Hoffman, David J.; Rattner, Barnett A.; Burton, G. Allen; Cairns, John

    2003-01-01

    The accident at the Chernobyl, Ukraine, nuclear reactor on 26 April 1986 released large amounts of radiocesium and other radionuclides into the environment, contaminating much of the northern hemisphere, especially Europe. In the vicinity of Chernobyl, at least 30 people died, more than 115,000 others were evacuated, and consumption of milk and other foods was banned because of radiocontamination. At least 14,000 human cancer deaths are expected in Russia, Belarus, and the Ukraine as a direct result of Chernobyl. The most sensitive local ecosystems, as judged by survival, were the soil fauna, pine forest communities, and certain populations of rodents. Elsewhere, fallout from Chernobyl significantly contaminated freshwater and terrestrial ecosystems and flesh and milk of domestic livestock; in many cases, radionuclide concentrations in biological samples exceeded current radiation protection guidelines. Reindeer (Rangifer tarandus) in Scandinavia were among the most seriously afflicted by Chernobyl fallout, probably because their main food during winter (lichens) is an efficient absorber of airborne particles containing radiocesium. Some reindeer calves contaminated with 137Cs from Chernobyl showed 137Cs-dependent decreases in survival and increases in frequency of chromosomal aberrations. Although radiation levels in the biosphere are declining with time, latent effects of initial exposure--including an increased frequency of thyroid and other cancers--are now measurable. The full effect of the Chernobyl nuclear reactor accident on natural resources will probably not be known for at least several decades because of gaps in data on long-term genetic and reproductive effects and on radiocesium cycling and toxicokinetics.

  14. The critical safety functions and plant operation

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Church, J.F.; Cross, M.T.; Guinn, W.M.; Porter, N.J.

    1981-01-01

    The operator's role in nuclear safety is outlined and the concept of ''safety functions'' introduced. Safety functions are a group of actions that prevent core melt or minimize radiation releases to the general public. They can be used to provide a hierarchy of practical plant protection that an operator should use. The plant safety evaluation uses four inputs in predicting the results of an event: the event initiator, the plant design, the initial plant conditions and setup, and the operator actions. If any of these inputs are not as assumed in the evaluation, confidence that the consequences will be as predicted is reduced. Based on the safety evaluation, the operator has three roles in assuring that the consequences of an event will be no worse than the predicted acceptable results: Maintain plant setup in readiness to properly respond. Operate the plant in a manner such that fewer, milder events minimize the frequency and the severity of adverse events. Monitor the plant to verify that the safety functions are accomplished. The operator needs a systematic approach to mitigating the consequences of an event. The concept of safety functions introduces this systematic approach and presents a hierarchy of protection. If the operator has difficulty identifying an event for any reason, the systematic safety function approach allows accomplishing the overall path of mitigating consequences. Ten functions designed to protect against core melt, preserve containment integrity, prevent indirect release of radioactivity, and maintain vital auxiliaries needed to support the other safety functions are identified

  15. Method for improving accident sequence recognition in nuclear power plant control rooms

    International Nuclear Information System (INIS)

    Heising, C.D.; Dinsmore, S.C.

    1983-01-01

    This work adapts fault trees from plant-specific probabilistic risk analyses (PRAs) to construct and quantitatively evaluate an alarm analysis system for the engineered safety features (ESFs). The purpose is to help improve reactor operator recognition and identification of potential accident sequences. The PRA system fault trees provide system failure mode information which can be used to construct alarm trees. These alarm trees provide a framework for assessing the plant indicators so that the plant conditions are made more readily apparent to plant personnel. In the alarm tree, possible states of each instrumented alarem are identified as true or false. In addition, a warning status is defined and integrated into an alarm analysis routine. The impact of this additional status conditioned on the Boolean laws used to evaluate the alarm trees is examined. An application is described for BWR high pressure coolant injection system (HPCI) that would be utilized during many severe reactor accidents

  16. Development of passive condensers for accident localization systems at nuclear power plants in the former USSR

    International Nuclear Information System (INIS)

    Kuznecov, M.V.

    1992-01-01

    The development is summarized of passive condensers for accident localization systems at nuclear power plants (with RBMK and WWER reactors) in the former USSR. Basic properties and criteria defining their availability are described, as are experimental tests and technical solution optimization results. (author) 2 fig

  17. MIGRATORY GAME BIRDS AS A SOURCE OF PUBLIC EXPOSURE FROM THE FUKUSHIMA NUCLEAR POWER PLANT ACCIDENT

    Directory of Open Access Journals (Sweden)

    I. P. Stamat

    2011-01-01

    Full Text Available This article examines assessments of the impact of the Fukushima nuclear power plant accident on exposure of the Russian Federation population related to the seasonal migration of game birds. Intake of artificial radionuclides with meat of migratory game birds is shown to be one of the major pathways for the population exposure in the Far Eastern region of the country.

  18. Quantification of severe accident source terms of a Westinghouse 3-loop plant

    International Nuclear Information System (INIS)

    Lee Min; Ko, Y.-C.

    2008-01-01

    Integrated severe accident analysis codes are used to quantify the source terms of the representative sequences identified in PSA study. The characteristics of these source terms depend on the detail design of the plant and the accident scenario. A historical perspective of radioactive source term is provided. The grouping of radionuclides in different source terms or source term quantification tools based on TID-14844, NUREG-1465, and WASH-1400 is compared. The radionuclides release phenomena and models adopted in the integrated severe accident analysis codes of STCP and MAAP4 are described. In the present study, the severe accident source terms for risk quantification of Maanshan Nuclear Power Plant of Taiwan Power Company are quantified using MAAP 4.0.4 code. A methodology is developed to quantify the source terms of each source term category (STC) identified in the Level II PSA analysis of the plant. The characteristics of source terms obtained are compared with other source terms. The plant analyzed employs a Westinghouse designed 3-loop pressurized water reactor (PWR) with large dry containment

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  20. Results of Questionnaire for the member of JHPS concerning the criticality accident at Tokai

    International Nuclear Information System (INIS)

    2000-01-01

    During the investigation of the criticality accident at Tokai occurring on Sep. 30, 1999, the project team in Japan Health Physics Society (JHPS) carried out a questionnaire for the member on the accident and this paper summarized its results. The effective answer was obtained in 36% of members. Major questions (and frequent answers) were: media of information obtained (internet 33%, TV and radio 22%, and newspaper 19%); concerning actions done by Japanese and local governments, the recommendation on Sep. 30 at 15:00 of evacuation for people living in the area within the radius of 350 m (necessary 92%), timing of its release on Oct. 2 at 18:30 (appropriate 41% and too late 36%) and its information to the people (more information needed 60%) and the recommendation on Sep. 30 at 22:30 of in-door refuge within 10 km radius (unnecessary 43% and necessary 41%), timing of its release on Oct. 1 at 16:40 (too late 49%) and its information to the people (more information needed 63%); and safety declaration for food etc. on Oct. 2 at 18:30 (necessary 92%). Based on above results and free description on the questionnaire, JHPS considered the necessity of described systems of JHPS for emergency.(K.H.)

  1. Early clinical consequences of victims in JCO criticality accident in Tokaimura

    International Nuclear Information System (INIS)

    Suzuki, Gen

    2000-01-01

    The JCO criticality accident occurred at 10:35 on September 30, 1999 when two workers (O and S) poured the solution of uranyl nitrate into the precipitation tank and one (Y) worked at desk in the neighboring room. O's symptoms were unconsciousness, rigidity and emesis, and S's, numbness. The three were moved to Mito National Hospital by an ambulance car at 12:07 and then to the Hospital of National Institute of Radiological Sciences by the helicopter and car at 15:25, where contamination of their cloths by Na-24, suggesting the exposure to neutron, was found. O exhibited emesis within 10 min after the accident and diarrhea, unconsciousness and severe pyrexia within 1 hr, suggesting he had undergone the lethal exposure of >8 Gy. S showed emesis, light unconsciousness and numbness within 1 hr, suggesting >6 Gy and Y did not show even emesis, less dose exposure than the two. They underwent firstly the drip of sodium hydrogen carbonate (due to possible internal exposure of uranium), oxygen inhalation and then corticoid injection as well as the drip of antibiotics. At that day, they had the special therapy with pentoxyphylline and L-glutamine+elementary diet. Later, in the Hospital of Tokyo University, O and S had the heamopoietic stem cell transplantation. At present, O passed away, S is still in hospital and Y is discharged. (K.H.)

  2. Surveillance of Strontium-90 in Foods after the Fukushima Daiichi Nuclear Power Plant Accident.

    Science.gov (United States)

    Nabeshi, Hiromi; Tsutsumi, Tomoaki; Uekusa, Yoshinori; Hachisuka, Akiko; Matsuda, Rieko; Teshima, Reiko

    2015-01-01

    As a result of the Fukushima Daiichi nuclear power plant (NPP) accident, various radionuclides were released into the environment. In this study, we surveyed strontium-90 ((90)Sr) concentrations in several foodstuffs. Strontium-90 is thought to be the third most important residual radionuclide in food collected after the Fukushima Daiichi, NPP accident after following cesium-137 ((137)Cs) and cesium-134 ((134)Cs). Results of (90)Sr analyses indicated that (90)Sr was detect in 25 of the 40 radioactive cesium (r-Cs) positive samples collected in areas around the Fukushima Daiichi NPP, ranging in distance from 50 to 250 km. R-Cs positive samples were defined as containing both (134)Cs and (137)Cs which are considered to be indicators of the after-effects of the Fukushima Daiichi NPP accident. We also detected (90)Sr in 8 of 13 r-Cs negative samples, in which (134)Cs was not detected. Strontium-90 concentrations in the r-Cs positive samples did not significantly exceed the (90)Sr concentrations in r-Cs negative samples or the (90)Sr concentration ranges in comparable food groups found in previous surveys before the Fukushima Daiichi NPP accident. Thus, (90)Sr concentrations in r-Cs positive samples were indistinguishable from the background (90)Sr concentrations arising from global fallout prior to the Fukushima accident, suggesting that no marked increase of (90)Sr concentrations has occurred in r-Cs positive samples as a result of the Fukushima Daiichi NPP accident.

  3. Response to the accident at TEPCO's Fukushima Daiichi Nuclear Power Plants

    International Nuclear Information System (INIS)

    Nei, Hisanori

    2012-01-01

    This article was reading from the author's plenary lecture at the thermal and nuclear power generation convention 2011, which was summary of the author edited report of Japanese government to IAEA ministerial conference on nuclear safety. The article consisted of (1) outlines of occurrence and development of the accident at TEPCO's Fukushima Daiichi Nuclear Power Plants (NPPs), (2) comparison of Fukushima Daiichi NPPs with other NPPs (Fukushima Daini, Onagawa and Tokai Daini NPPs), (3) major countermeasures to settle the situation regarding the accident, (4) comprehensive safety evaluation of other NPPs as response to the accident and (5) lessons learned from the accident so far. It was highly important to ensure power supplies and robust cooling functions of reactors, pressure containment vessels and spent fuel pools. 28 lessons were categorized into five groups such as (1) strengthen preventive measures against a severe accident, (2) enhancement of response measures against severe accidents, (3) enhancement of nuclear emergency responses, (4) reinforcement of safety infrastructure and (5) thoroughness of safety culture. (T. Tanaka)

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

    International Nuclear Information System (INIS)

    1986-08-01

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

  5. Quick evaluation of the neutron dose following a criticality accident by measurement of sodium 24 activity

    International Nuclear Information System (INIS)

    Tabardel, R.; Ricourt, A.; Parmentier, N.

    1984-07-01

    In order to quickly sort out the irradiated individuals following a criticality accident, the neutron dose can be evaluated quickly by measuring the sodium-24 activity induced in the human body. The report supplies the information necessary for this evaluation from the response of various detectors of current use in radiation protection. The first part describes the method of evaluation of sodium-24 activity (A) given by the reading (M) of each instrument. The second part describes the method of kerma evaluation from the measured sodium-24 activity. The third part is an experimental application of the method of kerma evaluation from the sodium-24 activity measured in a phantom irradiated in the SILENE reactor flux. The results given by radiation protection instruments are in good agreement with the calculated values for a front exposure and demonstrate the usefulness of measuring the induced sodium-24 activity by radiation protection instruments of current use [fr

  6. Improved set of criticality accident detectors used in the intercomparison experiment in Valduc

    International Nuclear Information System (INIS)

    Jozefowicz, K.; Golnik, N.

    1996-01-01

    An improved set of critically accident detectors has been elaborated for the needs of the Inst. of Atomic Energy in Swierk. The sets, which consist of fission track detectors, wide-base silicon diodes and RPL glasses, were tested in the international intercomparison experiment in Valduc, France. Comparison of our results with the reference measurements showed a good agreement (within 25%) for both the neutron and gamma measurements. Additionally, the diode response to neutron kerma was investigated more extensively in the dose range between 2 and 10 Gy, where the dependence of the diode signal versus neutron kerma was not well known. A possibility of the multiple use of the diodes has been proved. (author)

  7. The classification of cases related to Tokai-mura criticality accident. Mental care after radiation exposure

    International Nuclear Information System (INIS)

    Minoshita, Seiko; Satoh, Shinji

    2012-01-01

    Cases classified into each pattern, which the authors have met so far after the criticality accident JCO was introduced. Case is introduced, based on multiple cases actually met in medical institutions, has been created as a model case. When the cases that were considered related to the criticality accident in Tokai-mura was summarized, the cases could be classified by the time consultation. Therefore the cases were discussed along the time, also discussed about the time. From the first year to the second year, the most cases seen were the cases with high anxiety. Then, there were many cases which symptoms were worsened by the impact received through the residents meeting. Among the patients who received counseling from half a year to three years after the incident, the onset of mental illness, and the aggravation of the mental disease increased, too. After two or three years of the incident, there were a lot of consultation with women who were pregnant or had infants then. Four years later, men gradually came to have consultation at hospitals. In addition, the consultation of alcohol from problems of a family member has increased. In the first year, there were many patients that a symptom turned worse since they were shocked by the booing of the residents meeting. On the other hand, the patients that a symptom turned worse because of the prolonged issue increased four years later. Four or five years, after the incident the cases of because of bankruptcy or dismissal, life been deteriorated economically were increased, and some cases were led to the discrete of family in a chain reaction. Approximately 10 years later, due to the increase of the aging population, the amount of patient who were frightened because they got cancer increased since they lost the people around them as a result of cancer. (author)

  8. SARTEMP2 - A computer program to calculate power and temperatures in a transport flask during a criticality accident

    International Nuclear Information System (INIS)

    Shaw, P.M.

    1983-04-01

    The computer code SARTEMP2, an extended version of the original SARTEMP program, which calculates the power and temperatures in a transport flask during a hypothetical criticality accident is described. The accident arises, it is assumed, during the refilling of the flask with water, bringing the system to delayed critical. As the water level continues to rise, reactivity is added causing the power to rise, and thus temperatures in the fuel, clad and water to increase. The point kinetics equations are coupled to the one-dimensional heat conduction equation. The model used, the method of solution of the equations and the input data required are given. (author)

  9. Fukushima Nuclear Accident, the Third International Severe Nuclear Power Plant Accident

    International Nuclear Information System (INIS)

    Rashad, S.M.

    2013-01-01

    Japan is the world's third largest power user. Japan's last remaining nuclear reactor shutdown on Saturday 4 Th of May 2012 leaving the country entirely nuclear free. All of 50 of the nation's operable reactors (not counting for the four crippled reactors at Fukushima) are now offline. Before last year's Fukushima nuclear disaster, the country obtained 30% of its energy from nuclear plants, and had planned to produce up to 50% of its power from nuclear sources by 2030. Japan declared states of emergency for five nuclear reactors at two power plants after the units lost cooling ability in the aftermath of Friday 11 March 2011 powerful earthquake. Thousands of (14000) residents were immediately evacuated as workers struggled to get the reactors under control to prevent meltdowns. On March 11 Th, 2011, Japan experienced a sever earthquake resulting in the shutdown of multiple reactors. At Fukushima Daiichi site, the earthquake caused the loss of normal Ac power. In addition it appeals that the ensuing tsunami caused the loss of emergency Ac power at the site. Subsequent events caused damage to fuel and radiological releases offsite. The spent fuel problem is a wild card in the potentially catastrophic failure of Fukushima power plant. Since the Friday's 9.0 earthquake, the plant has been wracked by repeated explosions in three different reactors. Nuclear experts emphasized there are significant differences between the unfolding nuclear crisis at Fukushima and the events leading up to the Chernobyl disaster in 1986. The Chernobyl reactor exploded during a power surge while it was in operation and released a major cloud of radiation because the reactor had no containment structure around to. At Fukushima, each reactor has shutdown and is inside a 20 cm-thick steel pressure vessel that is designed to contain a meltdown. The pressure vessels themselves are surrounded by steel-lined, reinforced concrete shells. Chernobyl disaster was classified 7 on the International

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

    International Nuclear Information System (INIS)

    1986-01-01

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

  11. Accident for natural gas well with hydrogen sulfide in relation to nuclear power plant siting

    International Nuclear Information System (INIS)

    Tan Chengjun; Shangguang Zhihong; Sha Xiangdong

    2010-01-01

    In order to make assessment to the potential impact from accident of natural gas wells with hydrogen sulfide on the habitability of main control room of nuclear power plant (NPP), several assumptions such as source terms of maximum credible accident, conservative atmospheric conditions and release characteristics were proposed in the paper, and the impact on the habitability of main control room was evaluated using toxicity thresholds recommended by foreign authority. Case results indicate that the method can provide the reference for the preliminary assessment to external human-induced events during the siting phrase of NPP. (authors)

  12. Telephone counseling for the public after the Fukushima Daiichi Nuclear Power Plant accident

    International Nuclear Information System (INIS)

    Horiguchi, T.; Kojima, K.; Itoh, T.

    2011-01-01

    After the Fukushima Daiichi Nuclear Power Plant accident, Kinki University Atomic Energy Research Institute provided telephone counseling services in order to respond the public's growing concerns about radiation and nuclear energy. Three telephone lines were newly installed for the counseling and the number of consultation marked 705 between March 24 and April 2. In this report, by summarizing the contents of the counseling, we will show what the public concerned about shortly after the accident and report how we responded to the concerns. (author)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  14. Grundremmingen nuclear power plant: The accident that came from the cold

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The accident of January 13, 1977, is shortly described. The power plant was cut off from the network by an extreme temperature drop and high humidity, which caused a break in a number of porcelain insulators of two HV lines. The response of the turbine speed control was delayed, and there was a sudden pressure drop in the primary steam line. Weakly radioactive steam and water were released into the containment without polluting the environment, since all safety systems responded to the accident. (HP) [de

  15. Tests of qualification of national components of nuclear power plants under design basis accident

    International Nuclear Information System (INIS)

    Mesquita, A.Z.

    1990-01-01

    With the purpose of qualifying national components of nuclear power plants, whose working must be maintained during and after an accident, the Thermohydraulic Division of CDTN have done tests to check the equipment stability, under Design Basis Accident conditions. Until this moment, the following components were tested: electrical junction boxes (connectors); coating systems for wall, inside cover and steel containment; hydraulics components of personnel and equipment airlock. This work describes the test instalation, the tests performed and its results. The components tested, in a general way, fulfil the specified requirements. (author) [pt

  16. Theoretical Derivation of Simplified Evaluation Models for the First Peak of a Criticality Accident in Nuclear Fuel Solution

    International Nuclear Information System (INIS)

    Nomura, Yasushi

    2000-01-01

    In a reprocessing facility where nuclear fuel solutions are processed, one could observe a series of power peaks, with the highest peak right after a criticality accident. The criticality alarm system (CAS) is designed to detect the first power peak and warn workers near the reacting material by sounding alarms immediately. Consequently, exposure of the workers would be minimized by an immediate and effective evacuation. Therefore, in the design and installation of a CAS, it is necessary to estimate the magnitude of the first power peak and to set up the threshold point where the CAS initiates the alarm. Furthermore, it is necessary to estimate the level of potential exposure of workers in the case of accidents so as to decide the appropriateness of installing a CAS for a given compartment.A simplified evaluation model to estimate the minimum scale of the first power peak during a criticality accident is derived by theoretical considerations only for use in the design of a CAS to set up the threshold point triggering the alarm signal. Another simplified evaluation model is derived in the same way to estimate the maximum scale of the first power peak for use in judging the appropriateness for installing a CAS. Both models are shown to have adequate margin in predicting the minimum and maximum scale of criticality accidents by comparing their results with French CRiticality occurring ACcidentally (CRAC) experimental data

  17. Estimation of the economic impacts of Three Mile Island nuclear power plant accident

    International Nuclear Information System (INIS)

    Sagara, Aya; Fujimoto, Noboru; Fukuda, Kenji

    1998-01-01

    The Three Mile Island nuclear power plant accident had an immediate negative impact on the economy of the seven-country area which surrounds the plant site. In order to estimate the social effect of the nuclear power plant accident economically, immediate and short term economical impacts on some industrial classification have been evaluated. The economical effect to Metropolitan Edison Co., the circumstantial payment of the insurance and the lawsuit for the compensation for damages, etc. have been estimated at dollar 90 million for the manufacturing and nonmanufacturing industry, dollar 5 million for the tourist industry and dollar 50,000 for agriculture. The total loss for the state and country governments is about dollar 90,000. Metropolitan Edison Co. expended also dollar 111 million for the substitute energy and dollar 760 million for the decontamination cost. Since the lawsuit for the compensation for damages is still continuing, the total impacts cost is calculated more than a billion dollar. (author)

  18. Safety study on nuclear heat utilization system - accident delineation and assessment on nuclear steelmaking pilot plant

    International Nuclear Information System (INIS)

    Yoshida, T.; Mizuno, M.; Tsuruoka, K.

