Applying probabilistic methods for assessments and calculations for accident prevention

International Nuclear Information System (INIS)

Anon.

1984-01-01

The guidelines for the prevention of accidents require plant design-specific and radioecological calculations to be made in order to show that maximum acceptable expsoure values will not be exceeded in case of an accident. For this purpose, main parameters affecting the accident scenario have to be determined by probabilistic methods. This offers the advantage that parameters can be quantified on the basis of unambigious and realistic criteria, and final results can be defined in terms of conservativity. (DG) [de

Dispersion parameters: impact on calculated reactor accident consequences

Energy Technology Data Exchange (ETDEWEB)

Aldrich, D.C.

1979-01-01

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

Pilot program: NRC severe reactor accident incident response training manual: Severe reactor accident overview

International Nuclear Information System (INIS)

McKenna, T.J.; Martin, J.A.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Giitter, J.G.; Watkins, R.M.

1987-02-01

This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. Severe Reactor Accident Overview is the second in a series of volumes that collectively summarize the US Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume describes elementary perspectives on severe accidents and accident assesment. Each volume serves, respectively, as the text for a course of instruction in a series of courses. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are called out in the text

Response of Soviet VVER-440 accident localization systems to overpressurization

International Nuclear Information System (INIS)

Kulak, R.F.; Fiala, C.; Sienicki, J.J.

1989-01-01

The Soviet designed VVER-440 model V230 and VVER-440 model V213 reactors do not use full containments to mitigate the effects of accidents. Instead, these VVER-440 units employ a sealed set of interconnected compartments, collectively called the accident localization system (ALS), to reduce the release of radionuclides to the atmosphere during accidents. Descriptions of the VVER accident localization structures may be found in the report DOE NE-0084. The objective of this paper is to evaluate the structural integrity of the VVER-440 ALS at the Soviet design pressure, and to determine their response to pressure loadings beyond the design value. Complex, three-dimensional, nonlinear, finite element models were developed to represent the major structural components of the localization systems of the VVER-440 models V230 and V213. The interior boundary of the localization system was incrementally pressurized in the calculations until the prediction of gross failure. 6 refs., 9 figs

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

International Nuclear Information System (INIS)

Behling, S.R.

1985-05-01

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

NASA Medical Response to Human Spacecraft Accidents

Science.gov (United States)

Patlach, Robert

2011-01-01

This slide presentation reviews NASA's role in the response to spacecraft accidents that involve human fatalities or injuries. Particular attention is given to the work of the Mishap Investigation Team (MIT), the first response to the accidents and the interface to the accident investigation board. The MIT does not investigate the accident, but the objective of the MIT is to gather, guard, preserve and document the evidence. The primary medical objectives of the MIT is to receive, analyze, identify, and transport human remains, provide assistance in the recovery effort, and to provide family Casualty Coordinators with latest recovery information. The MIT while it does not determine the cause of the accident, it acts as the fact gathering arm of the Mishap Investigation Board (MIB), which when it is activated may chose to continue to use the MIT as its field investigation resource. The MIT membership and the specific responsibilities and tasks of the flight surgeon is reviewed. The current law establishing the process is also reviewed.

A systematic framework for effective uncertainty assessment of severe accident calculations; Hybrid qualitative and quantitative methodology

International Nuclear Information System (INIS)

Hoseyni, Seyed Mohsen; Pourgol-Mohammad, Mohammad; Tehranifard, Ali Abbaspour; Yousefpour, Faramarz

2014-01-01

This paper describes a systematic framework for characterizing important phenomena and quantifying the degree of contribution of each parameter to the output in severe accident uncertainty assessment. The proposed methodology comprises qualitative as well as quantitative phases. The qualitative part so called Modified PIRT, being a robust process of PIRT for more precise quantification of uncertainties, is a two step process for identifying and ranking based on uncertainty importance in severe accident phenomena. In this process identified severe accident phenomena are ranked according to their effect on the figure of merit and their level of knowledge. Analytical Hierarchical Process (AHP) serves here as a systematic approach for severe accident phenomena ranking. Formal uncertainty importance technique is used to estimate the degree of credibility of the severe accident model(s) used to represent the important phenomena. The methodology uses subjective justification by evaluating available information and data from experiments, and code predictions for this step. The quantitative part utilizes uncertainty importance measures for the quantification of the effect of each input parameter to the output uncertainty. A response surface fitting approach is proposed for estimating associated uncertainties with less calculation cost. The quantitative results are used to plan in reducing epistemic uncertainty in the output variable(s). The application of the proposed methodology is demonstrated for the ACRR MP-2 severe accident test facility. - Highlights: • A two stage framework for severe accident uncertainty analysis is proposed. • Modified PIRT qualitatively identifies and ranks uncertainty sources more precisely. • Uncertainty importance measure quantitatively calculates effect of each uncertainty source. • Methodology is applied successfully on ACRR MP-2 severe accident test facility

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.)