    1982-01-01

    This paper presents accident delineation and assessment on a nuclear steelmaking pilot plant as an example of nuclear heat utilization systems. The reactor thermal energy from VHTR is transported to externally located chemical process plant employing helium-heated steam reformer by an intermediate heat transport loop. This paper on the nuclear steelmaking pilot plant will describe (1) system transients under accident conditions, (2) impact of explosion and fire on the nuclear reactor and the public and (3) radiation exposure on the public. The results presented in this paper will contribute considerably to understanding safety features of nuclear heat utilization system that employs the intermediate heat transport loop and the helium-heated steam reformer

  19. Model experiments on depressurisation accidents in nuclear process heat plants (HTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Fritsching, G.; Wolf, G. [Internationale Atomreaktorbau G.m.b.H. (INTERATOM), Bergisch Gladbach (Germany, F.R.)

    1981-01-15

    The analysis of depressurisation accidents requires the use of digital computer programs to find out the dynamic loads acting on the plant structures. Because of the importance of such accidents in safety and licensing procedures of nuclear process heat plants, it is necessary to compare these computer results with suitable experiments to show the accuracy and the limits of the programs in question. For this purpose a series of depressurisation experiments has been started at INTERATOM on a small scale model of a primary loop of a nuclear process heat plant. Using the results of these experiments three different computer programs were tested with good success. The development of the experimental program and the estimation of the results was carried out in co-operation with KFA-Juelich and the Technische Hochschule Aachen.

  20. Model experiments on depressurisation accidents in nuclear process heat plants (HTGR)

    International Nuclear Information System (INIS)

    Fritsching, G.; Wolf, G.

    1981-01-01

    The analysis of depressurisation accidents requires the use of digital computer programs to find out the dynamic loads acting on the plant structures. Because of the importance of such accidents in safety and licensing procedures of nuclear process heat plants, it is necessary to compare these computer results with suitable experiments to show the accuracy and the limits of the programs in question. For this purpose a series of depressurisation experiments has been started at INTERATOM on a small scale model of a primary loop of a nuclear process heat plant. Using the results of these experiments three different computer programs were tested with good success. The development of the experimental program and the estimation of the results was carried out in co-operation with KFA-Juelich and the Technische Hochschule Aachen

  1. Protective action alternatives for accidents at nuclear power plants

    International Nuclear Information System (INIS)

    Tingle, A.; Pratt, W.T.; McGuire, S.A.

    1987-06-01

    Protective action calculations have been performed for five different light water reactors (LWRs) and containment designs using high and low fission product releases for early and late containment failures for each plant. These fission product release estimates were obtained from studies performed for the recently published ''Reactor Risk Reference Document'' (NUREG-1150). Five protective actions were considered for the risks of exceeding various dose levels to the red marrow versus centerline distance from the plants using site-specific meteorology. The strategies considered were 4 hours of normal activity, basement sheltering, large building sheltering, evacuation at release, and evacuation 1 hour after release. The evacuations were computed using 10 mph evacuation speed for all sites. Additional calculations were performed for the dose contributions due to the cloud, ground, and inhalation pathways

  2. Risks of potential accidents of nuclear power plants in Europe

    International Nuclear Information System (INIS)

    Slaper, H.; Eggink, G.J.; Blaauboer, R.O.

    1993-12-01

    This report is focussed on an integrated assessment of probabilistic cancer mortality risks due to possible accidental releases from the European nuclear power plants. For each of the European nuclear power plants the probability of accidental releases per year of operation is combined with the consequences in terms of the excess doses received over a lifetime (70 years). Risk estimates are restricted to cancer mortality and do not include immediate or short term deaths in the direct vicinity ( -8 per year in Western Europe. Going East the risks increase gradually to over 1000 x 10 -8 per year in regions of the former Soviet Union, where reactors of the Chernobyl type are located. The nuclear power plants in the East European countries dominate the estimated risk pattern and contribute at least 40-50% to the average risk in the West European countries. Improving the reactor safety in eastern European countries could lead to considerable reductions in estimated excess mortality risks. In western Europe the mortality risk might be reduced by a factor of two, and in eastern Europe by a factor of 100 to 1000. (orig.)

  3. PNNL Measurement Results for the 2016 Criticality Accident Dosimetry Exercise at the Nevada National Security Stite (IER-148)

    Energy Technology Data Exchange (ETDEWEB)

    Rathbone, Bruce A.; Morley, Shannon M.; Stephens, John A.

    2017-05-01

    The Pacific Northwest National Laboratory (PNNL) participated in a criticality accident dosimetry intercomparison exercise held at the Nevada National Security Site (NNSS) May 24-27, 2016. The exercise was administered by Lawrence Livermore National Laboratory (LLNL) and consisted of three exposures performed using the Godiva-IV critical assembly housed in the Device Assembly Facility (DAF) located on the NNSS site. The exercise allowed participants to test the ability of their nuclear accident dosimeters to meet the performance criteria in ANSI/HPS N13.3-2013, Dosimetry for Criticality Accidents and to obtain new measurement data for use in revising dose calculation methods and quick sort screening methods where appropriate. PNNL participated with new prototype Personal Nuclear Accident Dosimeter (PNAD) and Fixed Nuclear Accident Dosimeter (FNAD) designs as well as the existing historical PNAD design. The new prototype designs incorporate optically stimulated luminescence (OSL) dosimeters in place of thermoluminescence dosimeters (TLDs), among other design changes, while retaining the same set of activation foils historically used. The default dose calculation methodology established decades ago for use with activation foils in PNNL PNADs and FNADs was used to calculate neutron dose results for both the existing and prototype dosimeters tested in the exercise. The results indicate that the effective cross sections and/or dose conversion factors used historically need to be updated to accurately measure the operational quantities recommended for nuclear accident dosimetry in ANSI/HPS N13.3-2013 and to ensure PNAD and FNAD performance meets the ANSI/HPS N13.3-2013 performance criteria. The operational quantities recommended for nuclear accident dosimetry are personal absorbed dose, Dp(10), and ambient absorbed dose, D*(10).

  4. Knowledge data base for severe accident management of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    For the safety enhancement of Nuclear Power Plants (NPPs), continuous efforts are very important to take in the up-to-date scientific and technical knowledge positively and to reflect them into the safety regulation. The purpose of the present study is to gather effectively the scientific and technical knowledge about the severe accident (SA) phenomena and the accident management (AM) for prevention and mitigation of SA, and to take in the experimental data by participating in the international cooperative experiments regarding the important SA phenomena and the effectiveness of AM. Based on those data and knowledge, JNES is developing and improving severe accident analysis models to maintain the SA analysis codes and the AM knowledge base for assessment of the NPPs in Japan. The activities in fiscal year 2012 are as follows; Analytical study on OECD/NEA projects such as MCCI, SERENA and SFP projects, and support in making regulation for SA. (author)

  5. Knowledge data base for severe accident management of nuclear power plants

    International Nuclear Information System (INIS)

    2013-01-01

    For the safety enhancement of Nuclear Power Plants (NPPs), continuous efforts are very important to take in the up-to-date scientific and technical knowledge positively and to reflect them into the safety regulation. The purpose of the present study is to gather effectively the scientific and technical knowledge about the severe accident (SA) phenomena and the accident management (AM) for prevention and mitigation of SA, and to take in the experimental data by participating in the international cooperative experiments regarding the important SA phenomena and the effectiveness of AM. Based on those data and knowledge, JNES is developing and improving severe accident analysis models to maintain the SA analysis codes and the AM knowledge base for assessment of the NPPs in Japan. The activities in fiscal year 2012 are as follows; Analytical study on OECD/NEA projects such as MCCI, SERENA and SFP projects, and support in making regulation for SA. (author)

  6. Medical emergency planning in case of severe nuclear power plant accidents

    International Nuclear Information System (INIS)

    Ohlenschlaeger, L.

    1980-01-01

    This paper is an attempt to discuss a three-step-plan on medical emergency planning in case of severe accidents at nuclear power plants on the basis of own experiences in the regional area as well as on the basis of recommendations of the Federal Minister of the Interior. The medical considerations take account of the severity and extension of an accident whereby the current definitions used in nuclear engineering for accident situations are taken as basis. A comparison between obligatory and actual state is made on the possibilities of medical emergency planning, taking all capacities of staff, facilities, and equipment available in the Federal Republic of Germany into account. To assure a useful and quick utilization of the existing infra-structure as well as nation-wide uniform training of physicians and medical assistants in the field of medical emergency in case of a nuclear catastrophe, a federal law for health protection is regarded urgently necessary. (orig.) [de

  7. The ENEA criticality accident dosimetry system: a contribution to the 2002 international intercomparison at the SILENE reactor.

    Science.gov (United States)

    Gualdrini, G; Bedogni, R; Fantuzzi, E; Mariotti, F

    2004-01-01

    The present paper summarises the activity carried out at the ENEA Radiation Protection Institute for updating the methodologies employed for the evaluation of the neutron and photon dose to the exposed workers in case of a criticality accident, in the framework of the 'International Intercomparison of Criticality Accident Dosimetry Systems' (Silène reactor, IRSN-CEA-Valduc June 2002). The evaluation of the neutron spectra and the neutron dosimetric quantities relies on activation detectors and on unfolding algorithms. Thermoluminescent detectors are employed for the gamma dose measurement. The work is aimed at accurately characterising the measurement system and, at the same time, testing the algorithms. Useful spectral information were included, based on Monte Carlo simulations, to take into account the potential accident scenarios of practical interest. All along this exercise intercomparison a particular attention was devoted to the 'traceability' of all the experimental and computational parameters and therefore, aimed at an easy treatment by the user.

  8. The impact of the accident at Three Mile Island on plant control and instumentation philosophy

    International Nuclear Information System (INIS)

    Catlow, F.

    1983-01-01

    Independent commissions which were appointed to evaluate the causes of the accident at the Three Mile Island nuclear power plant in the USA exposed major weakness in the man/machine interface which they felt might be common to other similar plants. Strengthening this link is regarded as twofold: i) Educating the man to enhance his understanding of plant processes; ii) Improving the machine interface. The paper reviews suggested improvements in instumentation which would aid the control of a nuclear plant. These comprise mainly: a) The application of human factors engineering principles to control room design in order to make the 'machine' more manageable; b) improved data feedback so that the operator can make an accurate assessment of plant status at any instant. The author considers that there is a likelihood that the general philosophy of the man/machine interface being applied to the nuclear industry could be applied to some extent to conventional power plants and even other industries

  9. Tritium in Japanese precipitation following the March 2011 Fukushima Daiichi Nuclear Plant accident

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Takuya, E-mail: t.matsumoto@iaea.org [Isotope Hydrology Section, Division of Physical and Chemical Sciences, International Atomic Energy Agency, Vienna International Centre, 1400 Vienna (Austria); Maruoka, Teruyuki [Division of Integrative Environmental Sciences Graduate School of Life and Environmental Sciences, University of Tsukuba, 1-1-1 Tennodai, Tsukuba City, Ibaraki 305-8572 (Japan); Shimoda, Gen [Geological Survey of Japan, National Institute of Advanced Industrial Science and Technology, 1-1-1 Higashi, Tsukuba City, Ibaraki 305-8561 (Japan); Obata, Hajime [Atmosphere and Ocean Research Institute, The University of Tokyo, 5-1-5, Kashiwanoha, Kashiwa-shi, Chiba 277-8564 (Japan); Kagi, Hiroyuki [Geochemical Research Center, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-0033 (Japan); Suzuki, Katsuhiko [Japan Agency for Marin-Earth Science and Technology, 2-15, Natsushima, Yokosuka, Kanagawa 237-0061 (Japan); Yamamoto, Koshi [Graduate School of Environmental Studies, Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8601 (Japan); Mitsuguchi, Takehiro [215 Ooma Akadoji-cho Konan, 483-8226 (Japan); Usa Marine Biological Institute, Kochi University, 194 Inoshiri, Usa, Tosa, Kochi 781-1164 (Japan); Hagino, Kyoko; Tomioka, Naotaka [Institute for Study of the Earth' s Interior, Okayama University at Misasa, 827 Yamada, Misasa, Tottori 682-0193 (Japan); Sambandam, Chinmaya; Brummer, Daniela; Klaus, Philipp Martin; Aggarwal, Pradeep [Isotope Hydrology Section, Division of Physical and Chemical Sciences, International Atomic Energy Agency, Vienna International Centre, 1400 Vienna (Austria)

    2013-02-15

    Tritium concentrations in Japanese precipitation samples collected after the March 2011 accident at the Fukushima Dai-ichi Nuclear Power Plant (FNPP1) were measured. Values exceeding the pre-accident background were detected at three out of seven localities (Tsukuba, Kashiwa and Hongo) southwest of the FNPP1 at distances varying between 170 and 220 km from the source. The highest tritium content was found in the first rainfall in Tsukuba after the accident; however concentrations were 500 times less than the regulatory limit for tritium in drinking water. Tritium concentrations decreased steadily and rapidly with time, becoming indistinguishable from the pre-accident values within five weeks. The atmospheric tritium activities in the vicinity of the FNPP1 during the earliest stage of the accident was estimated to be 1.5 × 10{sup 3} Bq/m{sup 3}, which is potentially capable of producing rainwater exceeding the regulatory limit, but only in the immediate vicinity of the source. - Highlights: ► We measured the {sup 3}H contents of Japanese rain collected after the Fukushima accident. ► {sup 3}H level became 30 times higher than pre-accident level in the first rain at Tsukuba. ► Some locality within 220 km from the source showed elevated {sup 3}H levels. ► These high {sup 3}H signals disappear in a few weeks. ► Atmospheric {sup 3}H level at the source during the earliest stage was estimated to be 1500 Bq/m{sup 3}.

  10. Assessment of potential doses to workers during postulated accident conditions at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Hoover, M.D.; Newton, G.J.; Farrell, R.F.

    1996-01-01

    This qualitative hazard evaluation systematically assessed potential doses to workers during postulated accident conditions at the U.S. Department of Energy's Waste Isolation Pilot Plant (WIPP). Postulated accidents included the spontaneous ignition of a waste drum, puncture of a waste drum by a forklift, dropping of a waste drum from a forklift, and simultaneous dropping of seven drums during a crane failure. The descriptions and estimated frequencies of occurrence for these accidents were developed by the Hazard and Operability Study for CH TRU Waste Handling System (WCAP 14312). The estimated materials at risk, damage ratios, airborne release fractions and respirable fractions for these accidents were taken from the 1995 Safety Analysis Report (SAR) update and from the DOE handbook Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities (DOE-HDBK-3010-94). A Monte Carlo simulation was used to estimate the range of worker exposures that could result from each accident. Guidelines for evaluating the adequacy of defense-in-depth for worker protection at WIPP were adopted from a scheme presented by the International Commission on Radiological Protection in its publication on Protection from Potential Exposure: A Conceptual Framework (ICRP Publication 64). Probabilities of exposures greater than 5, 50, and 300 rem were less than 10 -2 , 10 -4 , and 10 -6 per year, respectively. In conformance with the guidance of DOE standard 3009-94, Appendix A (draft), we emphasize that use of these evaluation guidelines is not intended to imply that these numbers constitute acceptable limits for worker exposure under accident conditions. However, in conjunction with the extensive safety assessment in the 1995 SAR update, these results indicate that the Carlsbad Area Office strategy for the assessment of hazards and accidents assures the protection of workers, as well as members of the public and the environment

  11. Bolted Flanged Connection for Critical Plant/Piping Systems

    International Nuclear Information System (INIS)

    Efremov, Anatoly

    2006-01-01

    A novel type of Bolted Flanged Connection with bolts and gasket manufactured on a basis of advanced Shape Memory Alloys is examined. Presented approach combined with inverse flexion flange design of plant/piping joint reveals a significant increase of internal pressure under conditions of a variety of operating temperatures relating to critical plant/piping systems. (author)

  12. Overview of training methodology for accident management at nuclear power plants

    International Nuclear Information System (INIS)

    2005-04-01

    Many IAEA Member States operating nuclear power plants (NPPs) are at present developing accident management programmes (AMPs) for the prevention and mitigation of severe accidents. However, the level of implementation varies significantly between NPPs. The exchange of experience and best practices can considerably contribute to the quality and facilitate the implementation of AMPs at the plants. The main objective of this publication is to describe available material and technical support tools that can be used to support training of the personnel involved in the accident management (AM), and to highlight the current status of their application. The focus is on those operator aids that can help the plant personnel to take correct actions during an emergency to prevent and mitigate consequences of a severe accident. The second objective is to describe the available material for the training courses of those people who are responsible of the AMP development and implementation of an individual plant. The third objective is to collect a compact set of information on various aspects of AM training into a single publication. In this context, the AM personnel includes both the plant staff responsible for taking the decision and actions concerning preventive and mitigative AM and the persons involved in the management of off-site releases. Thus, the scope of this publication is on the training of personnel directly involved in the decisions and execution of the SAM actions during progression of an accident. The integration of training into the AMP development and implementation is summarized. The technical AM support tools and material are defined as operator aids involving severe accident guidelines, various computational aids and computerized tools. The operator aids make also an essential part of the training tools. The simulators to be applied for the AM training have been developed or are under development by various organizations in order to support the training on

  13. Analysis of the accident at Fukushima Daiichi nuclear power plant in an A BWR reactor

    International Nuclear Information System (INIS)

    Escorcia O, D.; Salazar S, E.

    2016-09-01

    The present work aims to recreate the accident occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, making use of an academic simulator of forced circulation of the A BWR reactor provided by the IAEA to know the scope of this simulator. The simulator was developed and distributed by the IAEA for academic purposes and contains the characteristics and general elements of this reactor to be able to simulate transients and failures of different types, allowing also to observe the general behavior of the reactor, as well as several phenomena and present systems in the same. Is an educational tool of great value, but it does not have a scope that allows the training of plant operators. To recreate the conditions of the Fukushima accident in the simulator, we first have to know what events led to this accident, as well as the actions taken by operators and managers to reduce the consequences of this accident; and the sequence of events that occurred during the course of the accident. Differences in the nuclear power plant behavior are observed and interpreted throughout the simulation, since the Fukushima plant technology and the simulator technology are not the same, although they have several elements in common. The Fukushima plant had an event that by far exceeded the design basis, which triggered in an accident that occurred in the first place by a total loss of power supply, followed by the loss of cooling systems, causing a level too high in temperature, melting the core and damaging the containment accordingly, allowing the escape of hydrogen and radioactive material. As a result of the simulation, was determined that the scope of the IAEA academic simulator reaches the entrance of the emergency equipment, so is able to simulate almost all the events occurred at the time of the earthquake and the arrival of the tsunami in the nuclear power plant of Fukushima Daiichi. However, due to its characteristics, is not able to simulate later

  14. The technical requirements concerning severe accident management in nuclear power plants

    International Nuclear Information System (INIS)

    Okamoto, Koji; Sugiyama, Tomoyuki; Kamata, Shinya

    2014-01-01

    The Great East Japan Earthquake with a magnitude of 9.0 (The 2011 off the Pacific coast of Tohoku Earthquake) occurred on March 11, 2011, and the beyond design-basis tsunami descended on the Fukushima Daiichi Nuclear Power Plant by the earthquake. Eventually, the core cooling systems of the units 1, 2 and 3 could not operate stably, they all suffered severe accident, and hydrogen explosions were triggered in the reactor buildings of units 1, 3 and 4. In the light of these circumstances, Atomic Energy Society of Japan (AESJ) decided to establish a standard that consolidates the concept of maintaining and improving severe accident management. In the SAM standard, the combination of hardware and software measures based on the risk assessment enables a scientific and rational approach to apply to scenarios of various severe accidents including low-frequency, high-impact events, and assures safety with functionality and flexibility. The SAM standard is already established in March, 2014. After publication of the SAM standard, with regard to effectiveness assessment for accident management and treatment of the uncertainty of severe accident analysis code, for example, the detailed guideline will be prepared as appendices of the standard. (author)

  15. TEPCO's costs and risks which invited the nuclear power plant accident

    International Nuclear Information System (INIS)

    Soeda, Takashi

    2017-01-01

    The National Diet of Japan Fukushima Nuclear Accident Independent Investigation Commission (Diet Accident Investigation Commission) considered two patterns against the tsunami risk of nuclear plant: (1) Risk management for the purpose of safety (Pattern A), and (2) Risk management for the purpose of utilization rate and cost of nuclear reactor (Pattern B). Pattern B emphasizes avoiding 'countermeasure cost generation' and 'operation shutdown' rather than preparing for a tsunami that we do not know when to come. Diet Accident Investigation Commission analyzed that the behavioral principles concerning the crisis response of Tokyo Electric Power Company (TEPCO) had the stronger tendency of Pattern B. Regarding the accident of TEPCO, there were class actions that asked the responsibility of TEPCO and the government. This paper examined the contents of the opinions of government-side experts submitted for this issue. The government-side experts argued that there was no 'scientific consensus' for tsunami forecast, and that preliminary measures against unexpected tsunami was impossible. However, both of these government's arguments are irrational due to difference from the fact. TEPCO president at the time of accident insisted in the firm that 'cost cut in another dimension' was indispensable and reduced expenses. TEPCO and the government had continued Pattern B, even knowing that tsunami risk measures were insufficient from more than ten years ago. (A.O.)

  16. Tritium in Japanese precipitation following the March 2011 Fukushima Daiichi Nuclear Plant accident.

    Science.gov (United States)

    Matsumoto, Takuya; Maruoka, Teruyuki; Shimoda, Gen; Obata, Hajime; Kagi, Hiroyuki; Suzuki, Katsuhiko; Yamamoto, Koshi; Mitsuguchi, Takehiro; Hagino, Kyoko; Tomioka, Naotaka; Sambandam, Chinmaya; Brummer, Daniela; Klaus, Philipp Martin; Aggarwal, Pradeep

    2013-02-15

    Tritium concentrations in Japanese precipitation samples collected after the March 2011 accident at the Fukushima Dai-ichi Nuclear Power Plant (FNPP1) were measured. Values exceeding the pre-accident background were detected at three out of seven localities (Tsukuba, Kashiwa and Hongo) southwest of the FNPP1 at distances varying between 170 and 220 km from the source. The highest tritium content was found in the first rainfall in Tsukuba after the accident; however concentrations were 500 times less than the regulatory limit for tritium in drinking water. Tritium concentrations decreased steadily and rapidly with time, becoming indistinguishable from the pre-accident values within five weeks. The atmospheric tritium activities in the vicinity of the FNPP1 during the earliest stage of the accident was estimated to be 1.5×10(3) Bq/m(3), which is potentially capable of producing rainwater exceeding the regulatory limit, but only in the immediate vicinity of the source. Copyright © 2012 Elsevier B.V. All rights reserved.

  17. Teaching of severe accident of Fukushima Daiichi Nuclear Power Plants of Tokyo Electric Power

    International Nuclear Information System (INIS)

    Saito, Shinzo

    2011-01-01

    The Great East Japan Earthquake and accompanied tsunami brought about the severe accident at Fukushima Daiichi Nuclear Power Plants of Tokyo Electric Power Co., Inc. For 'No more Fukushima', twelve teaching of the accident was pointed out as follows: 1) natural disasters and external events shall be taken into consideration, 2) severe accident shall be included into safety regulation, 3) all possibility of hydrogen explosion shall be excluded, 4) diversity of safety important component and equipment shall be added with sufficient period of outage, 5) siting of multiple units at the same site shall be avoided at quake-prone country like Japan, 6) accident response environment for operators shall be improved, 7) accident convergence termination system shall be established so as to concentrate technical experience and knowledge, 8) off-site center shall be improved, 9) resident evacuation, consumption limit of food, radiation exposure and soil contamination limit shall be decided openly, 10) nuclear regulation and prevention of disaster shall be conducted by unitary organization to gain public trust, 11) fostering of safety culture among relevant enterprises shall be more encouraged and 12) nuclear industry shall develop reactor such as with no core meltdown or no evacuation and environmental contamination even if reactor core would be meltdown. (T. Tanaka)

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

    International Nuclear Information System (INIS)

    Sonnenkalb, M.; Rohde, J.

    1999-01-01

    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)

  19. Internal event analysis for Laguna Verde Unit 1 Nuclear Power Plant. Accident sequence quantification and results

    International Nuclear Information System (INIS)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R.

    1994-01-01

    The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the I nternal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant, CNSNS-TR 004, in five volumes. The reports are organized as follows: CNSNS-TR 004 Volume 1: Introduction and Methodology. CNSNS-TR4 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR 004 Volume 3: System Analysis. CNSNS-TR 004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR 005 Volume 5: Appendices A, B and C. This volume presents the development of the dependent failure analysis, the treatment of the support system dependencies, the identification of the shared-components dependencies, and the treatment of the common cause failure. It is also presented the identification of the main human actions considered along with the possible recovery actions included. The development of the data base and the assumptions and limitations in the data base are also described in this volume. The accident sequences quantification process and the resolution of the core vulnerable sequences are presented. In this volume, the source and treatment of uncertainties associated with failure rates, component unavailabilities, initiating event frequencies, and human error probabilities are also presented. Finally, the main results and conclusions for the Internal Event Analysis for Laguna Verde Nuclear Power Plant are presented. The total core damage frequency calculated is 9.03x 10-5 per year for internal events. The most dominant accident sequences found are the transients involving the loss of offsite power, the station blackout accidents, and the anticipated transients without SCRAM (ATWS). (Author)

  20. Examination of some assumed severe reactor accidents at the Olkiluoto nuclear power plant

    International Nuclear Information System (INIS)

    Pekkarinen, E.; Rossi, J.

    1989-02-01

    Knowledge and analysis methods of severe accidents at nuclear power plants and of subsequent response of primary system and containment have been developed in last few years to the extent that realistic source tems of the specified accident sequences can be calculated for the Finnish nuclear power plants. The objective of this investigation was to calculate the source terms of off-site consequences brought about by some selected severe accident sequences initiated by the total loss of on-site and off-site AC power at the Olkiluoto nuclear power plant. The results describing the estimated off-site health risks are expressed as conditional assuming that the accident has taken place, because the probabilities of the occurence of the accident sequences considered have not been analysed in this study. The range and probabilities of occurence of health detriments are considered by calculating consequences in different weeather conditions and taking into account the annual frequency of each weather condition and statistical population distribution. The calculational results indicate that the reactor building provides and additional holdup and deposition of radioactive substance (except coble gases) released from the containment. Furthermore, the release fractions of the core inventory to the environment of volatile fission products such as iodine, cesium and tellurium remain under 0.03. No early health effects are predicted for the surrounding population in case the assumed short-tem countermeasures are performed effectively. Acute health effects are extremely improbable even without any active countermeasure. By reducing the long-term exposure from contaminated agricultural products, the collective dose from natural long-term background radiation, for instance in the sector of 30 degrees towards the southern Finland up to the distance of 300 kilometers, would be expected to increase with 2-20 percent depending on the release considered

  1. Dose estimation from food intake due to the Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    Yamaguchi, Ichiro; Terada, Hiroshi; Kunugita, Naoki; Takahashi, Kunihiko

    2013-01-01

    Since the Fukushima Daiichi nuclear power plant accident, concerns have arisen about the radiation safety of food raised at home and abroad. Therefore, many measures have been taken to address this. To evaluate the effectiveness of these measures, dose estimation due to food consumption has been attempted by various methods. In this paper, we show the results of dose estimation based on the monitoring data of radioactive materials in food published by the Ministry of Health, Labour and Welfare. The Radioactive Material Response Working Group in the Food Sanitation Subcommittee of the Pharmaceutical Affairs and Food Sanitation Council reported such dose estimation results on October 31, 2011 using monitoring data from immediately after the accident through September, 2011. Our results presented in this paper were the effective dose and thyroid equivalent dose integrated up to December 2012 from immediately after the accident. The estimated results of committed effective dose by age group derived from the radioiodine and radiocesium in food after the Fukushima Daiichi nuclear power plant accident showed the highest median value (0.19 mSv) in children 13-18 years of age. The highest 95% tile value, 0.33 mSv, was shown in the 1-6 years age range. These dose estimations from food can be useful for evaluation of radiation risk for individuals or populations and for radiation protection measures. It would also be helpful for the study of risk management of food in the future. (author)

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

    International Nuclear Information System (INIS)

    Abe, Hitoshi; Nishio; Gunji; Naito, Yoshitaka

    1993-11-01

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

  3. Flood control construction of Shidao Bay nuclear power plant and safety analysis for hypothetical accident of HTR-PM

    International Nuclear Information System (INIS)

    Chen Yongrong; Zhang Keke; Zhu Li

    2014-01-01

    A series of events triggered by tsunami eventually led to the Fukushima nuclear accident. For drawing lessons from the nuclear accident and applying to Shidao Bay nuclear power plant flood control construction, we compare with the state laws and regulations, and prove the design of Shidao Bay nuclear power plant flood construction. Through introducing the history of domestic tsunamis and the national researches before and after the Fukushima nuclear accident, we expound the tsunami hazards of Shidao Bay nuclear power plant. In addition, in order to verify the safety of HTR-PM, we anticipate the contingent accidents after ''superposition event of earthquake and extreme flood'', and analyse the abilities and measures of HTR-PM to deal with these beyond design basis accidents (BDBA). (author)

  4. Phenomenological uncertainty analysis of early containment failure at severe accident of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Su Won

    2011-02-15

    The severe accident has inherently significant uncertainty due to wide range of conditions and performing experiments, validation and practical application are extremely difficult because of its high temperature and pressure. Although internal and external researches were put into practice, the reference used in Korean nuclear plants were foreign data of 1980s and safety analysis as the probabilistic safety assessment has not applied the newest methodology. Also, it is applied to containment pressure formed into point value as results of thermal hydraulic analysis to identify the probability of containment failure in level 2 PSA. In this paper, the uncertainty analysis methods for phenomena of severe accident influencing early containment failure were developed, the uncertainty analysis that apply Korean nuclear plants using the MELCOR code was performed and it is a point of view to present the distribution of containment pressure as a result of uncertainty analysis. Because early containment failure is important factor of Large Early Release Frequency(LERF) that is used as representative criteria of decision-making in nuclear power plants, it was selected in this paper among various modes of containment failure. Important phenomena of early containment failure at severe accident based on previous researches were comprehended and methodology of 7th steps to evaluate uncertainty was developed. The MELCOR input for analysis of the severe accident reflected natural circulation flow was developed and the accident scenario for station black out that was representative initial event of early containment failure was determined. By reviewing the internal model and correlation for MELCOR model relevant important phenomena of early containment failure, the uncertainty factors which could affect on the uncertainty were founded and the major factors were finally identified through the sensitivity analysis. In order to determine total number of MELCOR calculations which can

  5. Development of the simulation system IMPACT for analysis of nuclear power plant severe accidents

    International Nuclear Information System (INIS)

    Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi

    1997-01-01

    The Nuclear Power Engineering Corporation (NUPEC) has initiated a long-term program to develop the simulation system IMPACT for analysis of hypothetical severe accidents in nuclear power plants. IMPACT employs advanced methods of physical modeling and numerical computation, and can simulate a wide spectrum of senarios ranging from normal operation to hypothetical, beyond-design-basis-accident events. Designed as a large-scale system of interconnected, hierarchical modules, IMPACT's distinguishing features include mechanistic models based on first principles and high speed simulation on parallel processing computers. The present plan is a ten-year program starting from 1993, consisting of the initial one-year of preparatory work followed by three technical phases: Phase-1 for development of a prototype system; Phase-2 for completion of the simulation system, incorporating new achievements from basic studies; and Phase-3 for refinement through extensive verification and validation against test results and available real plant data

  6. Safety improvements at Canadian nuclear power plants in the aftermath of Fukushima accident

    International Nuclear Information System (INIS)

    Rzentkowski, G.; Khouaja, H.

    2014-01-01

    This paper describes the safety review of operating nuclear power plants undertaken by the Canadian Nuclear Safety Commission in light of the March 11, 2011 accident at the Fukushima Daiichi Nuclear Power Plants (NPPs). The review confirmed that the Canadian NPPs are robust and have a strong design relying on multiple layers of defence to protect the public from credible external events. Nevertheless, in the spirit of continuous safety improvements, the review identified a number of recommendations to further strengthen reactor defence-in-depth in preventing and mitigating the consequences of beyond design basis accidents, enhance onsite and offsite emergency response, and improve the CNSC regulatory framework. Progress achieved to date, in implementing these measures, is described in this paper along with a summary of safety benefits for each level of the reactor defence-in-depth. (author)

  7. Safety improvements at Canadian nuclear power plants in the aftermath of Fukushima accident

    Energy Technology Data Exchange (ETDEWEB)

    Rzentkowski, G.; Khouaja, H. [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2014-07-01

    This paper describes the safety review of operating nuclear power plants undertaken by the Canadian Nuclear Safety Commission in light of the March 11, 2011 accident at the Fukushima Daiichi Nuclear Power Plants (NPPs). The review confirmed that the Canadian NPPs are robust and have a strong design relying on multiple layers of defence to protect the public from credible external events. Nevertheless, in the spirit of continuous safety improvements, the review identified a number of recommendations to further strengthen reactor defence-in-depth in preventing and mitigating the consequences of beyond design basis accidents, enhance onsite and offsite emergency response, and improve the CNSC regulatory framework. Progress achieved to date, in implementing these measures, is described in this paper along with a summary of safety benefits for each level of the reactor defence-in-depth. (author)

  8. Nuclear accidents at the Fukushima Dai-ichi power plant. History, events and consequences

    International Nuclear Information System (INIS)

    Berniolles, Jean Marc

    2011-01-01

    Written few weeks after the accident, this article first recalls the circumstances (earthquake and tsunami), and then describes the accidental process within the primary vessels of the Fukushima Dai-ichi number 1, 2 and 3 reactors. The author then describes the interventions which aimed at cooling these three reactors, the problem faced for the storage of used fuels, and then the sequence of accidents: loss of cooling means leading to an explosion, problems faced in the different storage pools. He describes the various steps of recovery (primary cooling, electricity supply), discusses the consequences in terms of radioactivity releases in the plant environment with a comparison with Chernobyl, and also in terms of nature and quantity of radioactive elements. He comments radioactivity controls and measurements, evacuation measures, measurements performed by the IAEA, measurements of sea radioactivity, and the establishment of maps of ground radioactivity around the plant. He discusses the perspectives associated with these measurements for the surroundings of the Fukushima site

  9. Enhancement of organizational resilience in light of the Fukushima Dai-ichi Nuclear Power Plant accident (1). Analysis of responding structure

    International Nuclear Information System (INIS)

    Yoshizawa, Atsufumi; Furuhama, Yutaka; Mutou, Keiko; Oba, Kyoko; Kitamura, Masaharu

    2014-01-01

    Through the critical situations experienced at the Fukushima Daiichi Nuclear Power Plant accident, it became evident that plant personnel are the essential driving force toward resilience for mitigating the severe nuclear accident. In such situation, the key factors are skill and attitude of human operators who struggled sincerely against the accident. It is also evident that future nuclear safety needs to aim at Safety- which is a newly introduced notion of safety proposed in conjunction with resilience engineering. Safety-I was defined as a condition where as little as possible went wrong;. Safety-II is defined as a condition where as much as possible goes wright. Among the four core capabilities (i.e. Learning, Responding, Monitoring, Anticipating) proposed in the framework of resilience engineering, the constituents of 'Responding' is mainly studied in this paper. Four constituents such as Skill Attitude, Health and Environment have been identified through in-depth reviewing of accident reports and reflection of one of the authors who served as a unit director of Fukushima Daiichi Nuclear Power Plant. (author)

  10. Aerosols released from solvent fire accidents in reprocessing plants

    International Nuclear Information System (INIS)

    Jordan, S.; Lindner, W.

    1985-01-01

    Thermodynamic, aerosol characterizing and radiological data of solvent fires in reprocessing plants have been established in experiments. These are the main results: Depending on the ventilation in the containment, kerosene-TBP mixtures burn at a rate up to 120 kg/m 2 h. The aqueous phase of inorganic-organic mixtures might be released during the fire. The gaseous reaction products contain unburnable acidic compounds. Solvents with TBP-nitrate complex shows higher (up to 25%) burning rates than pure solvents (kerosene-TBP). The nitrate complex decomposes violently at about 130 0 C with a release of acid and unburnable gases. Up to 20% of the burned kerosene-TBP solvents are released during the fire in the form of soot particles, phosphoric acid and TBP decomposition products. The particles have an aerodynamic mass median diameter of about 0.5 μm and up to 1.5% of the uranium fixed in the TBP-nitrate complex is released during solvent fires. (orig.)

  11. Report of the Fukushima nuclear accident by the National Academy of Science. Lessons learned from the Fukushima nuclear accident for improving safety of U.S. nuclear plants

    International Nuclear Information System (INIS)

    Nariai, Hideki

    2014-01-01

    U.S. National Academy of Science investigated the accident at the Fukushima Daiichi nuclear plant initiated by the Great East Japan Earthquake for two years and published a draft report in July 24, 2014. Investigation results were summarized in nine new findings and made ten recommendations in a wide horizon; (1) hardware countermeasures against severe accidents and training of operators, (2) upgrade of risk assessment capability for beyond design basis accident, (3) incorporation of new information about hazards in safety regulations, (4) needed improvement of off-site emergency preparedness, and (5) improvements of nuclear safety culture. New information about hazards related with tsunami assessment, new risk assessment for beyond design basis accident, advice of foreigner resident evacuations, regulatory capture, and safety culture and regulator's specialty were discussed as Japanese issues. (T. Tanaka)

  12. Common Risk Target for severe accidents of nuclear power plants based on IAEA INES scale

    International Nuclear Information System (INIS)

    Vitázková, Jiřina; Cazzoli, Errico

    2013-01-01

    The IAEA has repeatedly recommended that the nuclear community should arrive at a common understanding and definition of safety goals for severe accidents in nuclear power plants. The recommendation has only found partial answers, despite the numerous working groups and forums devoted to this effort. The most widely accepted definition of goals is based on the concept of Large (Early) Release Frequencies (L(E)RF) and its derivatives, a surrogate concept derived from results of Probabilistic Safety Assessments (PSAs) which was first introduced in the USA almost twenty years ago and much later accepted by the USNRC for risk informed decision making, but not for safety demonstrations. Other types of Safety Goals have been adopted by some nuclear authorities, but the main drawback of all current definitions is that they may apply only to LWRs. The lack of unifying safety/risk parameter throughout of PSAs worldwide is the basis of the present work, and an attempt is made to arrive at the definition of a Risk Target for severe accidents in NPPs, consistent with the IAEA definitions having a technical basis, which can be adopted without modifications for Generation IV power plants. The proposal of Common Risk Target in this work represents an attempt to define a Common Risk Target based on technical reasoning, reflecting IAEA definitions as well as harmonization requirements raised by the whole European Community in various OECD, ASAMPSA2 and SARNET (Guentay et al., 2006) conclusions and Council Directive of The European Union (Community Framework, 2009) as well as lastly performed stress tests of nuclear power plants throughout the Europe (Peer Review Report, 2012). The basic concept of CRT was first introduced and developed within the European project ASAMPSA2 by the authors of this article and was accepted by majority of world PSA experts participating in final evaluation and survey of the project (Guentay, 2011). In the proposed Risk Target concept an innovative

  13. Common Risk Target for severe accidents of nuclear power plants based on IAEA INES scale

    Energy Technology Data Exchange (ETDEWEB)

    Vitázková, Jiřina, E-mail: jirina@snus.sk [Vitázková-Vitty, Sládkovičova 24, 900 28 Ivanka pri Dunaji (Slovakia); Cazzoli, Errico, E-mail: erik.cazzoli@gmx.net [Cazzoli Consulting, Wiesenweg 14, CH-5415 Nussbaumen (Switzerland)

    2013-09-15

    The IAEA has repeatedly recommended that the nuclear community should arrive at a common understanding and definition of safety goals for severe accidents in nuclear power plants. The recommendation has only found partial answers, despite the numerous working groups and forums devoted to this effort. The most widely accepted definition of goals is based on the concept of Large (Early) Release Frequencies (L(E)RF) and its derivatives, a surrogate concept derived from results of Probabilistic Safety Assessments (PSAs) which was first introduced in the USA almost twenty years ago and much later accepted by the USNRC for risk informed decision making, but not for safety demonstrations. Other types of Safety Goals have been adopted by some nuclear authorities, but the main drawback of all current definitions is that they may apply only to LWRs. The lack of unifying safety/risk parameter throughout of PSAs worldwide is the basis of the present work, and an attempt is made to arrive at the definition of a Risk Target for severe accidents in NPPs, consistent with the IAEA definitions having a technical basis, which can be adopted without modifications for Generation IV power plants. The proposal of Common Risk Target in this work represents an attempt to define a Common Risk Target based on technical reasoning, reflecting IAEA definitions as well as harmonization requirements raised by the whole European Community in various OECD, ASAMPSA2 and SARNET (Guentay et al., 2006) conclusions and Council Directive of The European Union (Community Framework, 2009) as well as lastly performed stress tests of nuclear power plants throughout the Europe (Peer Review Report, 2012). The basic concept of CRT was first introduced and developed within the European project ASAMPSA2 by the authors of this article and was accepted by majority of world PSA experts participating in final evaluation and survey of the project (Guentay, 2011). In the proposed Risk Target concept an innovative

  14. Assessment of risks of accidents and normal operation at nuclear power plants

    International Nuclear Information System (INIS)

    Savolainen, Ilkka; Vuori, Seppo.

    1977-01-01

    A probabilistic assessment model for the analysis of risks involved in the operation of nuclear power plants is described. With the computer code ARANO it is possible to estimate the health and economic consequences of reactor accidents both in probabilistic and deterministic sense. In addition the code is applicable to the calculation of individual and collective doses caused by the releases during normal operation. The estimation of release probabilities and magnitudes is not included in the model. (author)

  15. Forest and Chernobyl: forest ecosystems after the Chernobyl nuclear power plant accident: 1986-1994

    International Nuclear Information System (INIS)

    Ipatyev, V.; Bulavik, I.; Baginsky, V.; Goncharenko, G.; Dvornik, A.

    1999-01-01

    This paper reports basic features of radionuclide migration and the prediction of the radionuclide redistribution and accumulation by forest phytocoenoses after the Chernobyl Nuclear Power Plant (CNPP) accident. The current ecological condition of forest ecosystems is evaluated and scientific aspects of forest management in the conditions of the large-scale radioactive contamination are discussed. (Copyright (c) 1999 Elsevier Science B.V., Amsterdam. All rights reserved.)

  16. Plant accident dynamics of high-temperature reactors with direct gas turbine cycle

    International Nuclear Information System (INIS)

    Waloch, M.L.