Emergency Response to Radioactive Material Transport Accidents

International Nuclear Information System (INIS)

EL-shinawy, R.M.K.

2009-01-01

Although transport regulations issued by IAEA is providing a high degree of safety during transport opertions,transport accidents involving packages containing radioactive material have occurred and will occur at any time. Whenever a transport accident involving radioactive material accurs, and many will pose no radiation safety problems, emergency respnose actioms are meeded to ensure that radiation safety is maintained. In case of transport accident that result in a significant relesae of radioactive material , loss of shielding or loss of criticality control , that consequences should be controlled or mitigated by proper emergency response actions safety guide, Emergency Response Plamming and Prepardness for transport accidents involving radioactive material, was published by IAEA. This guide reflected all requirememts of IAEA, regulations for safe transport of radioactive material this guide provide guidance to the publicauthorites and other interested organziation who are responsible for establishing such emergency arrangements

Method for consequence calculations for severe accidents

International Nuclear Information System (INIS)

Nielsen, F.; Thykier-Nielsn, S.

1987-03-01

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

System calculations related to the accident at Three-Mile Island using TRAC

International Nuclear Information System (INIS)

Ireland, J.R.

1980-01-01

The Three Mile Island nuclear plant (Unit 2) was modeled using the Transient Reactor Analysis Code (TRAC-P1A) and a base case calculation, which simulated the initial part of the accident that occurred on March 28, 1979, was performed. In addition to the base case calculation, several parametric calculations were performed in which a single hypothetical change was made in the system conditions, such as assuming the high pressure injection (HPI) system operated as designed rather than as in the accident. Some of the important system parameter comparisons for the base case as well as some of the parametric case results are presented

Using MARS to assist in managing a severe accident

International Nuclear Information System (INIS)

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

2004-01-01

During an accident, information about the current and possible future states of the plant provides guidance for accident managers in evaluating which actions should be taken. However, depending upon the nature of the accident and the stress levels imposed on the plant staff responding to the accident the current and future plant assessments may be very difficult or nearly impossible to perform without supplemental training and/or appropriate tools. The MAAP Accident Response System (MARS) has been developed as a calculational aid to assist the responsible accident management individuals. Specifically MARS provides additional insights on the current and possible future states of the plant during an accident including the influence of operator actions. In addition to serving as a calculational aid, the MARS software can be an effective means for providing supplemental training. The MARS software uses engineering calculations to perform an integral assessment of the plant status including a consistency assessment of the available instrumentation. In addition, it uses the Modular Accident Analysis Program (MAAP) to provide near term predictions of the plant response if corrective actions are taken. This paper will discuss the types of information that are beneficial to the accident manager and how MARS addresses each. The MARS calculational functions include: instrumentation, validation and simulation, projected operator response based on the EOPs, as well as estimated timing and magnitude of in-plant and off-site radiation dose releases. Each of these items is influential in the management of a severe accident. (author)

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Calculation of health risks from spent-nuclear-fuel transportation accidents

International Nuclear Information System (INIS)

Chen, S.Y.; Yuan, Y.C.

1987-01-01

Models developed to analyze potential radiological health risks from various accident scenarios during transportation of spent nuclear fuels are described. The models are designed both for detailed route-specific risk analyses and for use in conducting overall risk analyses for route selection and related decision-making activities. The radiological risks calculated include individual dose commitments, collective dose commitments, and long-term (100-year) environmental dose commitments to a population following release of radioactivity. To facilitate route-specific analysis, a state-level database was developed and incorporated into the model. Route-specific analysis is demonstrated by the calculation of radiological risks resulting from various accident scenarios, as postulated by the recent US Nuclear Regulatory Commission Modal Study, for four representative states selected from various regions of the United States. 10 refs., 3 figs., 3 tabs

The implementation of the IAEA accident response plan in Yugoslav practice

International Nuclear Information System (INIS)

Orlic, M.; Pavlovic, R.; Markovic, S.; Pavlovic, S.