    1977-01-01

    In the paper submitted, a one-dimensional accident simulation model for high-temperature reactors with direct-cycle gas turbine (single-cycle facilities) is described. The paper assesses the sudden failure of a gas duct caused by the double-ended break of one out of several parallel pipes before and behind the reactor for a non-integrated plant, leading to major loads in the reactor region, as well as the complete loss of vanes of the compressor for an integrated plant. The results of the calculations show especially high loads for the break of a hot-gas pipe immediately behind the flow restrictors of the reactor outlet, because of prolonged effects of pressure gradients in the reactor region and the maximum core differential pressure. A plant accident dynamics calculation therefore allows to find a compromise between the requirements of stable compressor operation, on the one hand, and small loads in the reactor in the course of an accident, on the other, by establishing in a co-ordinated manner the narrowing ratio of the flow restrictors. (GL) [de

  17. Measures for preventing and mitigating severe accidents of nuclear power plants

    International Nuclear Information System (INIS)

    Lin Chengge

    1993-01-01

    Safety goals, integrity of the containment, accident management, functions of existing equipment and measures and emergency preparedness are discussed as technical basis for implementing the new safety code on the nuclear power plant safety design (HAF-0200(91)). The main quantitative safety goals are presented as core melt frequency -5 /ry for new plants and -4 /ry for existing or constructed plants, and 0.1% I, Cs release frequency -6 /ry. To keep the integrity of the containment, main efforts should be placed on the prevention of early failure of the containment and by pass or isolation failures. Should a late failure of the containment occur at a high probability, measures such as filtering vent should be considered. The leak rate of the containment could be higher than the previous 0.1-0.5 wt%/day, depending on the source term and dose results. But, a limiting leak rate of 1 wt%/day is defined. Accident management involves emergency operating procedures, training and retraining for the AM and adding some supporting equipment and display and diagnostic system for the AM. Those requirements are described. Emergency preparedness and measures can reduced the risk significantly. In the most case of accidents, sheltering is preferred as an effective protective actions

  18. State of Level 2 analyses and severe accident management in Spanish nuclear power plants

    International Nuclear Information System (INIS)

    Otero, R.

    1998-01-01

    The state of the PSA/IPE studies in the Spanish NPPs is presented in this report, as well as the plans to implement the severe accident management strategy both in the Spanish BWRs and PWRs. First, the Spanish LWRs are introduced, and the scope of the IPE analyses required by the Spanish Regulatory Commission (CSN) is given. The general overview is completed with the current degree of development for the IPE studies in each plant. In the second part the methods and tools are shown which are used by the Spanish plants to develop their Level 2 analysis. The different strategies for severe accident management adopted by the BWPs and PWRs in Spain are also outlined. The sources and implementation of the Severe Accident Guidelines (SAG) are described. More detail is given in the following chapters to the containment analysis of Trillo (PWR, KWU design) and Cofrentes (BWR/6, GE design) NPPs, whose development is being carried out by IBERDROLA. The analysis which has been performed up to date for Trillo is limited to the Plant Damage State (PDS) definition. However, the main phenomena challenging its containment performance have been identified, and they are summarized here. The Level 2 analysis for Cofrentes is comparatively more developed. The main phenomena and the key equipment affecting its containment behaviour are presented. Finally the conclusions of this report are elaborated. (author)

  19. Procedural and submittal guidance for the individual plant examination of external events (IPEEE) for severe accident vulnerabilities

    International Nuclear Information System (INIS)

    Chen, J.T.; Chokshi, N.C.; Kenneally, R.M.; Kelly, G.B.; Beckner, W.D.; McCracken, C.; Murphy, A.J.; Reiter, L.; Jeng, D.

    1991-06-01

    Based on a Policy statement on Severe Accidents, the licensee of each nuclear power plant is requested to perform an individual plant examination. The plant examination systematically looks for vulnerabilities to severe accidents and cost-effective safety improvements that reduce or eliminate the important vulnerabilities. This document presents guidance for performing and reporting the results of the individual plant examination of external events (IPEEE). The guidance for reporting the results of the individual plant examination of internal events (IPE) is presented in NUREG-1335. 53 refs., 1 figs., 2 tabs

  20. Assessment of potential doses to workers during postulated accident conditions at the Waste Isolation Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hoover, M.D.; Farrell, R.F. [DOE, Carlsbad, NM (United States); Newton, G.J.

    1995-12-01

    The recent 1995 WIPP Safety Analysis Report (SAR) Update provided detailed analyses of potential radiation doses to members of the public at the site boundary during postulated accident scenarios at the U.S. Department of Energy`s Waste Isolation Pilot Plant (WIPP). The SAR Update addressed the complete spectrum of potential accidents associated with handling and emplacing transuranic waste at WIPP, including damage to waste drums from fires, punctures, drops, and other disruptions. The report focused on the adequacy of the multiple layers of safety practice ({open_quotes}defense-in-depth{close_quotes}) at WIPP, which are designed to (1) reduce the likelihood of accidents and (2) limit the consequences of those accidents. The safeguards which contribute to defense-in-depth at WIPP include a substantial array of inherent design features, engineered controls, and administrative procedures. The SAR Update confirmed that the defense-in-depth at WIPP is adequate to assure the protection of the public and environment. As a supplement to the 1995 SAR Update, we have conducted additional analyses to confirm that these controls will also provide adequate protection to workers at the WIPP. The approaches and results of the worker dose assessment are summarized here. In conformance with the guidance of DOE Standard 3009-94, we emphasize that use of these evaluation guidelines is not intended to imply that these numbers constitute acceptable limits for worker exposures under accident conditions. However, in conjunction with the extensive safety assessment in the 1995 SAR Update, these results indicate that the Carlsbad Area Office strategy for the assessment of hazards and accidents assures the protection of workers, members of the public, and the environment.

  1. The Fukushima Daiichi Nuclear Power Plant Accident: OECD/NEA Nuclear Safety Response and Lessons Learnt

    International Nuclear Information System (INIS)

    2013-01-01

    Following the March 2011 accident at the Fukushima Daiichi nuclear power plant, all NEA member countries took early action to ensure and confirm the continued safety of their nuclear power plants and the protection of the public. After these preliminary safety reviews, all countries with nuclear facilities carried out comprehensive safety reviews, often referred to as 'stress tests', which reassessed safety margins of nuclear facilities with a primary focus on challenges related to conditions experienced at the Fukushima Daiichi nuclear power plant, for example extreme external events and the loss of safety functions, or capabilities to cope with severe accidents. As appropriate, improvements are being made to safety and emergency response systems to ensure that nuclear power plants are capable of withstanding events that lead to loss of electrical power and/or cooling capability. In the weeks following the accident, the NEA immediately began establishing expert groups in the nuclear safety and radiological protection areas, as well as contributing to information exchange with the Japanese authorities and other international organisations. It promptly provided a forum for high-level decision makers and regulators within the G8-G20 frameworks. The NEA actions taken at the international level in response to the accident have been carried out primarily by the three NEA standing technical committees concerned with nuclear and radiation safety issues - the Committee on Nuclear Regulatory Activities (CNRA), the Committee on the Safety of Nuclear Installations (CSNI) and the Committee on Radiation Protection and Public Health (CRPPH) - under the leadership of the CNRA. More than two years following the accident, the NEA continues to assist the Japanese authorities in dealing with their nuclear safety and recovery efforts as well as to facilitate international co-operation on nuclear safety and radiological protection matters. It is strongly supporting the establishment of

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  3. Criticality accident dosimetry systems: an international intercomparison at the SILENE reactor in 2002.

    Science.gov (United States)

    Médioni, R; Asselineau, B; Verrey, B; Trompier, F; Itié, C; Texier, C; Muller, H; Pelcot, G; Clairand, I; Jacquet, X; Pochat, J L

    2004-01-01

    In criticality accident dosimetry and more generally for high dose measurements, special techniques are used to measure separately the gamma ray and neutron components of the dose. To improve these techniques and to check their dosimetry systems (physical and/or biological), a total of 60 laboratories from 29 countries (America, Europe, Asia) participated in an international intercomparaison, which took place in France from 9 to 21 June 2002, at the SILENE reactor in Valduc and at a pure gamma source in Fontenay-aux-Roses. This intercomparison was jointly organised by the IRSN and the CEA with the help of the NEA/OCDE and was partly supported by the European Communities. This paper describes the aim of this intercomparison, the techniques used by the participants and the two radiation sources and their characteristics. The experimental arrangements of the dosemeters for the irradiations in free air or on phantoms are given. Then the dosimetric quantities measured and reported by the participants are summarised, analysed and compared with the reference values. The present paper concerns only the physical dosimetry and essentially experiments performed on the SILENE facility. The results obtained with the biological dosimetry are published in two other papers of this issue.

  4. Fukushima Daiichi Nuclear Power Plant accident: facts, environmental contamination, possible biological effects, and countermeasures.

    Science.gov (United States)

    Anzai, Kazunori; Ban, Nobuhiko; Ozawa, Toshihiko; Tokonami, Shinji

    2012-01-01

    On March 11, 2011, an earthquake led to major problems at the Fukushima Daiichi Nuclear Power Plant. A 14-m high tsunami triggered by the earthquake disabled all AC power to Units 1, 2, and 3 of the Power Plant, and carried off fuel tanks for emergency diesel generators. Despite many efforts, cooling systems did not work and hydrogen explosions damaged the facilities, releasing a large amount of radioactive material into the environment. In this review, we describe the environmental impact of the nuclear accident, and the fundamental biological effects, acute and late, of the radiation. Possible medical countermeasures to radiation exposure are also discussed.

  5. Modelling of radioactive fallout in the vicinity of Chernobyl nuclear power plant accident

    International Nuclear Information System (INIS)

    Israel, Y.A.; Petrov, V.N.; Severov, D.A.

    1988-03-01

    Deposition of radioactive products escaping into the atmosphere for a long time from the Chernobylsk-4 reactor resulting in residual radioactive contamination of the region at a distance of up to 100 km from the nuclear power plant is considered. The suggested model may be used for estimation of the possible scope of nuclear danger in the regions of nuclear power plants and creation of conditions ensuring safety of the population at possible accidents. The following topics are developed: height of elevation and conditions of radionuclide transfer in the atmosphere; dynamics of release and dispersive composition of radioactive products; calculations of radiation levels at a close trace [fr

  6. The main regularities of 137Cs accumulation by medicinal plants after nuclear accidents

    International Nuclear Information System (INIS)

    Orlov, A.A.; Krasnov, V.P.; Get'manchuk, A.I.

    2004-01-01

    The main regularities of 137 Cs accumulation by medicinal plants after nuclear accidents have been analyzed. Tendencies in study of this problem have been underlined on literary data. The mean values of transfer factor of 137 Cs from soil to medicinal row in different habitat types have been elucidated for Ukrainian Polessye. It was found that species with the wide ecological amplitude were characterized by the highest intensity of 137 Cs accumulation in forest habitats in comparison with non-forest ones. For some species of medicinal plants multiyear dynamics of 137 Cs specific activity has been shown on stationary experimental plots. (author)

  7. Effect of marine condition on feature of natural circulation after accident in floating nuclear power plant

    International Nuclear Information System (INIS)

    Yang Fan; Zhang Dan; Tan Changlu; Ran Xu; Yu Hongxing

    2015-01-01

    The incline and swing effect on natural circulation of floating nuclear power plant under site black out (SBO) accident is studied using self-developing marine condition system code RELAP5/MC. It shows that, for floating nuclear power plant under marine condition, the pressurizer fluctuating flow rate, the parallel heat sink (steam generator) have significant influences on the direct passive reactor heat removal (PRHR) system, which is different from other secondary PRHR under marine condition. The flow exchange between the loop and the pressurizer have major effect on cooling capacity for the left side loop. (authors)

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

    International Nuclear Information System (INIS)

    Zhu Xinguan

    1987-12-01

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

  9. Development of hydrogeological modelling approaches for assessment of consequences of hazardous accidents at nuclear power plants

    International Nuclear Information System (INIS)

    Rumynin, V.G.; Mironenko, V.A.; Konosavsky, P.K.; Pereverzeva, S.A.

    1994-07-01

    This paper introduces some modeling approaches for predicting the influence of hazardous accidents at nuclear reactors on groundwater quality. Possible pathways for radioactive releases from nuclear power plants were considered to conceptualize boundary conditions for solving the subsurface radionuclides transport problems. Some approaches to incorporate physical-and-chemical interactions into transport simulators have been developed. The hydrogeological forecasts were based on numerical and semi-analytical scale-dependent models. They have been applied to assess the possible impact of the nuclear power plants designed in Russia on groundwater reservoirs

  10. Criticality safety assessment of a TRIGA reactor spent-fuel pool under accident conditions

    International Nuclear Information System (INIS)

    Glumac, B.; Ravnik, M.; Logar, M.

    1997-01-01

    Additional criticality safety analysis of a pool-type storage for TRIGA spent fuel at the Jozef Stefan Institute in Ljubljana, Slovenia, is presented. Previous results have shown that subcriticality is not guaranteed for some postulated accidents (earthquake with subsequent fuel rack disintegration resulting in contact fuel pitch) under the assumption that the fuel rack is loaded with fresh 12 wt% standard fuel. To mitigate this deficiency, a study was done on replacing a certain number of fuel elements in the rack with cadmium-loaded absorber rods. The Monte Carlo computer code MCNP4A with an ENDF/B-V library and detailed three-dimensional geometrical model of the spent-fuel rack was used for this purpose. First, a minimum critical number of fuel elements was determined for contact pitch, and two possible geometries of rack disintegration were considered. Next, it was shown that subcriticality can be ensured when pitch is decreased from a rack design pitch of 8 cm to contact, if a certain number of fuel elements (8 to 20 out of 70) are replaced by absorber rods, which are uniformly mixed into the lattice. To account for the possibility that random mixing of fuel elements and absorber rods can occur during rack disintegration and result in a supercritical configuration, a probabilistic study was made to sample the probability density functions for random absorber rod lattice loadings. Results of the calculations show that reasonably low probabilities for supercriticality can be achieved (down to 10 -6 per severe earthquake, which would result in rack disintegration and subsequent maximum possible pitch decrease) even in the case where fresh 12 wt% standard TRIGA fuel would be stored in the spent-fuel pool

  11. The JCO criticality accident at Tokai-mura, Japan: an overview of the sampling campaign and preliminary results

    International Nuclear Information System (INIS)

    Komura, K.; Yamamoto, M.; Muroyama, T.; Murata, Y.; Nakanishi, T.; Hoshi, M.; Takada, J.; Ishikawa, M.; Takeoka, S.; Kitagawa, K.; Suga, S.; Endo, S.; Tosaki, N.; Mitsugashira, T.; Hara, M.; Hashimoto, T.; Takano, M.; Yanagawa, Y.; Tsuboi, T.; Ichimasa, M.; Ichimasa, Y.; Imura, H.; Sasajima, E.; Seki, R.; Saito, Y.; Kondo, M.; Kojima, S.; Muramatsu, Y.; Yoshida, S.; Shibata, S.; Yonehara, H.; Watanabe, Y.; Kimura, S.; Shiraishi, K.; Ban-nai, T.; Sahoo, S.K.; Igarashi, Y.; Aoyama, M.; Hirose, K.; Uehiro, T.; Doi, T.; Tanaka, A.; Matsuzawa, T.

    2000-01-01

    A criticality accident occurred on September 30, 1999 at the uranium conversion facility of the JCO Company Ltd. in Tokai-mura, Japan. A collaborating scientific investigation team was organized in two groups, the first to carry out research on the environmental impact (the environmental research group) and the second to assess the radiation effects on residents (the biological research group). This report concerns only the activities of the environmental research group. Four investigative teams were sent on different dates to the accident site and its vicinity to collect samples. About 400 samples were collected and subjected to analysis. An outline of the sampling campaign is presented here along with a brief chronology of the accident and the preliminary key results obtained by the independent research group are summarised in this Special Issue of the Journal of Environmental Radioactivity

  12. Design parameters and testing techniques for criticality accident detection systems used in various nuclear establishments - a review

    International Nuclear Information System (INIS)

    Janardhanan, S.; Krishnamony, S.; Krishnamurthi, T.N.; Gopalan, C.S.

    1981-01-01

    Accidental criticality excursion is a potential hazard in operations involving fissile material. In this review paper, design criteria for criticality detection systems, associated requirements for reliable functioning of the instrument and recent advances in the field are discussed. Systems based on integrated dose and rate of change of dose rate concepts are explained. A criticality accident simulator using a pneumatically driven 60 Co source for testing the detector is described. The paper also discusses the relative advantages of gamma and neutron sensing devices. (author)

  13. Design parameters and testing techniques for criticality accident detection systems used in various nuclear establishments - a review

    Energy Technology Data Exchange (ETDEWEB)

    Janardhanan, S.; Krishnamony, S.; Krishnamurthi, T.N.; Gopalan, C.S. (Bhabha Atomic Research Centre, Bombay (India). Health Physics Div.)

    Accidental criticality excursion is a potential hazard in operations involving fissile material. In this review paper, design criteria for criticality detection systems, associated requirements for reliable functioning of the instrument and recent advances in the field are discussed. Systems based on integrated dose and rate of change of dose rate concepts are explained. A criticality accident simulator using a pneumatically driven /sup 60/Co source for testing the detector is described. The paper also discusses the relative advantages of gamma and neutron sensing devices.

  14. Application of the accident consequences model of the German risk study to assessments of accident risks in different types of nuclear power plants

    International Nuclear Information System (INIS)

    Ehrhardt, J.; Bayer, A.

    1982-01-01

    Within the scope of the 'German Risk Study for Nuclear Power Plants' (Phase A) the accident consequence model UFOMOD was developed in the Karlsruhe Nuclear Research Center. This model originally developed for pressurized water reactors has now been extended in order to obtain results about accidental releases of activity from fast breeder and high-temperature reactors, too. (RW) [de

  15. NARCISS critical stand experiments for studying the nuclear safety in accident water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoj, N.N.; Glushkov, E.S.; Bubelev, V.G.

    2005-01-01

    A brief description of the Topaz-2 SNPS designed under scientific supervision of RRC KI in Russia, and of the NARCISS critical facility, is given. At the NARCISS critical facility, neutronic peculiarities and nuclear safety issues of the Topaz-2 system reactor were studied experimentally. This work is devoted to a detailed description of experiments on investigation of criticality safety in accident water immersion og highly enriched uranium dioxide fuel elements, performed at the NARCISS facility. The experiments were carried out at water-moderated critical assemblies with varying height, number, and spacing of fuel elements. The results obtained in the critical experiments, computational models of the investigated critical configurations, and comparison of the computational and experimental results are given [ru

  16. Study of Containment Vent Strategies During Severe Accident Progression for the CANDU6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Youngho; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    In March, 2011, Fukushima daichi nuclear power plants experienced a long term station blackout. Severe core damage occurred and a large amount of radioactive materials are released outside of the plants. After this terrible accident Nuclear Safety and Security Commission (NSSC) enforced to increase nuclear safety for all operating plants in Korea. To increase plant safety, both hardware reinforcement and software improvement are encouraged. Hardware reinforcement includes the preparation of the external water injection paths to the RCS and the spent fuel pool, a filtered containment venting system (CFVS), and AC power generating truck. Software improvement includes the increase of the effectiveness of the severe accident management guidance (SAMG) and plant staff training. To comply with NSSC's request, Wolsong Unit 1 has fulfilled the hardware reinforcement including the installation of a CFVS and started the extension of a SAMG to the low power and shutdown operation mode. Current SAMG deals accident occurred during full power operation only. The CFVS is designed to open and to close isolation valves manually. It does not require AC power. The operation of the CFVS prevents the reactor containment building failure due to the over-pressurization but it may release radioactive materials out of the reactor containment building. This paper discusses the radiological source terms for the containment vent strategy during severe accident progression which occurred during shutdown operation mode. This work is a part of the development of shutdown SAMG.. The CFVS is an effective means to control the containment pressure when the local air coolers are unavailable. Radioactive materials may release through the CFVS, but their amounts are reduced significantly. The alternative means, i.e., containment vent through the ventilation system which does not have an effective filter, is not a good choice to control the containment condition. It can maintain the containment

  17. Emergency response and nuclear risk governance. Nuclear safety at nuclear power plant accidents

    International Nuclear Information System (INIS)

    Kuhlen, Johannes

    2014-01-01

    The present study entitled ''Emergency Response and Nuclear Risk Governance: nuclear safety at nuclear power plant accidents'' deals with issues of the protection of the population and the environment against hazardous radiation (the hazards of nuclear energy) and the harmful effects of radioactivity during nuclear power plant accidents. The aim of this study is to contribute to both the identification and remediation of shortcomings and deficits in the management of severe nuclear accidents like those that occurred at Chernobyl in 1986 and at Fukushima in 2011 as well as to the improvement and harmonization of plans and measures taken on an international level in nuclear emergency management. This thesis is divided into a theoretical part and an empirical part. The theoretical part focuses on embedding the subject in a specifically global governance concept, which includes, as far as Nuclear Risk Governance is concerned, the global governance of nuclear risks. Due to their characteristic features the following governance concepts can be assigned to these risks: Nuclear Safety Governance is related to safety, Nuclear Security Governance to security and NonProliferation Governance to safeguards. The subject of investigation of the present study is as a special case of the Nuclear Safety Governance, the Nuclear Emergency governance, which refers to off-site emergency response. The global impact of nuclear accidents and the concepts of security, safety culture and residual risk are contemplated in this context. The findings (accident sequences, their consequences and implications) from the analyses of two reactor accidents prior to Fukushima (Three Mile Iceland in 1979, Chernobyl in 1986) are examined from a historical analytical perspective and the state of the Nuclear Emergency governance and international cooperation aimed at improving nuclear safety after Chernobyl is portrayed by discussing, among other topics, examples of &apos

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

    International Nuclear Information System (INIS)

    Sykora, D.; Sykorova, I.

    1988-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Fruchard, Y.; Lavie, J.M

    1966-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Fruchard, Y; Lavie, J M

    1966-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-15

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

  2. Simulation and dose analysis of a hypothetical accident in Sanmen nuclear power plant

    International Nuclear Information System (INIS)

    Zhu, Yangmo; Guo, Jianghua; Nie, Chu; Zhou, Youhua

    2014-01-01

    Highlights: • Atmospheric dispersion following a hypothetical accident in Sanmen NPP is simulated. • Japan, North Korea and Russia are slightly influenced in this accident. • In Taiwan and South Korea, population on 100% and 35% of the land should be given information about reducing dose. • In mainland China, about 284 thousand people are likely to get cancer. - Abstract: In November 2013, an AP1000 nuclear power plant (NPP) will be put into commercial operation. An atmospheric dispersion of radionuclides during a severe hypothetical accident in Sanmen NPP, Zhejiang province, China, is simulated with a Lagrangian particle dispersion model FLEXPART. The accident assumes that a station blackout (SBO) accident occurred on August 25, 2011, 55% core was damaged and 49 radionuclides were released into the atmosphere. Our simulation indicates that, during this dispersion, the radioactive plume will cover the mainland China, Taiwan, Japan, North Korea, South Korea and Russia. The radiation dose levels in Japan, North Korea and Russia are the lightest, usually less than 1 mSv. The influenced areas in these countries are 9901 km 2 , 31,736 km 2 and 2,97,524 km 2 , respectively; dose levels in Taiwan and South Korea are moderate, no more than 20 mSv. Information about reducing dose should be given to the public. Total influenced areas in these two countries are 3621 km 2 and 42,370 km 2 , which take up 100% of the land in Taiwan and 35% of the land in South Korea; the worst situation happens in mainland China. The total influenced area is 3 × 106 km 2 and 1,40,000 km 2 in this area has a dose level higher than 20 mSv. Measurement must be taken to reduce the dose. More than 284 thousand residents will face the risk of developing cancer. Furthermore, 96% of this population is mainly concentrated in Zhejiang province, where Sanmen NPP locates

  3. Are the sea foods on our table safe after the Fukushima Nuclear Power Plant Accident?

    International Nuclear Information System (INIS)

    Enriquez, E.B.; Palad, L.J.H.; Encabo, R.R.; Cruz, P.T.F.; Garcia, T.Y.

    2015-01-01

    The Fukushima Nuclear Power Plant Accident that occurred in March 2011 raised immediate public concern on the safety of marine organisms caught in the Philippine waters. This is because of Japan’s proximity to the Philippine archipelago and the threat of contamination reaching our shores became a major issue. Immediately after the accident, the Philippine Nuclear Research Institute (PNRI) established an extensive marine monitoring program to assess any possible effect of one of the worst nuclear accidents that occurred in recent time. This study is under the framework of the project entitled “Radiological Impact Assessment of the Fukushima Nuclear Accident in the Philippine Marine Environment”. Species of fish, mollusks and crustaceans were analyzed for anthropogenic radionuclides cesium-134 (Cs-134) and Cesium-137 (Cs-137) Using a high purify Germanium (HPGe) detector from ORTEC Inc. Major fishing grounds and coastal areas in the west and eastern seaboards as well as in the north and south were selected as sources of biota samples. Commonly eaten and popular species of fish such as tuna, mackerel, sardines, etc were purchased from local markets in the area; processed and analyzed by gamma spectrometry. The results showed that the average Cs-137 activity concentration in fish samples (n=103) was found to be 0.74 ± 0.28 Becquerel/kilogram wet. The Cs-137 concentrations in mollusks (n=12) and crustaceans (n=4) were all below the Lowest Limit of Detection (LLD). Cs-134 was not detected in any of the samples analyzed. The low concentration of the radionuclides studied showed that, thus far, the Fukushima NPP accident has no Impact to the Philippines marine environment. (author)

  4. Short and medium effects on the environment of Valencia, Spain of the Chernobyl nuclear plant accident

    International Nuclear Information System (INIS)

    Moreno, A.; Navarro, E.; Senent, F.; Baeza, A.; Miro, C.; Rio, M. del

    1991-01-01

    As a consequence of the 26 April 1986 accident at the Chernobyl nuclear plant, a large amount of radioactivity was released into the atmosphere. The radioactive plume formed could be detected in practically the whole of the Northern Hemisphere a few days later. The zone most affected by the radioactive cloud over Spain was that of the Mediterranean coast and the Balearic Islands. In this paper, the authors examine the level of the radioactive contamination reached in various receptive media in Valencia, such as air, dry-fallout, water, soil, grass and milk samples collected in Valencia immediately after the accident. The activity levels are compared with those found during 1964 and 1965 due to the Chinese nuclear atmospheric explosions. The levels of contamination presented by four species of migratory birds which spend the winter in this area is analyzed. Lastly, an estimate is made of the absorbed dose

  5. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks.

  6. Development of resilience evaluation method for nuclear power plant. Part 1. Proposal of resilience index for assessment of safety of nuclear power plant under severe accident

    International Nuclear Information System (INIS)

    Demachi, Kazuyuki; Suzuki, Masaaki; Itoi, Tatsuya

    2016-01-01

    In this research, a new index 'The Resilience Index' was proposed to evaluate the capability of nuclear power plant to recover from the situation of safety function lost. Three elements assumed to evaluate the resilience index are the achievement rate, necessary time, and probability of success of each accident management activity. The resilience index is expected to visualize the improvement of safety of each nuclear power plant against severe accidents. (author)

  7. Basic study on BWR plant behavior under the condition of severe accident (2)

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Ueda, Masataka; Sasaki, Hajime

    2016-01-01

    In this paper, we report on the results using the BWR plant simulator about the plant behavior under the condition of the two types of severe accidents that LOCA occurs but ECCS fails the water irrigation into the reactor core and SBO occurs and at the same time the reclosed failure of SRV occurs. The simulation experiments were carried out for the cases that LOCA has occurred in the main feed-water piping. As for the results about the relationship between the LOCA area and the time from LOCA occurs until the fuel temperature rise start, the effect that RCIC operated was extremely big for small and middle LOCA area. In the case of main feed-water system LOCA, the core water level suddenly decreased for large LOCA of 2000 cm"2 area, however, if the irrigation into the reactor core was carried out 30 min after LOCA occurrence, the core had little damage. In addition, the H_2 concentration in the containment vessel did not exceed both limits of H_2 explosion nor detonation. The pressure of the containment vessel was around 3 kg/cm"2 of design value, so the soundness of the containment vessel was confirmed. On the other hand, for the accident of SBO with reclosed failure of SRV, it has been shown that the accidents continue to progress rapidly as compared with the case of normally operating of SRV. Because SRV has the function that keep the inside pressure of reactor core by repeating opened and closed in response of the inside pressure and prevent the decrease of water level inside reactor core. However, if the irrigation into the reactor core was carried out 30 min after SBO occurrence, the core had little damage and also the H_2 concentration in the containment vessel did not exceed limits of H_2 explosion. Further, as for the accident of reclosed failure of SRV, it has been shown that there are very good correspondence with the simulation results of main steam piping LOCA of area 180 cm"2 corresponding to the inlet cross-sectional area SRV installed on the piping

  8. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  9. TL detectors for gamma-ray dose measurements in critically accidents

    International Nuclear Information System (INIS)

    Miljanic, S.; Knezevic, Z.; Zorko, B.; Gregori, B.

    2005-01-01

    Full text: Determination of gamma-ray dose in mixed neutron + gamma-ray fields is still a challenging task. Dosemeters used for gamma-ray dosimetry are usually in some extent sensitive to neutrons and their response variations depend on neutron energy i.e. on neutron spectra. Besides, it is necessary to take into account the energy dependence of dosimeter responses to gamma-rays. To reduce all these influences, design of dosemeter holders is of special importance. In this work, several types of thermoluminescent detectors (TLD) placed in different holders used for gamma-ray dose determination in mixed fields were examined. Dosemeters were from three different institutions: Ruder Boscovic Institute (RBI), Croatia, Jozef Stefan Institute (JSI), Slovenia and Autoridad Regulatoria Nuclear (ARN), Argentina. All dosemeters were irradiated during the International Intercomparison of Criticality Accident Dosimetry Systems at the SILENE Reactor, Valduc, June 2002. At that exercise three accidental scenarios were reproduced: bare reactor, free evolution; lead shielded reactor, steady state; and lead shielded reactor, free evolution. In each irradiation dosemeters were exposed placed on the front of phantom and 'free-in-air'. Also, dosemeters were irradiated in a pure gamma ray field of 60 Co source. Following types of TLDs were used: 7 LiF (TLD-700), CaF 2 :Mn and AI 2 O 3 :Mg,Y - all from RBI; CaF 2 :Mn from JSI and 7 LiF (TLD-700) from ARN. Reported doses were compared with the reference values as well as with the mean participants' values. The results show satisfactory agreement with other dosimetry systems used in the Intercomparison. The influence of different types of holders and applied corrections of dosemeters' readings are discussed. (author)

  10. Critical heat flux for APR1400 lower head vessel during a severe accident

    International Nuclear Information System (INIS)

    Noh, Sang W.; Suh, Kune Y.

    2013-01-01

    Highlights: ► Studied boiling on downward-facing hemispherical vessel with asymmetric thermal insulator. ► Scaled the APR1400 lower head linearly down by 1/10 including ICI tubes and shear keys. ► Performed thermal analysis using ANSYS V11.0 to determine the internal temperature and heat flux. ► Performed tests to obtain the CHF with saturated demineralized water at atmospheric pressure. ► Measured CHF accounting for 3D random flow effect expected in the APR1400 application. -- Abstract: Corium Ablation Stopper Apparatus (CASA) has a downward-facing hemispherical vessel and geometrically asymmetric thermal insulator of the Advanced Power Reactor 1400 MWe (APR1400) scaled linearly down by 1/10, as well as sixty-one in-core instrumentation (ICI) tubes and four shear keys. The heated vessel plays a pivotal role in CASA depending on the configuration of the oxide pool and metal layer to bring about the focusing effect expected of a molten pool in the lower head during a severe accident. The heated vessel was designed through a trial-and-error method and thermal analysis. Thermal analysis was performed using ANSYS V11.0 to investigate the effect of the internal temperature and heat flux on the integral hemispherical copper vessel. The CASA tests were carried out to obtain the critical heat flux (CHF) with saturated and demineralized water at the atmospheric pressure (0.1 MPa). The CHF in the metal layer through the hemispherical channel was found to be lower than that in the ULPU-2400 configuration V data through the streamlined thermal insulator. The experimental CHF was measured and obtained through the CASA hemispherical heated surface accounting for the three-dimensional random flow effect expected in the APR1400 application

  11. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    International Nuclear Information System (INIS)

    Purcell, P.C.; Dallongeville, M.

    2004-01-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

  12. Optimization of the Severe Accident Management Strategy for Domestic Plants and Validation Experiments

    International Nuclear Information System (INIS)

    Kim, S. B.; Kim, H. D.; Koo, K. M.; Park, R. J.; Hong, S. H.; Cho, Y. R.; Kim, J. T.; Ha, K. S.; Kang, K. H.

    2007-04-01

    nuclear power plants, a technical basis report and computational aid tools were developed in parallel with the experimental and analytical works for the resolution of the uncertain safety issues. ELIAS experiments were carried out to quantify the boiling heat removal rate at the upper surface of a metallic layer for precise evaluations on the effect of a late in-vessel coolant injection. T-HERMES experiments were performed to examine the two-phase natural circulation phenomena through the gap between the reactor vessel and the insulator in the APR1400. Detailed analyses on the hydrogen control in the APR1400 containment were performed focused on the effect of spray system actuation on the hydrogen burning and the evaluation of the hydrogen behavior in the IRWST. To develop the technical basis report for the severe accident management, analyses using SCDAP/RELAP5 code were performed for the accident sequences of the OPR1000. Based on the experimental and analytical results performed in this study, the computational aids for the evaluations of hydrogen flammability in the containment, criteria of the in-vessel corium cooling, criteria of the external reactor vessel cooling were developed. An ASSA code was developed to validate the signal from the instrumentations during the severe accidents and to process the abnormal signal. Since ASSA can perform the signal processing from the direct input of the nuclear power plant during the severe accident, it can be platform of the computational aids. In this study, the ASSA was linked with the computaional aids for the hydrogen flammability

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-05-21

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

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

    Science.gov (United States)

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

    2001-05-01

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

  15. Comprehensive Health Risk Management after the Fukushima Nuclear Power Plant Accident.

    Science.gov (United States)

    Yamashita, S

    2016-04-01

    Five years have passed since the Great East Japan Earthquake and the subsequent Fukushima Daiichi Nuclear Power Plant accident on 11 March 2011. Countermeasures aimed at human protection during the emergency period, including evacuation, sheltering and control of the food chain were implemented in a timely manner by the Japanese Government. However, there is an apparent need for improvement, especially in the areas of nuclear safety and protection, and also in the management of radiation health risk during and even after the accident. Continuous monitoring and characterisation of the levels of radioactivity in the environment and foods in Fukushima are now essential for obtaining informed consent to the decisions on living in the radio-contaminated areas and also on returning back to the evacuated areas once re-entry is allowed; it is also important to carry out a realistic assessment of the radiation doses on the basis of measurements. Until now, various types of radiation health risk management projects and research have been implemented in Fukushima, among which the Fukushima Health Management Survey is the largest health monitoring project. It includes the Basic Survey for the estimation of external radiation doses received during the first 4 months after the accident and four detailed surveys: thyroid ultrasound examination, comprehensive health check-up, mental health and lifestyle survey, and survey on pregnant women and nursing mothers, with the aim to prospectively take care of the health of all the residents of Fukushima Prefecture for a long time. In particular, among evacuees of the Fukushima Nuclear Power Plant accident, concern about radiation risk is associated with psychological stresses. Here, ongoing health risk management will be reviewed, focusing on the difficult challenge of post-disaster recovery and resilience in Fukushima. Copyright © 2016 The Royal College of Radiologists. Published by Elsevier Ltd. All rights reserved.

  16. Optimization of the Severe Accident Management Strategy for Domestic Plants and Validation Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. B.; Kim, H. D.; Koo, K. M.; Park, R. J.; Hong, S. H.; Cho, Y. R.; Kim, J. T.; Ha, K. S.; Kang, K. H

    2007-04-15

    nuclear power plants, a technical basis report and computational aid tools were developed in parallel with the experimental and analytical works for the resolution of the uncertain safety issues. ELIAS experiments were carried out to quantify the boiling heat removal rate at the upper surface of a metallic layer for precise evaluations on the effect of a late in-vessel coolant injection. T-HERMES experiments were performed to examine the two-phase natural circulation phenomena through the gap between the reactor vessel and the insulator in the APR1400. Detailed analyses on the hydrogen control in the APR1400 containment were performed focused on the effect of spray system actuation on the hydrogen burning and the evaluation of the hydrogen behavior in the IRWST. To develop the technical basis report for the severe accident management, analyses using SCDAP/RELAP5 code were performed for the accident sequences of the OPR1000. Based on the experimental and analytical results performed in this study, the computational aids for the evaluations of hydrogen flammability in the containment, criteria of the in-vessel corium cooling, criteria of the external reactor vessel cooling were developed. An ASSA code was developed to validate the signal from the instrumentations during the severe accidents and to process the abnormal signal. Since ASSA can perform the signal processing from the direct input of the nuclear power plant during the severe accident, it can be platform of the computational aids. In this study, the ASSA was linked with the computaional aids for the hydrogen flammability.

  17. Transcriptional plant responses critical for resistance towards necrotrophic pathogens

    Directory of Open Access Journals (Sweden)

    Rainer P. Birkenbihl

    2011-11-01

    Full Text Available Plant defenses aimed at necrotrophic pathogens appear to be genetically complex. Despite the apparent lack of a specific recognition of such necrotrophs by products of major R genes, biochemical, molecular, and genetic studies, in particular using the model plant Arabidopsis, have uncovered numerous host components critical for the outcome of such interactions. Although the JA signaling pathway plays a central role in plant defense towards necrotrophs additional signaling pathways contribute to the plant response network. Transcriptional reprogramming is a vital part of the host defense machinery and several key regulators have recently been identified. Some of these transcription factors positively affect plant resistance whereas others play a role in enhancing host susceptibility towards these phytopathogens.

  18. A human reliability analysis of the Three Mile power plant accident considering the THERP and ATHEANA methodologies

    International Nuclear Information System (INIS)

    Fonseca, Renato Alves da

    2004-03-01

    The main purpose of this work is the study of human reliability using the THERP (Technique for Human Error Prediction) and ATHEANA methods (A Technique for Human Error Analysis), and some tables and also, from case studies presented on the THERP Handbook to develop a qualitative and quantitative study of nuclear power plant accident. This accident occurred in the TMI (Three Mile Island Unit 2) power plant, PWR type plant, on March 28th, 1979. The accident analysis has revealed a series of incorrect actions, which resulted in the Unit 2 shut down and permanent loss of the reactor. This study also aims at enhancing the understanding of the THERP method and ATHEANA, and of its practical applications. In addition, it is possible to understand the influence of plant operational status on human failures and of these on equipment of a system, in this case, a nuclear power plant. (author)

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

    International Nuclear Information System (INIS)

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

    1978-04-01

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

  20. Strategy generation in accident management support

    International Nuclear Information System (INIS)

    Sirola, M.

    1995-01-01

    An increased interest for research in the field of Accident Management can be noted. Several international programmes have been started in order to be able to understand the basic physical and chemical phenomena in accident conditions. A feasibility study has shown that it would be possible to design and develop a computerized support system for plant staff in accident situations. To achieve this goal the Halden Project has initiated a research programme on Computerized Accident Management Support (CAMS project). The aim is to utilize the capabilities of computerized tools to support the plant staff during the various accident stages. The system will include identification of the accident state, assessment of the future development of the accident and planning of accident mitigation strategies. A prototype is developed to support operators and the Technical Support Centre in decision making during serious accident in nuclear power plants. A rule based system has been built to take care of the strategy generation. This system assists plant personnel in planning control proposals and mitigation strategies from normal operation to severe accident conditions. The ideal of a safety objective tree and knowledge from the emergency procedures have been used. Future prediction requires good state identification of the plant status and some knowledge about the history of some critical variables. The information needs to be validated as well. Accurate calculations in simulators and a large database including all important information form the plant will help the strategy planning. (author). 12 refs, 2 figs

  1. German (GRS) approach to accident analysis (part I). German licensing basis for accident analyses. Applicants accident analyses in second part license for Konvoi-plants. Appendix 1. Assessor accident analyses in second part license for Konvoi-plants. Appendix 2. Reference list of DBA to be considered in the safety status analysis of a PSR. Appendix 3a. Reference list of special very rare and BDB plant conditions to be considered in the safety status analysis of a PSE. Appendix 3b

    International Nuclear Information System (INIS)

    Velkov, K.