1996-01-01

One of the important lessons from the Chernobyl accident is the necessity of existence of operational national emergency response plan. Summarizing consequences and experiences after Chernobyl accident, expert groups from IAEA, ICRP and other international scientific organizations, have been extensively worked on reviewing old ones, and preparing new radiation protection and nuclear safety principals and codes. One of the important issue is national emergency response plan for radiological accident. The nuclear accident response plan in Yugoslavia is presented in this paper. It is essentially based on IAEA model national response plan for radiological accident. This model has to be adjusted to the specificity of member states. The optimum society organization for emergency management in the case of accidents in ionizing radiation sources practices is suggested in this paper. Specific characteriztics of Yugoslav state organization relating to accident response are emphasised. (author)

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

International Nuclear Information System (INIS)

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

1983-02-01

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

Radionuclide release calculations for selected severe accident scenarios

International Nuclear Information System (INIS)

Denning, R.S.; Leonard, M.T.; Cybulskis, P.; Lee, K.W.; Kelly, R.F.; Jordan, H.; Schumacher, P.M.; Curtis, L.A.

1990-08-01

This report provides the results of source term calculations that were performed in support of the NUREG-1150 study. ''Severe Accident Risks: An Assessment for Five US Nuclear Power Plants.'' This is the sixth volume of a series of reports. It supplements results presented in the earlier volumes. Analyses were performed for three of the NUREG-1150 plants: Peach Bottom, a Mark I, boiling water reactor; Surry, a subatmospheric containment, pressurized water reactor; and Sequoyah, an ice condenser containment, pressurized water reactor. Complete source term results are presented for the following sequences: short term station blackout with failure of the ADS system in the Peach Bottom plant; station blackout with a pump seal LOCA for the Surry plant; station blackout with a pump seal LOCA in the Sequoyah plant; and a very small break with loss of ECC and spray recirculation in the Sequoyah plant. In addition, some partial analyses were performed which did not require running all of the modules of the Source Term Code Package. A series of MARCH3 analyses were performed for the Surry and Sequoyah plants to evaluate the effects of alternative emergency operating procedures involving primary and secondary depressurization on the progress of the accident. Only thermal-hydraulic results are provided for these analyses. In addition, three accident sequences were analyzed for the Surry plant for accident-induced failure of steam generator tubes. In these analyses, only the transport of radionuclides within the primary system and failed steam generator were examined. The release of radionuclides to the environment is presented for the phase of the accident preceding vessel meltthrough. 17 refs., 176 figs., 113 tabs

Explanation of procedure on site medical emergency response for nuclear accident

International Nuclear Information System (INIS)

Liu Yulong; Jiang Zhong

2012-01-01

National occupational health standard-Procedure on Site Medical Emergency Response for Nuclear Accident has been approved and issued by the Ministry of Health. This standard is formulated according to the Emergency Response Law of the People's Republic of China, Law of the People 's Republic of China on Prevention and Control of Occupational Diseases, Regulations on Emergency Measures for Nuclear Accidents at Nuclear Power Plants, and Health Emergency Plans for Nuclear and Radiological Accidents of Ministry of Health, supporting the use of On-site Medical Emergency Planning and Preparedness for Nuclear Accidents and Off-site Medical Emergency Planning and Preparedness for Nuclear Accidents. Nuclear accident on-site medical response procedure is a part of the on-site emergency plan. The standard specifies the basic content and requirements of the nuclear accident on-site medical emergency response procedures of nuclear facilities operating units to guide and regulate the work of nuclear accident on-site medical emergency response of nuclear facilities operating units. The criteria-related contents were interpreted in this article. (authors)

The effect of gamma-ray transport on afterheat calculations for accident analysis

International Nuclear Information System (INIS)

Reyes, S.; Latkowski, J.F.; Sanz, J.