    2002-01-01

    Appendix 1: The Safety Analysis Report (S.A.R.) is presented from 3 Handbooks - ECC Handbook (LOCA), Plant Dynamics Handbook (Transients incl. ATWS), and Core Design Handbook. The first one Conceived as Living handbook, Basis for design, catalogue of transients, specifications and licensing. Handbook contains LOCA in primary system, it contains also core damage analysis, and description of codes, description of essential plant data and code input data. The second one consists of Basis for design, commissioning, operation, and catalogue of transients, specifications and licensing, as well as specified operation, disturbed operation, incidents, non-LOCA, SS-procedures and Code description. The third book consists of Reactivity balance and reactivity coefficients, efficiency of shutdown systems. Calculation of burn up cycle, power density distribution, and critical boron concentration. Also Codes used, as SAV79A standard analysis methodology including FASER for nuclear data generation, MEDIUM and PANBOX for static and transient core calculations. Appendix 2: The three TUEV (Technical Inspection Agencies) responsible for the three individual plants of type KONVOI: TUEV Bayern for ISAR-2, TUV-Hanover for KKE, TUEV-Stuttgart for GKN-2 and GRS performed the safety assessment. TUV-Bayern for disturbance and failure of secondary heat sink without loss of coolant (failure of main heat sink, erroneous operation of valves in MS and in FW system, failure of MFW supply), long term LONOP, performance of selected SBLOCA analyses. TUV Hanover for disturbances due to failure of MCPs, short term LONOP, damages of SG tubes incl. SGTR, performance of selected LOCA analyses (blowdown phase of LBLOCA). TUV-Stuttgart for breaks and leaks in MS and FW system with and without leaks in SG tubes. GRS for ATWS, sub-cooling transients due to disturbances on secondary side, initial and boundary conditions for transients with opening of pressurizer valves with and without stuck-open, most of the

  2. Safety-critical human factors issues derived from analysis of the TEPCO Fukushima Daiichi accident investigation reports

    International Nuclear Information System (INIS)

    Sakuda, Hiroshi; Takeuchi, Michiru

    2013-01-01

    The Fukushima Daiichi nuclear power plant accident on March 11, 2011 had a large impact both in and outside Japan, and is not yet concluded. After Tokyo Electric Power Co.'s (TEPCO's) Fukushima accident, electric power suppliers have taken measures to respond in the event that the same state of emergency occurs - deploying mobile generators, temporary pumps and hoses, and training employees in the use of this equipment. However, it is not only the “hard” problems including the design of equipment, but the “soft” problems such as organization and safety culture that have been highlighted as key contributors in this accident. Although a number of organizations have undertaken factor analysis of the accident and proposed issues to be reviewed and measures to be taken, a systematic overview about electric power suppliers' organization and safety culture has not yet been undertaken. This study is based on three major reports: the report by the national Diet of Japan Fukushima Nuclear Accident Independent Investigation Commission (the Diet report), the report by the Investigation Committee on the Accident at Fukushima Nuclear Power Stations of Tokyo Electric Power Company (Government report), and the report by the non-government committee supported by the Rebuild Japan Initiative Foundation (Non-government report). From these reports, the sections relevant to electric power suppliers' organization and safety culture were extracted. These sections were arranged to correspond with the prerequisites for the ideal organization, and 30 issues to be reviewed by electric power suppliers were extracted using brainstorming methods. It is expected that the identified issues will become a reference for every organization concerned to work on preventive measures hereafter. (author)

  3. Implementation of hydrogen mitigation techniques during severe accidents in nuclear power plants

    International Nuclear Information System (INIS)

    1996-01-01

    concentration and under special geometric conditions, an accelerated flame or even a local detonation may occur which would produce higher dynamic loads than a deflagration and a more serious threat to equipment and structures. Should it occur in spite of its low probability, a global detonation, following prolonged and extensive accumulation of hydrogen in the containment atmosphere, would be a major threat to the containment integrity. The goal of hydrogen mitigation techniques is to prevent loads, resulting from hydrogen combustion, which could threaten containment integrity. The risk of containment failure depends on the overall hydrogen concentration which is dependent on the amount of hydrogen released and the containment volume. A possible containment failure also depends on the containment structure and design which is very important in the resistance of the containment to a global combustion. Geometrical sub-compartmentalization is also very important, because significant amounts of hydrogen could accumulate in compartments to create high local concentrations of hydrogen that could be well within the detonability limits. Once accident management measures aimed at preventing severe accidents from occurring have failed and hydrogen is being generated and released to the containment atmosphere in large amounts, the first step is to reduce the possibility of hydrogen accumulating to flammable concentrations. Where flammable concentrations cannot be precluded, the next step is to minimize the volume of gas at flammable concentrations and the third and last step is to prevent further increasing hydrogen levels from the flammable to detonable mixture concentrations. The purpose of this paper is to present a snapshot, from a technical viewpoint, of the current situation regarding the implementation of hydrogen mitigation techniques for severe accident conditions in nuclear power plants. Broader aspects related to overall accident management policies are not considered here

  4. Phenix plant - Complementary safety assessment of the Phenix plant (INB 71) in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Phenix reactor to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. The Phenix reactor stands on the Marcoule site of CEA and was stopped definitely in 2009 for electricity production. Robustness is the ability for the facility to withstand events beyond the level for which the facility was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence (cliff edge effect). Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like crisis organization and work organization via subcontracting are also taken into consideration. This report is divided into 9 main chapters: 1) main features of the Phenix facility, 2) identification of cliff edge risks as well as structures and equipment essential to safety, 3) earthquake risk, 4) flood risk, 5) risks due to other extreme natural disasters, 6) the loss of electrical power supplies and of cooling systems, 7) management of severe accidents, 8) subcontracting policy, 9) synthesis. This study shows that it is necessary to take some measures to reinforce the robustness of the plant concerning flood risks. (A.C.)

  5. Integrated framework for the external cost assessment of nuclear power plant accident considering risk aversion: The Korean case

    International Nuclear Information System (INIS)

    Lee, Sang Hun; Kang, Hyun Gook

    2016-01-01

    Recently, the estimation of accident costs within the social costs of nuclear power plants (NPPs) has garnered substantial interest. In particular, the risk aversion behavior of the public toward an NPP accident is considered an important factor when integrating risk aversion into NPP accident cost. In this study, an integrated framework for the external cost assessment of NPP accident that measures the value of statistical life (VSL) and the relative risk aversion (RRA) coefficient for NPP accident based on an individual-level survey is proposed. To derive the willingness to pay and the RRA coefficient for NPP accident risks, a survey was conducted on a sample of 1550 individuals in Korea. The estimation obtained a mean VSL of USD 2.78 million and an RRA coefficient of 1.315. Based on the estimation results in which various cost factors were considered, a multiplication factor of 5.16 and an external cost of NPP accidents of 4.39E−03 USD-cents/kW h were estimated. This study is expected to provide insight to energy policy decision-makers on analyzing the economic validity of NPP compared to other energy sources by reflecting the estimated external cost of NPP accident into the unit electricity generation cost of NPP. - Highlights: •External cost assessment framework for NPP is proposed considering risk aversion. •VSL was derived from WTP for mortality risk reduction from hypothetical NPP accident. •RRA was derived to integrate public risk aversion into external cost of NPP accident. •Individual-level survey was conducted to derive WTP and RRA for NPP accident risk. •The external cost was estimated considering the direct cost factors of NPP accident.

  6. Cytogenetic characteristics of children who suffered as a result of the Chernobyl power plant accident

    International Nuclear Information System (INIS)

    Vorobtsova, I.E.; Kolyubaeva, S.N.; Vorob'eva, M.V.; Korotkov, D.V.; Komar, V.E.

    1993-01-01

    Cytogenetic examinations of children living in the St. Peterburg district who suffered as a result of the Chernobyl power plant accident (liquidators' children and children evacuated from radioactive pollution areas) and of control children were carried out. The chromosomal system stability was assessed by several parameters: spontaneous levels of chromosomal radiosensitivity in in vitro irradiation of lymphocytes in dose 1.5 Gy of 60 Co γ-irradiation. Children with an increased incidence of chromosomal aberrations and an increased chromosomal radiosensitivity were referred to a risk group

  7. Hemostatic homeostasis in liquidators of the aftereffects of the Chernobyl power plant accident

    International Nuclear Information System (INIS)

    Chekalina, S.I.; Lyasko, L.I.; Sushkevich, G.N.; Pashkov, E.I.; Savina, H.P.

    1995-01-01

    The function of the hemostasis system was examined in 128 participants in the liquidation of the aftereffects of the Chernobyl power plant accident 4 years, on an average, after their work in the radioactive zone of the 4th energy block. Signs of functional disorganization in the hemostasis system wer revealed: hemocoagulation and platelet aggregation activation in the presence of reduced fibrinolysis activity and antithrombogenic properties of vascular walls. The said trends were best of all detected by functional loading (local circulatory hypoxia) of the vascular wall. 11 refs

  8. Estimation of water pollution by domestic in-land nuclear power plant under severe accident

    International Nuclear Information System (INIS)

    Li Hong; Fang Sheng; Fang Dong

    2013-01-01

    In-land nuclear power plant sites of China are usually located in densely populated area and are close to large surface water. This paper proposed scenarios and corresponding calculation models for water contamination caused by radioactive plume release after a severe accident. The models were applied to an imaginary lake (reservoir)-adjacent site in the south of China. The results showed that, the short-time concentration of radioactivity in the lake due to dry and wet deposition and runoff was higher than the generic action levels for foodstuffs in GB 18871-2002, and the public dose resulted was unacceptable. (authors)

  9. Safety Implementation of Hydrogen Igniters and Recombiners for Nuclear Power Plant Severe Accident Management

    Institute of Scientific and Technical Information of China (English)

    XIAO Jianjun; ZHOU Zhiwei; JING Xingqing

    2006-01-01

    Hydrogen combustion in a nuclear power plant containment building may threaten the integrity of the containment. Hydrogen recombiners and igniters are two methods to reduce hydrogen levels in containment buildings during severe accidents. The purpose of this paper is to evaluate the safety implementation of hydrogen igniters and recombiners. This paper analyzes the risk of deliberate hydrogen ignition and investigates three mitigation measures using igniters only, hydrogen recombiners only or a combination of recombiners and igniters. The results indicate that steam can effectively control the hydrogen flame acceleration and the deflagration-to-detonation transition.

  10. Break spectrum analyses for small break loss of coolant accidents in a RESAR-3S Plant

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Kullberg, C.M.

    1986-03-01

    A series of thermal-hydraulic analyses were performed to investigate phenomena occurring during small break loss-of-coolant-accident (LOCA) sequences in a RESAR-3S pressurized water reactor. The analysis included simulations of plant behavior using the TRAC-PF1 and RELAP5/MOD2 computer codes. Series of calculations were performed using both codes for different break sizes. The analyses presented here also served an audit function in that the results shown here were used by the US Nuclear Regulatory Commission (NRC) as an independent confirmation of similar analyses performed by Westinghouse Electric Company using another computer code. 10 refs., 62 figs., 14 tabs

  11. Medical countermeasure for Tokyo Electric Power Co. Fukushima Nuclear Power Plant accident

    International Nuclear Information System (INIS)

    Kondo, Hisayoshi

    2013-01-01

    DMAT (Disaster Medical Assistance Team) is a group of professional medical personnel organized to provide rapid-response medical care at the emergent stage of disasters. At the accident of Fukushima Daiichi Nuclear Power Plant, medical response was difficult because many infrastructures were destroyed. Under this situation, emergent medical treatment for heavy irradiation or contamination, cares for habitants and transportation of patients were conducted. Through these activities, it is suggested that rapid response for the radiation exposure should be definitely include in the medical system for usual disasters. (J.P.N.)

  12. The influence of Goiania radiological accident on Brazilian public opinion concerning new nuclear electric plants

    International Nuclear Information System (INIS)

    Meldonian, Nelson Leon; Mattos, Luis Antonio Terribile de

    1997-01-01

    The Brazilian society is against applications of nuclear energy, mainly respecting to construction of new nuclear power plants, believing that they are harmful to population's welfare and the environment. By this reason, Brazilian nuclear sector would promote a more intensive program of public discussion, not limited to technical and scientific community. Intending to contribute to a better judgment by society about the differences between diverse employment of nuclear energy, arguments concerned to its benefits are presented, pointing out that adverse accounts to nuclear electricity based on Goiania radiological accident, are not justified

  13. Technical feasibility and costs of the retention of radionuclides during accidents in nuclear power plants demonstrated by the example of a pressurized water reactor

    International Nuclear Information System (INIS)

    Braun, H.; Grigull, R.; Lahner, K.; Gutowski, H.; Weber, J.

    1985-01-01

    The maximum allowable radiation doses during accidents in nuclear power plants, i.e., 5 rem whole-body dose and 15 rem thyroid dose, have been laid down in the German Radiation Protection Act. In order to ensure that these limits are not exceeded for all exposure paths including the ingestion path or, if possible, to remain far below them, the Federal Ministry of the Interior has initiated a study on the effectiveness and cost of additional safety features for reducing the release of activity and the dose exposure during accidents in nuclear power plants. Detailed investigations were carried out for the following three radiologically representative types of accidents: break of a reactor coolant line, break of an instrument line in one of the outer ring rooms, and break of a main stream line outside the containment. The technical basis of the study was a BBR-type nuclear power plant with pressurized water reactor and once-through steam generator. I-131 was chosen for determining the activity release as this is the critical nuclide for the ingestion path. Altogether 33 feasible technical measures were investigated and their potential improvement was assessed

  14. The accident of the Fukushima-Daiichi nuclear plant. Status two years after the event

    International Nuclear Information System (INIS)

    2013-03-01

    In a first part, this report briefly recalls the circumstances and occurrence of the accident, gives an overview of actions undertaken by the IRSN (calculations of installation damages, modelling of contaminated air movements, simulations of radionuclide dispersion in the sea environment, information of French nationals in Japan, press and public information), and an overview of strength tests of nuclear installations (additional safety assessments and European stress tests). The second part gives an overview of the situation in Japan two years after the accident: evolution of governance in terms of nuclear risk management, condition of the Fukushima plant in January 2013, health and environmental impact and post-accidental management, actions undertaken by the IRSN (assessment of doses potentially received by populations, strengthening of cooperation between Japan and France in the field of severe accidents, participation to the Fukushima Dialogue). The third part presents the contribution of the IRSN to the strengthening of nuclear safety and radiation protection at the international level, at the European level, and in France

  15. Accident source terms for Light-Water Nuclear Power Plants. Final report

    International Nuclear Information System (INIS)

    Soffer, L.; Burson, S.B.; Ferrell, C.M.; Lee, R.Y.; Ridgely, J.N.

    1995-02-01

    In 1962 tile US Atomic Energy Commission published TID-14844, ''Calculation of Distance Factors for Power and Test Reactors'' which specified a release of fission products from the core to the reactor containment for a postulated accident involving ''substantial meltdown of the core''. This ''source term'', tile basis for tile NRC's Regulatory Guides 1.3 and 1.4, has been used to determine compliance with tile NRC's reactor site criteria, 10 CFR Part 100, and to evaluate other important plant performance requirements. During the past 30 years substantial additional information on fission product releases has been developed based on significant severe accident research. This document utilizes this research by providing more realistic estimates of the ''source term'' release into containment, in terms of timing, nuclide types, quantities and chemical form, given a severe core-melt accident. This revised ''source term'' is to be applied to the design of future light water reactors (LWRs). Current LWR licensees may voluntarily propose applications based upon it

  16. Evaluation of hazards from industrial activity near nuclear power plants. Study of typical accidents

    International Nuclear Information System (INIS)

    Lannoy, A.; Gobert, T.; Granier, J.P.

    1981-08-01

    The design and dimensioning of nuclear power plant structures necessitate the evaluation of risks due to industrial activity. Among these risks, those due to the storage or transport of dangerous products merit special attention. They result, inter alia, in the explosion of flammable gas clouds. Such clouds can drift before igniting and, once alight, the resulting pressure wave can cause serious damage, even at a distance. A methodology both deterministic and probabilistic enabling this risk to be quantified has therefore been developed. It is based in part on an analysis of the statistics of actual accidents that have occurred. After briefly recalling the probabilistic model, the typical accidents selected are described and for three usual cases (storage under pressure, rail tank cars and road units) the main characteristics of the rupture are explicited. The deterministic models that have been worked out to calculate the consequences of such an accident: flow rate at the bursting point, evaporation, drift and atmospheric dispersion of the cloud formed, explosion of this cloud, are then described. At the present time the overpressure wave is quantified against a TNT equivalent of the explosive mixture. Some data are given as examples for three commonly employed hydrocarbons (butane, propane, propylene) [fr

  17. In-Plant Fission Product Behavior in SGTR Accident with Long-Term SBO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Woon; Han, Seok Jung; Ahn, Kwang Il [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The off-site AC power was recovered in 9 days after the accident in the NPS. Therefore safety injection by fire pump truck with fresh water or seawater is only available in the Fukushima accident. However, safety injection by fire pump truck is not always effective due to the high pressure of RPV inside or leakages of alternative water injection flow paths. In the SBO situations in pressurized water reactor plant (PWR), turbine driven auxiliary feedwater (TD-AFW) pump can inject water to the secondary side of steam generator. However, turbine inlet steam flow control valve cannot work properly when loss of vital DC power occurs. Vital DC power is designed to be maintained during 4 or 8 hours in the SBO conditions. In this paper motor-driven and turbine driven AFW pumps are all assumed to be not working at time 0 sec as a worst case assumption. Iodine pool-scrubbing can occur in the secondary side of the faulted steam generator. However, iodine pool-scrubbing in the secondary side of the faulted steam generator is assumed not to be working, due to the assumption of the loss of DC battery for turbine inlet flow control valve. Iodine pool-scrubbing is one of the long-term research issues in safety assessment of nuclear power plant severe accident. PHEBUS FPT series and THAI experiment projects are typical projects on the resolving source term issues in severe accident of nuclear power plants. However, iodine retention by pool scrubbing is still a debating issue. In such containment bypass sequences, fission products can be released out to environment directly from RCS without retention or deposition in containment structures. SGTR is one of the hazardous accident scenarios in the typical PSA, because SGTR induces a large release amount of source term to environment directly. A key operation strategy was the isolation of the broken reactor coolant system loop from the intact loop. Typical core degradation in SGTR scenarios occurs with multiple failures of the isolation

  18. Dose estimation and evaluation of protector measures for a power plant's accidents scenario, using geographical information system

    International Nuclear Information System (INIS)

    Costa, E.M.; Biagio, R.M.S.; Alves, R.N.

    1999-01-01

    Since the initial phase of a project of a nuclear plant several environmental studies are carried out, and a considerable amount of relevant information is generated. Therefore, there is an increasing need of an integrated analysis of this information in order to better evaluate the potential impact associated to hypothetical accident scenarios of such plants. This paper presents a case-study, in which a hypothetical accident scenario is analysed taking into account the environmental and populational information of the Brazilian nuclear power plants region by using a geographical information system. Important areas for planning of protective measures are identified to provide a basis for further analysis. (author)

  19. Socioeconomic consequences of nuclear reactor accidents

    International Nuclear Information System (INIS)

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

    1984-06-01

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

  20. Accident information needs

    International Nuclear Information System (INIS)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information

  1. Accident information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-12-31

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

  2. Accident information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

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

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  4. Report of a Special Committee on the Review of U.S. Nuclear Power Plant Accident, second report

    International Nuclear Information System (INIS)

    1979-01-01

    Following on the issuance of the first report, for the accident in Three Mile Island Nuclear Power Plant in the United States there has appeared detailed information of such as reactor operation and radiation control. This has enabled technical evaluation of those items involved in nuclear power safety. The review results up to the beginning of September 1979 are presented, to meet popular desires to know the accident situation and to reflect the results in the nation's nuclear power generation. Contents are features and background of the TMI Nuclear Power Plant accident consequences, safety measures to be taken in Japan, and (in the appendix) the data on the TMI accident, countermeasures taken in Japan, etc. (Mori, K.)

  5. Hand-calculation technique for the evaluation of public risk from a severe accident at a nuclear power plant

    International Nuclear Information System (INIS)

    Linn, M.A.; Schmoyer, R.E.

    1993-01-01

    The Nuclear Regulatory Commission (NRC) is in the process of promulgating a proposed rule 10 CFR Part 54, ''Requirements for Renewal of Operating Licensees for Nuclear Power Plants,'' which will allow licenses to renew the operating licenses on their nuclear power plants for an additional 20 years beyond the original 40-year limit. A Generic Environmental Impact Statement (GEIS) prepared by the Oak Ridge National Laboratory (ORNL) in conjunction with and for the Nuclear Regulatory Commission to assess the environmental issues associated with this proposed rule. The evaluation of the environmental impact from postulated severe accidents was included in the GEIS. During this evaluation of postulated severe accidents, a method was developed to estimate the public health consequences of atmospheric releases from severe accidents that is much simpler to use than existing consequence computer codes. From the results of this work, it is concluded that the simplified methodology does provide reasonable and conservative estimates of public risk from atmospheric releases from severe accidents

  6. Development of Draft Regulatory Guide on Accident Analysis for Nuclear Power Plants with New Safety Design Features

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Woo, Sweng Woong; Hwang, Tae Suk [KINS, Daejeon (Korea, Republic of); Sim, Suk K; Hwang, Min Jeong [Environment and Energy Technology, Daejeon (Korea, Republic of)

    2016-05-15

    The present paper discusses the development process of the draft version of regulatory guide (DRG) on accident analysis of the NPP having the NSFD and its result. Based on the consideration on the lesson learned from the previous licensing review, a draft regulatory guide (DRG) on accident analysis for NPP with new safety design features (NSDF) was developed. New safety design features (NSDF) have been introduced to the new constructing nuclear power plants (NPP) since the early 2000 and the issuance of construction permit of SKN Units 3 and 4. Typical examples of the new safety features includes Fluidic Device (FD) within Safety Injection Tanks (SIT), Passive Auxiliary Feedwater System (PAFS), ECCS Core Barrel Duct (ECBD) which were adopted in APR1400 design and/or APR+ design to improve the safety margin of the plants for the postulated accidents of interest. Also several studies of new concept of the safety system such as Hybrid ECCS design have been reported. General and/or specific guideline of accident analysis considering the NSDF has been requested. Realistic evaluation of the impact of NSDF on accident with uncertainty and separated accident analysis accounting the NSDF impact were specified in the DRG. Per the developmental process, identification of key issues, demonstration of the DRG with specific accident with specific NSDF, and improvement of DGR for the key issues and their resolution will be conducted.

  7. Trial evaluation on criticality safety of the fuel assemblies at falling accident as spent fuel transport and storage cask

    International Nuclear Information System (INIS)

    Tadano, Tomoaki

    2016-01-01

    The authors conducted critical safety assessment on the supposed event at the time of a fall accident of cask, and examined the influence on criticality safety. If the spacer of fuel assembly is sound, it is assumed that the pitch of fuel rod interval changes, and if the spacer is broken, it is assumed that the fuel rod is unevenly distributed in the basket. For the critical calculation of fuel assembly basket system, they performed it using a calculation code. For both of the single cell and assembly, calculation results showed an increase in the effective multiplication factor of reactivity of 2-3%. When this reactivity is applied to the criticality analysis result of PWR fuel assembly, the value approaches to the limit 0.95 of the neutron effective multiplication factor keff. However, the keff when new fuel is loaded is sufficiently lower than 0.93. Therefore, it is unlikely that the criticality analysis result approaches to 0.95 at all burnups, and the possibility to become criticality is very low in actual spent fuel transport. When considering the reactivity of this research, it is possible that the design condition for the assumption of novel fuel loading becomes severer. Furthermore, criticality analysis under non-uniform pitch will become necessary, and criticality safety analysis for BWR fuel with heterogeneous enrichment degree and burnup degree will become also necessary. (A.O.)

  8. Impact of the great east Japan earthquake, tsunami, nuclear power plant accident

    International Nuclear Information System (INIS)

    Hosoya, Mitsuaki

    2013-01-01

    The title subject is described mainly from the pediatric aspect. Shortly after the Quake (Mar. 11, 2011), Disaster Medical Assistance Team (DMAT) and Japan Medical Association Team (JMAT) started on their disaster emergent activity with various personnel and students of Fukushima Medical University (FMU). The Power Plant Accident broke out on the next day, and FMU was the base in charge of multiple managements of radiological medicare as well as the ordinary emergent one. Number of children emergently admitted in or requiring the general pediatric consultation was rather small, and problems of insufficient pediatric articles were virtually solved within 2 weeks. Pediatric support by FMU was done from April to May end for children in evacuation places. At 3 months after the disaster, the birth number markedly decreased near the Plant, and pediatric in- and out-patient number also diminished in the whole Fukushima prefecture, suggesting that many at pregnancy or having infants had evacuated out of the prefecture probably because of their concern of possible radiation health hazard. In consideration of epidemiology of A-bomb survivors and of victims in Chernobyl Accident, so much increased prevalence of pediatric thyroid cancer is conceived to be hardly observed after Fukushima Accident. The project of Fukushima Health Management Survey involves fundamental and detailed examinations. The former subject is to all of prefectural residents who lived at the time of the Quake and the latter, to all children's thyroid with the age <18 y by ultrasonographic follow-up, to all residents dwelled in the evacuation areas by the detailed physical and mental/life-style examinations and to pregnant women by questionnaire and follow-up. Residents' concern is mostly toward the health of infants who are sensitive to radiation. (T.T.)

  9. Containment failure modes preliminary analysis for Atucha-I nuclear power plant during severe accidents

    International Nuclear Information System (INIS)

    Baron, J.; Caballero, C.; Zarate, S.M.

    1997-01-01

    The present work has the objective to analyze the containment behavior of the Atucha-I nuclear power plant during a severe accident, as part of a probabilistic safety assessment (PSA). Initially, a generic description of the containment failure modes considered in other PSAs is performed. Then, the possible containment failure modes for Atucha I are qualitatively analyzed, according to it design peculiarities. These failure modes involve some substantial differences from other PSAs, due to the particular design of Atucha I. Among others, it is studied the influence of: moderator/coolant separation, existence of cooling Zircaloy channels, existence of filling bodies inside the pressure vessel, reactor cavity geometry, on-line refueling mode, and existence of a double shell containment (steel and concrete) with an annular separation room. As a functions of the before mentioning analysis, a series of parameters to be taken into account is defined, on a preliminary basis, for definition of the plant damage states. (author) [es

  10. Antidiabetic plant-derived nutraceuticals: a critical review.

    Science.gov (United States)

    Naveen, Jayapal; Baskaran, Vallikannan

    2018-06-01

    Diabetes mellitus (DM) is one of the major health problems in the world, especially amongst the urban population. Chemically synthesized drugs used to decrease the ill effects of DM and its secondary complications cause adverse side effects, viz., weight gain, gastrointestinal disturbances, and heart failure. Currently, various other approaches, viz., diet control, physical exercise and use of antidiabetic plant-derived molecules/foods are advocated to manage DM, as they are economical with fewer or no side effects. This review mainly focuses on antidiabetic plants, chemically characterized plant molecules and plant-based foods in the treatment of DM. Very little science-based evidence is available on the mechanism of action of plant-derived food molecules on the DM targets. Critical DM targets include α-amylase, α-glucosidase, DPP-IV, aldose reductase, PPAR-γ, AMP kinase and GLUT4. In-depth studies carried out on a few of those targets with specific mechanisms of action are addressed in this review. This review may help future researchers in identifying a right plant molecule to treat DM or to develop food formulations for DM management.

  11. Questions concerning safety and risk after the nuclear accidents in Japan. Deepened accident analysis for the Fukushima Daiichi power plant; Sicherheits- und Risikofragen im Nachgang zu den nuklearen Stoer- und Unfaellen in Japan. Vertiefte Ereignisanalyse zur Anlage Fukushima-Daini

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Englert, Matthias [Oeko-Institut e.V. - Institut fuer Angewandte Oekologie, Darmstadt (Germany)

    2015-02-25

    The study questions concerning safety and risk in Japanese power plants following the disastrous nuclear accident covers the following issues: the nuclear facility Fukushima Daiichi, site characterization, important technical equipment, important electro-technical equipment, personal; description of the accident progression in the Fukushima nuclear power plant: impact of the earthquake, impact of the tsunami, short-term measures of the operating personnel, pressure and temperature situation in the containments, restoration of the after-heat cooling system in the units 1/2 and 4, fuel element storage pool, summarized parameters during the accident progress; comparative analysis of the accident progression at the Fukushima Daiichi site.

  12. Safety requirement of the nuclear power plants, after TMI-2 accident and their possible implementation on Bushehr NPP

    International Nuclear Information System (INIS)

    Mirhabibi, N.; Tochai, M.T.M.; Ashrafi, A.; Farnoudi, E.

    1985-01-01

    Based on the lessons learned from the TMI-2 accident and other research and developments, many improvements have been required for the design, manufacturing and operation of nuclear power plants in recent years. These requirements have already been implemented to the plants in operation and considered as new safety requirements for new plants. In the present paper these requirements and their possible implementation on Bushehr NPP are discussed. (Author)

  13. Criticality safety philosophy for the Sellafield MOX plant

    International Nuclear Information System (INIS)

    Edge, Jane; Gulliford, Jim

    2003-01-01

    The Sellafield MOX Plant (SMP) has been operational since 2001, blending plutonium dioxide from THORP reprocessing operations, with uranium dioxide to produce Mixed Oxide (MOX) fuel elements. In handling the quantities of fuel associated with a commercial fuel fabrication plant, it is necessary to impose criticality controls. Plutonium dioxide (PuO 2 ), uranium dioxide (UO 2 ) and recycled MOX are mixed together in batches. An Engineered Protection System (EPS) prevents the production of MOX powder in excess of 20w/o Pu(fissile)/(Pu+U), achieved through the combination of a weight-based' system and a diverse 'neutron monitoring' radiometric system. The 'neutron monitoring' component of the EPS determines the fissile enrichment of the batch of MOX powder, based on pessimistic isotopic requirements of the PuO 2 feedstock powder. Guaranteeing the maximum MOX enrichment of 20w/o Pu(fissile)/(Pu + U) at an early stage of the fuel manufacturing process enables the criticality safety assessor to demonstrate that normal operations are deterministically safe. This paper describes in detail the EPS at the front end of plant and the engineered and operational protection in downstream areas. In addition plant operational experience in producing the first fuel assemblies is discussed. (author)

  14. Real-time assessment of radiation burden of the population in the vicinity of nuclear power plants during radiation accidents

    International Nuclear Information System (INIS)

    Stubna, M.

    1986-01-01

    The method is presented of real-time calculation of the radiation situation (dose equivalents) in the environs of a nuclear power plant in case of an accident involving the release of radioactive substances into the atmosphere, this for the potentially most significant exposure paths in the initial and medium stages of the accident. The method allows to take into consideration the time dependence of the rate of radioactive substance release from the nuclear power plant and to assess the development in space and time of the radiation situation in the environs of the nuclear power plant. The use of the method is illustrated by an example of the calculation of the development of the radiation situation for model accidents of a hypothetical PWR with containment. (author)

  15. Flamanville plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Flamanville plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 2 parts: one part dedicated to the first 2 reactors of the plant and the second part to the EPR that is being built. Each part is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  16. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix I. Accident definition and use of event trees

    International Nuclear Information System (INIS)

    1975-10-01

    Information is presented concerning accident definition and use of event trees, event tree methodology, potential accidents covered by the reactor safety study, analysis of potential accidents involving the reactor core, and analysis of potential accidents not involving the core

  17. Cancer incidence in northern Sweden before and after the Chernobyl nuclear power plant accident.

    Science.gov (United States)

    Alinaghizadeh, Hassan; Tondel, Martin; Walinder, Robert

    2014-08-01

    Sweden received about 5 % of the total release of (137)Cs from the Chernobyl nuclear power plant accident in 1986. The distribution of the fallout mainly affected northern Sweden, where some parts of the population could have received an estimated annual effective dose of 1-2 mSv per year. It is disputed whether an increased incidence of cancer can be detected in epidemiological studies after the Chernobyl nuclear power plant accident outside the former Union of Soviet Socialist Republics. In the present paper, a possible exposure-response pattern between deposition of (137)Cs and cancer incidence after the Chernobyl nuclear power plant accident was investigated in the nine northernmost counties of Sweden (2.2 million inhabitants in 1986). The activity of (137)Cs from the fallout maps at 1986 was used as a proxy for the received dose of ionizing radiation. Diagnoses of cancer (ICD-7 code 140-209) from 1980 to 2009 were received from the Swedish Cancer Registry (273,222 cases). Age-adjusted incidence rate ratios, stratified by gender, were calculated with Poisson regression in two closed cohorts of the population in the nine counties 1980 and 1986, respectively. The follow-up periods were 1980-1985 and 1986-2009, respectively. The average surface-weighted deposition of (137)Cs at three geographical levels; county (n = 9), municipality (n = 95) and parish level (n = 612) was applied for the two cohorts to study the pre- and the post-Chernobyl periods separately. To analyze time trends, the age-standardized total cancer incidence was calculated for the general Swedish population and the population in the nine counties. Joinpoint regression was used to compare the average annual percent change in the general population and the study population within each gender. No obvious exposure-response pattern was seen in the age-adjusted total cancer incidence rate ratios. A spurious association between fallout and cancer incidence was present, where areas with the

  18. Cancer incidence in northern Sweden before and after the Chernobyl nuclear power plant accident

    International Nuclear Information System (INIS)

    Alinaghizadeh, Hassan; Tondel, Martin; Walinder, Robert

    2014-01-01

    Sweden received about 5 % of the total release of "1"3"7Cs from the Chernobyl nuclear power plant accident in 1986. The distribution of the fallout mainly affected northern Sweden, where some parts of the population could have received an estimated annual effective dose of 1-2 mSv per year. It is disputed whether an increased incidence of cancer can be detected in epidemiological studies after the Chernobyl nuclear power plant accident outside the former Union of Soviet Socialist Republics. In the present paper, a possible exposure-response pattern between deposition of "1"3"7Cs and cancer incidence after the Chernobyl nuclear power plant accident was investigated in the nine northernmost counties of Sweden (2.2 million inhabitants in 1986). The activity of "1"3"7Cs from the fallout maps at 1986 was used as a proxy for the received dose of ionizing radiation. Diagnoses of cancer (ICD-7 code 140-209) from 1980 to 2009 were received from the Swedish Cancer Registry (273,222 cases). Age-adjusted incidence rate ratios, stratified by gender, were calculated with Poisson regression in two closed cohorts of the population in the nine counties 1980 and 1986, respectively. The follow-up periods were 1980-1985 and 1986-2009, respectively. The average surface-weighted deposition of "1"3"7Cs at three geographical levels; county (n = 9), municipality (n = 95) and parish level (n = 612) was applied for the two cohorts to study the pre- and the post-Chernobyl periods separately. To analyze time trends, the age-standardized total cancer incidence was calculated for the general Swedish population and the population in the nine counties. Joinpoint regression was used to compare the average annual percent change in the general population and the study population within each gender. No obvious exposure-response pattern was seen in the age-adjusted total cancer incidence rate ratios. A spurious association between fallout and cancer incidence was present, where areas with the lowest

  19. Cancer incidence in northern Sweden before and after the Chernobyl nuclear power plant accident

    Energy Technology Data Exchange (ETDEWEB)

    Alinaghizadeh, Hassan; Tondel, Martin; Walinder, Robert [Uppsala University, Occupational and Environmental Medicine, Department of Medical Sciences, Uppsala (Sweden)

    2014-08-15

    Sweden received about 5 % of the total release of {sup 137}Cs from the Chernobyl nuclear power plant accident in 1986. The distribution of the fallout mainly affected northern Sweden, where some parts of the population could have received an estimated annual effective dose of 1-2 mSv per year. It is disputed whether an increased incidence of cancer can be detected in epidemiological studies after the Chernobyl nuclear power plant accident outside the former Union of Soviet Socialist Republics. In the present paper, a possible exposure-response pattern between deposition of {sup 137}Cs and cancer incidence after the Chernobyl nuclear power plant accident was investigated in the nine northernmost counties of Sweden (2.2 million inhabitants in 1986). The activity of {sup 137}Cs from the fallout maps at 1986 was used as a proxy for the received dose of ionizing radiation. Diagnoses of cancer (ICD-7 code 140-209) from 1980 to 2009 were received from the Swedish Cancer Registry (273,222 cases). Age-adjusted incidence rate ratios, stratified by gender, were calculated with Poisson regression in two closed cohorts of the population in the nine counties 1980 and 1986, respectively. The follow-up periods were 1980-1985 and 1986-2009, respectively. The average surface-weighted deposition of {sup 137}Cs at three geographical levels; county (n = 9), municipality (n = 95) and parish level (n = 612) was applied for the two cohorts to study the pre- and the post-Chernobyl periods separately. To analyze time trends, the age-standardized total cancer incidence was calculated for the general Swedish population and the population in the nine counties. Joinpoint regression was used to compare the average annual percent change in the general population and the study population within each gender. No obvious exposure-response pattern was seen in the age-adjusted total cancer incidence rate ratios. A spurious association between fallout and cancer incidence was present, where areas with

  20. Determination of gamma-ray exposure rate from short-lived fission products under criticality accident conditions

    International Nuclear Information System (INIS)

    Yanagisawa, Hiroshi; Ohno, Akio; Aizawa, Eijyu

    2002-01-01

    For the assessment of γ-ray doses from short-lived fission products (FPs) under criticality accident conditions, γ-ray exposure rates varying with time were experimentally determined in the Transient Experiment Critical Facility (TRACY). The data were obtained by reactivity insertion in the range of 1.50 to 2.93$. It was clarified from the experiments that the contribution of γ-ray from short-lived FPs to total exposure during the experiments was evaluated to be 15 to 17%. Hence, the contribution cannot be neglected for the assessment of γ-ray doses under criticality accident conditions. Computational analyses also indicated that γ-ray exposure rates from short-lived FPs calculated with the Monte Carlo code, MCNP4B, and photon sources based on the latest FP decay data, the JENDL FP Decay Data File 2000, well agreed with the experimental results. The exposure rates were, however, extremely underestimated when the photon sources were obtained by the ORIGEN2 code. The underestimation is due to lack of energy-dependent photon emission data for major short-lived FP nuclides in the photon database attached to the ORIGEN2 code. It was also confirmed that the underestimation arose in 1,000 or less of time lapse after an initial power burst. (author)

  1. Nuclear accidents

    International Nuclear Information System (INIS)

    1987-01-01

    On 27 May 1986 the Norwegian government appointed an inter-ministerial committee of senior officials to prepare a report on experiences in connection with the Chernobyl accident. The present second part of the committee's report describes proposals for measures to prevent and deal with similar accidents in the future. The committee's evaluations and proposals are grouped into four main sections: Safety and risk at nuclear power plants; the Norwegian contingency organization for dealing with nuclear accidents; compensation issues; and international cooperation

  2. Study on radioactive fallout from Fukushima nuclear accident by plant samples using an imaging plate system

    International Nuclear Information System (INIS)

    Minowa, Haruka

    2011-01-01

    The radioactive fallout from the Fukushima nuclear accident was investigated by the radiation images of plant samples using an Imaging Plate System. Plant samples exposed by an imaging plate BASIII 2040 (Fujifilm, Japan) in overnight to one week, and radiation images were read by Typhoon FLA7000 (GE Healthcare Japan Corp.). Identifying and quantitative analysis of radionuclides were measured by Auto Well Gamma System ARC-380CL (Aloha Co. Ltd., Tokyo, Japan). In the cross-sectional images of the bamboo shoot, the radioactive material is shown in heterogeneous distribution, it was found that it concentrated on the tip of the edible portion, and thin skin. These radionuclides were identified as "1"3"7Cs, "1"3"4Cs, and "4"0K. "4"0K is a natural radionuclide, on the other hand "1"3"7Cs and "1"3"4Cs would be derived from the accident of the Fukushima Daiichi Nuclear Power Plant. A high concentration of "1"3"4Cs was shown at the distance of 150 mm from the base of the bamboo shoot by cross-sectional cutting into the width of about 1 mm. It was estimated about 1 kBq of "1"3"4Cs would be included in about 400 g (wet weight) of this one bamboo shoot in an edible part. Imaging data suggests that the contamination of radioactive cesium in this bamboo shoot was caused not by the extraneous attachment but by the absorption from roots. Because bamboo is gather water from extensive area, bamboo shoot concentrates the radioactive material contained in the rain even at low concentrations of radioactive materials in soil. (author)

  3. Description of Survey Data Regarding the Chemical Repackaging Plant Accident West Helena, Arkansas

    Energy Technology Data Exchange (ETDEWEB)

    Sorensen, J.H.; Vogt, B.M.

    1999-03-01

    Shortly after 1:00 p.m. on Thursday, May 8, 1997, clouds of foul-smelling smoke began pouring from an herbicide and pesticide packaging plant in West Helena, Arkansas. An alert was sounded, employees evacuated, and the West Helena fire department was called. As three firefighters prepared to enter the plant, the chemical compounds exploded, collapsing a solid concrete block wall, and killing all three firefighters. As the odorous smoky cloud drifted away from the plant, authorities ordered residents in a 2-mile area downwind of the plant to evacuate and those in the 2- to 3-mile zone to shelter in place. This study examines and compares the responses to a mail survey of those ordered to evacuate and those told to shelter in place. Among the variables examined are compliance with official orders and perceived warnings, threat perception, time and source of first warning, response times, and behavior characteristics for both populations. The findings indicate that 90% of those that were told to evacuate did so but only 27% of those told to shelter-in-place did so, with 68% opting to evacuate instead. The implications of these findings for emergency managers is that people will likely choose to evacuate when both warnings to evacuate and warnings to shelter are issued to residents in close proximity to each other. The findings on warning times closely resemble other findings from evacuations when chemical accidents occur and route notification is used for warning residents.

  4. WWER-440/V-230 Confinement modernization to upgrade the critical safety function 'Containment integrity' in case of severe accident

    International Nuclear Information System (INIS)

    Sartmadjiev, A.

    1999-01-01

    In this lecture the WWER-440/V-230 confinement modernization to upgrade the critical safety function 'Containment integrity' in case of severe accident is presented. There are discussed: design limitations of the location system; consequence from these design limitations; a few confinement reconstruction concepts of this type of units worldwide; and purpose of the confinement reconstruction - to improve significantly the original design, ensuring (1) localization for all possible primary breaks and (2) limitation of the radiological consequences for the personnel, the population and the environment below the regulatory requirements

  5. Oxidation behavior analysis of cladding during severe accidents with combined codes for Qinshan Phase II Nuclear Power Plant

    International Nuclear Information System (INIS)

    Shi, Xingwei; Cao, Xinrong; Liu, Zhengzhi

    2013-01-01

    Highlights: • A new verified oxidation model of cladding has been added in Severe Accident Program (SAP). • A coupled analysis method utilizing RELAP5 and SAP codes has been developed and applied to analyze a SA caused by LBLOCA. • Analysis of cladding oxidation under a SA for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) has been performed by SAP. • Estimation of the production of hydrogen has been achieved by coupled codes. - Abstract: Core behavior at a high temperature is extremely complicated during transition from Design Basic Accident (DBA) to the severe accident (SA) in Light Water Reactors (LWRs). The progression of core damage is strongly affected by the behavior of fuel cladding (oxidation, embrittlement and burst). A Severe Accident Program (SAP) is developed to simulate the process of fuel cladding oxidation, rupture and relocation of core debris based on the oxidation models of cladding, candling of melted material and mechanical slumping of core components. Relying on the thermal–hydraulic boundary parameters calculated by RELAP5 code, analysis of a SA caused by the large break loss-of-coolant accident (LBLOCA) without mitigating measures for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) was performed by SAP for finding the key sequences of accidents, estimating the amount of hydrogen generation and oxidation behavior of the cladding

  6. Prediction of hydrogen concentration in nuclear power plant containment under severe accidents using cascaded fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Geon Pil; Kim, Dong Yeong; Yoo, Kwae Hwan; Na, Man Gyun, E-mail: magyna@chosun.ac.kr

    2016-04-15

    Highlights: • We present a hydrogen-concentration prediction method in an NPP containment. • The cascaded fuzzy neural network (CFNN) is used in this prediction model. • The CFNN model is much better than the existing FNN model. • This prediction can help prevent severe accidents in NPP due to hydrogen explosion. - Abstract: Recently, severe accidents in nuclear power plants (NPPs) have attracted worldwide interest since the Fukushima accident. If the hydrogen concentration in an NPP containment is increased above 4% in atmospheric pressure, hydrogen combustion will likely occur. Therefore, the hydrogen concentration must be kept below 4%. This study presents the prediction of hydrogen concentration using cascaded fuzzy neural network (CFNN). The CFNN model repeatedly applies FNN modules that are serially connected. The CFNN model was developed using data on severe accidents in NPPs. The data were obtained by numerically simulating the accident scenarios using the MAAP4 code for optimized power reactor 1000 (OPR1000) because real severe accident data cannot be obtained from actual NPP accidents. The root-mean-square error level predicted by the CFNN model is below approximately 5%. It was confirmed that the CFNN model could accurately predict the hydrogen concentration in the containment. If NPP operators can predict the hydrogen concentration in the containment using the CFNN model, this prediction can assist them in preventing a hydrogen explosion.