2000-01-01

Radioactive afterheat is an important source term for the release of radionuclides in fusion systems under accident conditions. Heat transfer calculations are used to determine time-temperature histories in regions of interest, but the true source term needs to be the effective afterheat, which considers the transport of penetrating gamma rays. Without consideration of photon transport, accident temperatures may be overestimated in others. The importance of this effect is demonstrated for a simple, one-dimensional problem. The significance of this effect depends strongly on the accident scenario being analyzed

Calculation of ex-core detector responses

Energy Technology Data Exchange (ETDEWEB)

Wouters, R. de; Haedens, M. [Tractebel Engineering, Brussels (Belgium); Baenst, H. de [Electrabel, Brussels (Belgium)

2005-07-01

The purpose of this work carried out by Tractebel Engineering, is to develop and validate a method for predicting the ex-core detector responses in the NPPs operated by Electrabel. Practical applications are: prediction of ex-core calibration coefficients for startup power ascension, replacement of xenon transients by theoretical predictions, and analysis of a Rod Drop Accident. The neutron diffusion program PANTHER calculates node-integrated fission sources which are combined with nodal importance representing the contribution of a neutron born in that node to the ex-core response. These importance are computed with the Monte Carlo program MCBEND in adjoint mode, with a model of the whole core at full power. Other core conditions are treated using sensitivities of the ex-core responses to water densities, computed with forward Monte Carlo. The Scaling Factors (SF), or ratios of the measured currents to the calculated response, have been established on a total of 550 in-core flux maps taken in four NPPs. The method has been applied to 15 startup transients, using the average SF obtained from previous cycles, and to 28 xenon transients, using the SF obtained from the in-core map immediately preceding the transient. The values of power (P) and axial offset (AOi) reconstructed with the theoretical calibration agree well with the measured values. The ex-core responses calculated during a rod drop transient have been successfully compared with available measurements, and with theoretical data obtained by alternative methods. In conclusion, the method is adequate for the practical applications previously listed. (authors)

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

International Nuclear Information System (INIS)

Kelsey, Charles T. IV

2011-01-01

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

Reconfigurable mobile manipulation for accident response

International Nuclear Information System (INIS)

Anderson, Robert J.; Morse, William D.; Shirey, David L.; Cdebaca, DanielL M.; Hoffman, John P. Jr.; Lucy, William E.

2000-01-01

The need for a telerobotic vehicle with hazard sensing and integral manipulation capabilities has been identified for use in transportation accidents where nuclear weapons are involved. The Accident Response Mobile Manipulation System (ARMMS) platform has been developed to provide remote dexterous manipulation and hazard sensing for the Accident Response Group (ARG) at Sandia National Laboratories. The ARMMS' mobility platform is a military HMMWV [High Mobility Multipurpose Wheeled Vehicle] that is teleoperated over RF or Fiber Optic communication channels. ARMMS is equipped with two high strength Schilling Titan II manipulators and a suite of hazardous gas and radiation sensors. Recently, a modular telerobotic control architecture call SMART (Sandia Modular Architecture for Robotic and Teleoperation) has been applied to ARMMS. SMART enables input devices and many system behaviors to be rapidly configured in the field for specific mission needs. This paper summarizes current SMART developments applied to ARMMS

Improvements in the nuclear accident response system in Brazil

International Nuclear Information System (INIS)

Estrada, J.J.S.; Azevedo, E.M.; Knofel, T.M.J.; Recio, J.C.A.; Alves, R.N.

1998-01-01

The National Commission on Nuclear Energy has been making outstanding effort to improve its nuclear and radiological accident response systems since the tragic accident in Goiania. Most of this effort is related to nuclear area although the radiological accident has been also considered. This paper describes the improvements in the CNEN response system structure, discusses several topics involving those related to emergency planning and preparedness, and points out some deficiencies that need to be corrected also. The situation during the Goiania accident was more disadvantageous than nowadays, so it is believed that none of the actual deficiencies are sufficient to guess that the population and the environment will not be protected in case of a nuclear or radiological accident

Multi-objective evolutionary emergency response optimization for major accidents

International Nuclear Information System (INIS)

Georgiadou, Paraskevi S.; Papazoglou, Ioannis A.; Kiranoudis, Chris T.; Markatos, Nikolaos C.

2010-01-01

Emergency response planning in case of a major accident (hazardous material event, nuclear accident) is very important for the protection of the public and workers' safety and health. In this context, several protective actions can be performed, such as, evacuation of an area; protection of the population in buildings; and use of personal protective equipment. The best solution is not unique when multiple criteria are taken into consideration (e.g. health consequences, social disruption, economic cost). This paper presents a methodology for multi-objective optimization of emergency response planning in case of a major accident. The emergency policy with regards to protective actions to be implemented is optimized. An evolutionary algorithm has been used as the optimization tool. Case studies demonstrating the methodology and its application in emergency response decision-making in case of accidents related to hazardous materials installations are presented. However, the methodology with appropriate modification is suitable for supporting decisions in assessing emergency response procedures in other cases (nuclear accidents, transportation of hazardous materials) or for land-use planning issues.