  7. Health status and follow-up of the Chernobyl Nuclear Power Plant accident liquidators in Latvia

    International Nuclear Information System (INIS)

    Curbakova, E.; Dzerve, B.; Eglite, M.; Frickausa, I.; Zvagule, T.

    1996-01-01

    The accident at the Nuclear Power Plant in Chernobyl create a new problem for health professionals in Latvia due to the fact that 6475 inhabitants (mainly healthy and men of reproductive age) of Latvia took part in clear-up works in Chernobyl within the period 1986-1991. Chernobyl clear-up workers were exposed γ-radiation and they also incorporated radionuclides. The doses documented for the clear-up workers are variable; they are estimated to be between 0.01-0.5 Gy although the specialists working on the precision of received doses think that they could be even 2 or 3 times higher. The aim of this work is to evaluate the health status of liquidators investigating them on a long-term basis: to create the correct system of health status evaluation of Chernobyl clear-up workers, to improve the register of Chernobyl clear-up workers and of their children, to analyze the data about the incidence of different diseases and mortality gained from follow-ups, to evaluate health status and clinical picture within the period of time, to work out and use adequate methods of treatment. Chernobyl clear-up workers more often than the control group suffer from diseases of the nervous, the endocrine and the metabolic and immune system. They also have higher rate of incidence for diseases of digestive and respiratory system and for diseases of bones, muscles and connective tissue higher rates of accidents and suicides. Now, ten years after the accident there are Chernobyl clear-up workers who are chronically ill and their health status is expected to be worse in the next few years. Regular follow-up and medical examination of Chernobyl clear-up workers and their children should be carried out every year. Regular rehabilitation of Chernobyl clear-up workers should be provided by the government

  8. Radioactive contamination of Danish territory after core-melt accidents at the Barsebaeck power plant

    International Nuclear Information System (INIS)

    Gjoerup, H.L.; Jensen, N.O.; Hedemann Jensen, P.; Kristensen, L.; Nielsen, O.J.; Petersen, E.L.; Petersen, T.; Roed, J.; Thykier-Nielsen, S.; Heikel Vinter, F.; Warming, L.; Aarkrog, A.

    1982-03-01

    An assessment is made of the radioactive contamination of Danish territory in the event of a core-melt accident at the Barsebaeck nuclear power plant in Sweden. Accidents including both core melt-down and containment failure are considered. Consequences are calculated for a BWR-3 release under common meteorological conditions and for a BWR-2 release under extreme meteorological conditions. Calculations are based on experiments and theoretical work relating to deposition velocities for different types of surface, shielding effect of structures, and weathering. The effects are described of different dose-reducing measures, e.g., decontamination, relocation, destruction of contaminated foodstuffs. The collective effective dose equivalent from external gamma radiation from deposited activity integrated over a time period of 30 years, is calculated to be 3.6 Megamanrem in the BWR-3 case without dose-reducing measures. For the BWR-2 case, the corresponding dose is approx. 41 Megamanrem. A combination of temporary relocation, hosing of roads etc. and digging of gardens is estimated to reduce these doses to approx. 2.5 Megamanrem and approx. 15 Megamanrem, respectively. The collective committed effective dose equivalent from the consumption of contaminated foodstuffs is calculated to 23 Megamanrem in the BWR-3 case without dose-reducing measures. This dose could be reduced to 0.2 Megamanrem if contaminated crops are destroyed during the first year after the accident and if changes are made in agricultural production in the contaminated area. The corresponding doses in the BWR-2 case would be 197 Megamanrem and 1.4 Megmanrem, respectively. (author)

  9. Health status and follow-up of the Chernobyl Nuclear Power Plant accident liquidators in Latvia

    Energy Technology Data Exchange (ETDEWEB)

    Curbakova, E; Dzerve, B; Eglite, M; Frickausa, I; Zvagule, T [Centre of Occupational and Radiological Medicine of P. Stradins State clinical Hospital, Riga (Latvia)

    1996-07-01

    The accident at the Nuclear Power Plant in Chernobyl create a new problem for health professionals in Latvia due to the fact that 6475 inhabitants (mainly healthy and men of reproductive age) of Latvia took part in clear-up works in Chernobyl within the period 1986-1991. Chernobyl clear-up workers were exposed {gamma}-radiation and they also incorporated radionuclides. The doses documented for the clear-up workers are variable; they are estimated to be between 0.01-0.5 Gy although the specialists working on the precision of received doses think that they could be even 2 or 3 times higher. The aim of this work is to evaluate the health status of liquidators investigating them on a long-term basis: to create the correct system of health status evaluation of Chernobyl clear-up workers, to improve the register of Chernobyl clear-up workers and of their children, to analyze the data about the incidence of different diseases and mortality gained from follow-ups, to evaluate health status and clinical picture within the period of time, to work out and use adequate methods of treatment. Chernobyl clear-up workers more often than the control group suffer from diseases of the nervous, the endocrine and the metabolic and immune system. They also have higher rate of incidence for diseases of digestive and respiratory system and for diseases of bones, muscles and connective tissue higher rates of accidents and suicides. Now, ten years after the accident there are Chernobyl clear-up workers who are chronically ill and their health status is expected to be worse in the next few years. Regular follow-up and medical examination of Chernobyl clear-up workers and their children should be carried out every year. Regular rehabilitation of Chernobyl clear-up workers should be provided by the government.

  10. Calculation notes that support accident scenario and consequence determination of a waste tank criticality

    International Nuclear Information System (INIS)

    Marusich, R.M. Westinghouse Hanford

    1996-01-01

    The purpose of this calculation note is to provide the basis for criticality consequences for the Tank Farm Safety Analysis Report (FSAR). Criticality scenario is developed and details and description of the analysis methods are provided

  11. Mitigation of severe accidents in AREVA's Gen 3+ nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, M., E-mail: manfred.fischer@areva.com; Henning, A.; Surmann, R.

    2014-04-01

    The current AREVA Gen 3+ PWR designs (EPR™ and ATMEA1) are based on the proven defense-in-depth safety concepts inherited from their predecessors, the French “N4” and the German “Konvoi” reactors. Complemented by specific enhancements, including higher redundancy and diversity as well as the use of passive systems, this leads to very low values of the core damage frequency (CDF). Notwithstanding this very low probability, dedicated design measures have been implemented to improve the response of the plant in case of a postulated severe accident (SA) with core melting. This way not only the frequency of large-early-releases (LERF) but also the related radiological consequences are drastically reduced. Situations that potentially lead to high loads that can challenge the short-term integrity of the containment, like RPV melt-through under high pressure, energetic hydrogen/steam explosions, as well as long-term containment failure caused by internal over-pressure are avoided by a combination of preventive measures and dedicated systems. At the example of the EPR{sup TM}, the paper gives an overview of the severe accident mitigation strategy and the related measures and systems of AREVAs current Gen 3+ reactors, with special focus on the function of the core melt stabilization system.

  12. Reducing logistical barriers to radioactive soil remediation after the Fukushima No. 1 nuclear power plant accident

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, K., E-mail: keizo.ishii@qse.tohoku.ac.jp [Research Center for Remediation Engineering of Living Environments Contaminated with Radioisotopes, Department of Quantum Science and Energy Engineering, Tohoku University, 6-6 Aza-Aoba, Aramaki, Aoba-ku, Sendai 980-8579 (Japan); Terakawa, A.; Matsuyama, S.; Kikuchi, Y.; Fujishiro, F.; Ishizaki, A.; Osada, N.; Arai, H.; Sugai, H.; Takahashi, H.; Nagakubo, K.; Sakurada, T. [Research Center for Remediation Engineering of Living Environments Contaminated with Radioisotopes, Department of Quantum Science and Energy Engineering, Tohoku University, 6-6 Aza-Aoba, Aramaki, Aoba-ku, Sendai 980-8579 (Japan); Yamazaki, H.; Kim, S. [Cyclotron and Radioisotope Center, Tohoku University, 6-3 Aza-Aoba, Aramaki, Aoba-ku, Sendai 980-8578 (Japan)

    2014-01-01

    We present an updated assessment of soil contamination due to the nuclear accident at the Fukushima No. 1 nuclear power plant on 11 March 2011. A safe limit for the spatial dose rate (micro-Sv/h) of gamma rays from {sup 134,137}Cs has been established in this work. Based on this value, the highly contaminated region within Fukushima Prefecture that must be decontaminated could be defined. Moreover, a conceptual model for the chemical speciation that occurred during the accident has been delineated. The compound model Cs{sub 2}CO{sub 3} was found to be meaningful and practical (non-radioactive) to simulate contamination in our decontamination experiments. Finally, we explain the mechanism of action of our soil remediation technique, which effectively reduces the total volume of contaminated soil by isolating the highly Cs-adsorptive clay fraction. The adsorption of non-radioactive Cs atoms on clay particles with diameters <25 μm were analyzed using micro-particle-induced X-ray emission (PIXE)

  13. The European Research on Severe Accidents in Generation-II and -III Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Jean-Pierre Van Dorsselaere

    2012-01-01

    Full Text Available Forty-three organisations from 22 countries network their capacities of research in SARNET (Severe Accident Research NETwork of excellence to resolve the most important remaining uncertainties and safety issues on severe accidents in existing and future water-cooled nuclear power plants (NPP. After a first project in the 6th Framework Programme (FP6 of the European Commission, the SARNET2 project, coordinated by IRSN, started in April 2009 for 4 years in the FP7 frame. After 2,5 years, some main outcomes of joint research (modelling and experiments by the network members on the highest priority issues are presented: in-vessel degraded core coolability, molten-corium-concrete-interaction, containment phenomena (water spray, hydrogen combustion…, source term issues (mainly iodine behaviour. The ASTEC integral computer code, jointly developed by IRSN and GRS to predict the NPP SA behaviour, capitalizes in terms of models the knowledge produced in the network: a few validation results are presented. For dissemination of knowledge, an educational 1-week course was organized for young researchers or students in January 2011, and a two-day course is planned mid-2012 for senior staff. Mobility of young researchers or students between the European partners is being promoted. The ERMSAR conference is becoming the major worldwide conference on SA research.

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

    Science.gov (United States)

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

    2014-02-18

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

  15. The FIRAC code - its applicability and boundary conditions for fire accident analysis in a reprocessing plant

    International Nuclear Information System (INIS)

    Roewekamp, M.

    1991-01-01

    After a short description of the modelling capabilities and the implementation of the computer code the possible applications of FIRAC are demonstrated by means of two test-examples. The so gained experiences with respect to the variation of different parameters, convergency criteria, etc. can be used for the simulation of a fire accident in the storage area for unconditioned combustible low active waste (LAW) of the planned reprocessing plant at Wackersdorf. The code is prepared for calculating direct effects (of the fire) in the fire room as well as particularly effects on adjacent rooms and ventilation systems. Source terms for the release of radioactive particles outside a building can also be investigated. The temperature and pressure curves for the fire room as well as for other areas in the facility show that no damages caused by temperature effects are expected for the considered fire of low active waste. As a result of the calculated mass and volumetric flows radioactive aerosole particles could be transported into normally non-active areas. The FIRAC code renders the possibility of a more detailed analysis of those parameters relevant for fire accidents and by this means completes the so far phenomenological procedure of the fire hazard analysis in nuclear facilities. (orig.) [de

  16. Risk of Thyroid Cancer after the Fukushima Nuclear Power Plant Accident

    International Nuclear Information System (INIS)

    Yamashita, S.

    2016-01-01

    Full text: A sound scientific understanding about the relationship between radiation dose and health risk is needed to apply any countermeasure against radiological and nuclear accidents. Since the Great East Japan earthquake and the Fukushima Daiichi Nuclear Power Plant accident in Japan, Fukushima Prefecture has started the Fukushima Health Management Survey Project since June 2011 for the purpose of long-term health care administration for the prefectural residents. There are considerable differences between Chernobyl and Fukushima regarding radiation dose to the public, however, it is still difficult to estimate retrospectively accurate internal exposure dose individually from the short-lived radioactive iodines. Another difficult challenge is to how to manage non-radiation–related health effects, such as post-disaster mental impact and lifestyle changes. As we support residents in their recovery and return to their homes, understanding each individual’s state with respect to radiation and regular monitoring of their health conditions contribute to the region’s rebirth and restoration. Therefore, as one of the tools of risk communication, the necessity of thyroid ultrasound examination in Fukushima and the intermediate results of this survey targeting children will be reviewed and discussed in order to avoid any misunderstanding or misinterpretation of the high detection rate of childhood and adolescent thyroid cancer by mass screening. (author

  17. Severe accident management at nuclear power plants - emergency preparedness and response actions

    International Nuclear Information System (INIS)

    Pawar, S.K.; Krishnamurthy, P.R.

    2015-01-01

    This paper describes the current level of emergency planning and preparedness and also improvement in the emergency management programme over the years including lessons learned from Fukushima accident, hazard analysis and categorization of nuclear facilities into hazard category for establishing the emergency preparedness class, classification of emergencies based on the Emergency Action Levels (EAL), development of EAL’s for PHWR, Generic Criteria in terms of projected dose for initiating protective actions (precautionary urgent protective actions, urgent protective actions, early protective actions), operational intervention levels (OIL), Emergency planning zones and distances, protection strategy and reference levels, use of residual dose for establishing reference levels for optimization of protection strategy, criteria for termination of emergency, transition of emergency exposure situation to existing exposure situation or planned exposure situation, criteria for medical managements of exposed persons and guidance for controlling the dose of emergency workers. This paper also highlights the EALs for typical PHWR type reactors for all types of emergencies (plant, site and offsite), transition from emergency operating procedures (EOP) to accident management guidelines (AMG) to emergency response actions and proposed implementation of guidelines

  18. Criticality safety analysis of TK-13 cask in Bushehr nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mohammadi, Ashgar; Omidvari, Nima [Iran Radioactive Waste Management Company, Tehran (Iran, Islamic Republic of); Hassanzadeh, Mostafa [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of)

    2017-12-15

    Spent fuel production is one of the main problems of nuclear power plants that should be managed properly considering the strategy of each country. Today, in most of nuclear power owner countries, the interim storage has been selected as the temporary solution of spent fuel management because of absence of deep geological repositories and no tendency for reprocessing. On the other side, considering the merits of storage in dual purpose casks based on dry storage, this method was chosen for interim storage. By taking into account that the only operating reactor of Iran is of Water-Water Energetic Reactor (WWER)-1000 type, proposed TK-13 cask by Russia which is the manufacturer of these types of reactors has been considered. In this study, the calculation of basket holding spent fuel assembly criticality of this cask has been analyzed for two modes of fresh and spent fuel by ORIGEN2.1 and MCNPX2.6 nuclear codes. The criterion of the nuclear criticality safety for effective multiplication factor (k{sub eff}) should be 0.95 and 0.98 for many ordinary and accident conditions, respectively. Therefore, the results show that a cylindrical basket with 66 cm diameter and 28 cm pitch with internal holding basket made of borated steel with 0.1% borate and steel free from borate would meet the criticality of cask, respectively.

  19. Criticality safety analysis of TK-13 cask in Bushehr nuclear power plant

    International Nuclear Information System (INIS)

    Mohammadi, Ashgar; Omidvari, Nima; Hassanzadeh, Mostafa

    2017-01-01

    Spent fuel production is one of the main problems of nuclear power plants that should be managed properly considering the strategy of each country. Today, in most of nuclear power owner countries, the interim storage has been selected as the temporary solution of spent fuel management because of absence of deep geological repositories and no tendency for reprocessing. On the other side, considering the merits of storage in dual purpose casks based on dry storage, this method was chosen for interim storage. By taking into account that the only operating reactor of Iran is of Water-Water Energetic Reactor (WWER)-1000 type, proposed TK-13 cask by Russia which is the manufacturer of these types of reactors has been considered. In this study, the calculation of basket holding spent fuel assembly criticality of this cask has been analyzed for two modes of fresh and spent fuel by ORIGEN2.1 and MCNPX2.6 nuclear codes. The criterion of the nuclear criticality safety for effective multiplication factor (k eff ) should be 0.95 and 0.98 for many ordinary and accident conditions, respectively. Therefore, the results show that a cylindrical basket with 66 cm diameter and 28 cm pitch with internal holding basket made of borated steel with 0.1% borate and steel free from borate would meet the criticality of cask, respectively.

  20. CE/Bechtel design containment response to severe accident phenomenology: A comparison among several combustion engineering plants

    International Nuclear Information System (INIS)

    Khalil, Y.F.; Schneider, R.E.

    1995-01-01

    The objectives of this paper are to: (1) discuss the types of severe accident phenomena that drive containment failure modes in CE plants and (2) contribute to the current state of knowledge of CE/Bechtel-design containment response to severe accident phenomenology. The second objective is addressed by providing a comparative study of containment response to severe accidents among several CE plants including Millstone Unit 2 (MP2), Palisades (Consumers Power), Calvert Cliffs (Baltimore Gas and Electric Company), Palo Verde (Arizona Public Service), and SONGS Units 2 and 3 (Southern California Edison). The motivation for addressing the second objective is based on the current lack of comprehensive literature on CE/Bechtel design containment failure modes and mechanisms for accidents that progress beyond the design basis limits. The first part of this paper addresses severe accident phenomena-related failure mechanisms in CE/Bechtel-designed containments. The second part of this work provides a comparative study of containment response among several CE plants

  1. Golfech plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Golfech plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  2. Tricastin plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Tricastin plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  3. Bugey plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Bugey plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  4. Fessenheim plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Fessenheim plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  5. Chinon plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Chinon B plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  6. Saint-Alban plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Saint-Alban plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  7. Blayais plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Blayais plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  8. Civaux plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Civaux plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  9. Cattenom plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Cattenom plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  10. Gravelines plant - Report on the complementary safety assessment of nuclear facilities in the light of the Fukushima accident

    International Nuclear Information System (INIS)

    2011-01-01

    This CSA (Complementary Safety Assessment) analyses the robustness of the Gravelines plant to extreme situations such as those that led to the Fukushima accident and proposes a series of improvements. Robustness is the ability for the plant to withstand events beyond the level for which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Safety is not only a matter of design or of engineered systems, it is also a matter of organization. So issues like EDF's crisis organization, the organization of radiation protection, and work organization via subcontracting are also taken into consideration. The creation of a nuclear rapid action force (FARN) is proposed: this will be a national emergency force made up of specialized teams equipped to intervene in less than 24 hours on a nuclear site hit by an accident. This report is divided into 8 main chapters: 1) features of the site, 2) earthquake risk, 3) flooding risk, 4) risks due to other extreme natural disasters, 5) the loss of electrical power supplies and of heat sink, 6) management of severe accidents (accidents with core melt), 7) task subcontracting policy, 8) synthesis and list of improvements. 4 following appendices review: EDF's crisis organization, the FARN, radiation protection organization and accidental event trees. (A.C.)

  11. Diagnostic and prognostic system for identification of accident scenarios and prediction of 'source term' in nuclear power plants under accident conditions

    International Nuclear Information System (INIS)

    Santhosh; Gera, B.; Kumar, Mithilesh

    2014-01-01

    Nuclear power plant experiences a number of transients during its operations. These transients may be due to equipment failure, malfunctioning of process support systems etc. In such a situation, the plant may result in an abnormal state which is undesired. In case of such an undesired plant condition, the operator has to carry out diagnostic and corrective actions. When an event occurs starting from the steady state operation, instruments' readings develop a time dependent pattern and these patterns are unique with respect to the type of the particular event. Therefore, by properly selecting the plant process parameters, the transients can be distinguished. In this connection, a computer based tool known as Diagnostic and Prognostic System has been developed for identification of large pipe break scenarios in 220 MWe Pressurised Heavy Water Reactors (PHWRs) and for prediction of expected 'Source Term' and consequence for a situation where Emergency Core Cooling System (ECCS) is not available or partially available. Diagnostic and Prognostic System is essentially a transient identification and expected source term forecasting system. The system is based on Artificial Neural Networks (ANNs) that continuously monitors the plant conditions and identifies a Loss Of Coolant Accident (LOCA) scenario quickly based on the reactor process parameter values. The system further identifies the availability of injection of ECCS and in case non-availability of ECCS, it can forecast expected 'Source Term'. The system is a support to plant operators as well as for emergency preparedness. The ANN is trained with a process parameter database pertaining to accident conditions and tested against blind exercises. In order to see the feasibility of implementing in the plant for real-time diagnosis, this system has been set up on a high speed computing facility and has been demonstrated successfully for LOCA scenarios. (author)

  12. The estimation of economic impacts resulting from the severe accidents of a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Jung, Won dea

    2001-03-01

    The economic impacts resulting from the severe accidents of a nuclear power plant were estimated for the different combinations of a release parameters and metrorological data. According to the cost estimation for the basic scenarios, the population dependent cost is dominant. The cost for the protective actions such as evacuation and relocation have a small portion in the total cost and show little variation from scenario to scenario. The economic cost estimation for the seasonal scenarios show very similar trend as that for the basic scenarios. There are little or small variation in the economic cost for the different scenarios for each season except for the season-5 scenario. The health effect value shows maximum in Summer and minimum in Fall. On the contrast, the economic cost shows maximum in Fall and minimum in Summer. The result will be used as basic data in the establishment of effective emergency response and in the cost/benefit analysis in developing optimum risk reduction strategies.

  13. The estimation of economic impacts resulting from the severe accidents of a nuclear power plant

    International Nuclear Information System (INIS)

    Jeong, Jong Tae; Jung, Won dea

    2001-03-01

    The economic impacts resulting from the severe a