Development of the model for the stress calculation of fuel assembly under accident load

International Nuclear Information System (INIS)

Kim, Il Kon

1993-01-01

The finite element model for the stress calculation in guide thimbles of a fuel assembly (FA) under seismic and loss-of-coolant-accident (LOCA) load is developed. For the stress calculation of FA under accident load, at first the program MAIN is developed to select the worst bending mode shaped FA from core model. And then the model for the stress calculation of FA is developed by means of the finite element code. The calculated results of program MAIN are used as the kinematic constraints of the finite element model of a FA. Compared the calculated results of the stiffness of the finite element model of FA with the test results they have good agreements. (Author)

Method for consequence calculations for servere accidents

International Nuclear Information System (INIS)

Nielsen, F.

1987-01-01

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

Fukushima nuclear accident and the social responsibility of science

International Nuclear Information System (INIS)

Yoshioka, Hitoshi

2011-01-01

Five months had passed since Fukushima Daiichi Nuclear Power Plant (NPP) accident occurred but still there was no knowing when the accident ended. Released radioactivity seemed to be greater than one million terra Bq and if there occurred an explosive rupture of containment vessel due to the failure of containment vent or occurrence of steam explosion, the amount of released radioactivity might amount to be at least equivalent to or surpass that of Chernobyl NPP accident. There existed still a risk that overheating and meltdown of nuclear fuels might reoccur with loss of cooling due to a possible giant aftershock. This article described total views on significant disaster that the accident brought about on many neighboring residents or wide range of people. After a general discussion about what was social responsibility of scientists, social responsibility of scientists for Fukushima Daiichi NPP accident was discussed. Responsibility of omission was also argued. (T. Tanaka)

Preliminary evaluation of the Accident Response Mobile Manipulation System for accident site salvage operations

International Nuclear Information System (INIS)

Trujillo, J.M.; Morse, W.D.; Jones, D.P.

1994-01-01

This paper describes and evaluates operational experiences with the Accident Response Mobile Manipulation System (ARMMS) during simulated accident site salvage operations which might involve nuclear weapons. The ARMMS is based upon a teleoperated mobility platform with two Schilling Titan 7F Manipulators

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

International Nuclear Information System (INIS)

SANDGREN, K.R.

2006-01-01

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

Response to the Chernobyl accident in Japan

International Nuclear Information System (INIS)

Anon.

1986-01-01

The worst nuclear accident in history happened at No.4 unit of the Chernobyl Atomic Power Station in USSR. Since the Chernobyl accident, a number of measures have been introduced in many countries, including the reconsideration of programs for construction and operation of nuclear power plants. In Japan, the press and television first reported the accident on April 29. The next day, all the relevant governmental agencies began to collect and analyze information in order to prepare possible countermeasures. The Nuclear Safety Commission issued a statement covering three points: 1) the radioactive substances released by the accident will have virtually no influence on the health of people in Japan, 2) a Special Committee on the Chernobyl Atomic Power Station Accident will be established, and 3) the Soviet government must provide all detailed information about the accident as soon as it is available. On April 30, the Committee on Radioactivity decided to increase radioactivity observations by the Science and Technology Agency, the Defence Agency, and the Meteorological Agency. On the same day, the Ministry of International Trade and Industry set up a survey committee for the Chernobyl accident with the responsibility of collecting and analyzing information about the accident. A review is also made in this article as to how the Japanese media reported the accident and how people reacted on reading the newspapers and watching TV on the accident. (Nogami, K.)

Application of uncertainty analysis method for calculations of accident conditions for RP AES-2006

International Nuclear Information System (INIS)

Zajtsev, S.I.; Bykov, M.A.; Zakutaev, M.O.; Siryapin, V.N.; Petkevich, I.G.; Siryapin, N.V.; Borisov, S.L.; Kozlachkov, A.N.

2015-01-01

An analysis of some accidents using the uncertainly assessment methods is given. The list of the variable parameters incorporated the model parameters of the computer codes, initial and boundary conditions of reactor plant, neutronics. On the basis of the performed calculations of the accident conditions using the statistical method, errors assessment is presented in the determination of the main parameters comparable with the acceptance criteria. It was shown that in the investigated accidents the values of the calculated parameters with account for their error obtained from TRAP-KS and KORSAR/GP Codes do not exceed the established acceptance criteria. Besides, these values do not exceed the values obtained in the conservative calculations. A possibility in principle of the actual application of the method of estimation of uncertainty was shown to justify the safety of WWER AES-2006 using the thermal-physical codes KORSAR/GP and TRAP-KS, PANDA and SUSA programs [ru

The early medical response to the Goiania accident

International Nuclear Information System (INIS)

Valverde, N.J.; Oliveira, A.R.

2000-01-01

The Goiania accident was the most severe radiological one that ever happened in the western hemisphere. The response to its human, social, environmental, economical and psychological burdens represented a huge challenge. Thanks to a multi-institutional intervention the consequences of the accident were greatly minimised. The medical response followed the same pattern and was based on a three-level system of progressive assistance. The early medical response encompassed medical and 'radiological' triage, admission to a specially prepared ward of a local hospital and treatment at a reference center in Rio de Janeiro. (author)

Transport accident emergency response plan

International Nuclear Information System (INIS)

Vallette-Fontaine, M.; Frantz, P.

1998-01-01

To comply with the IAEA recommendations for the implementation of an Emergency Response Plan as described in Safety Series 87, Transnucleaire, a company deeply involved in the road and rail transports of the fuel cycle, masters means of Emergency Response in the event of a transport accident. This paper aims at analyzing the solutions adopted for the implementation of an Emergency Response Plan and the development of a technical support and adapted means for the recovery of heavy packagings. (authors)

ARAC response to the Chernobyl reactor accident

International Nuclear Information System (INIS)

Dickerson, M.H.; Sullivan, T.J.

1986-07-01

This report summarizes the assessments provided by ARAC during the first two weeks after the Chernobyl reactor accident began. Results of this work and measurements made by European countries during that same period show that no major short-term acute health effects would be expected in Europe as a result of this accident. Statistical long-term health effects were not addressed in these studies. Both measured and calculated I-131 concentrations in milk in the US were over an order of magnitude below the USDA guideline of 15,000 pCi/l

Fuel models and results from the TRAC-PF1/MIMAS TMI-2 accident calculation

International Nuclear Information System (INIS)

Schwegler, E.C.; Maudlin, P.J.

1983-01-01

A brief description of several fuel models used in the TRAC-PF1/MIMAS analysis of the TMI-2 accident is presented, and some of the significant fuel-rod behavior results from this analysis are given. Peak fuel-rod temperatures, oxidation heat production, and embrittlement and failure behavior calculated for the TMI-2 accident are discussed. Other aspects of fuel behavior, such as cladding ballooning and fuel-cladding eutectic formation, were found not to significantly affect the accident progression

The passive response of the Integral Fast Reactor concept to the chilled inlet accident

International Nuclear Information System (INIS)

Vilim, R.B.

1990-01-01

Simple methods are described for bounding the passive response of a metal fueled liquid-metal cooled reactor to the chilled inlet accident. Calculation of these bounds for a prototype of the Integral Fast Reactor concept shows that failure limits --- eutectic melting, sodium boiling and fuel pin failure --- are not exceeded. 2 refs., 1 fig., 2 tabs

Severe accident management guidelines tool

International Nuclear Information System (INIS)

Gutierrez Varela, Javier; Tanarro Onrubia, Augustin; Martinez Fanegas, Rafael

2014-01-01

Severe Accident is addressed by means of a great number of documents such as guidelines, calculation aids and diagnostic trees. The response methodology often requires the use of several documents at the same time while Technical Support Centre members need to assess the appropriate set of equipment within the adequate mitigation strategies. In order to facilitate the response, TECNATOM has developed SAMG TOOL, initially named GGAS TOOL, which is an easy to use computer program that clearly improves and accelerates the severe accident management. The software is designed with powerful features that allow the users to focus on the decision-making process. Consequently, SAMG TOOL significantly improves the severe accident training, ensuring a better response under a real situation. The software is already installed in several Spanish Nuclear Power Plants and trainees claim that the methodology can be followed easier with it, especially because guidelines, calculation aids, equipment information and strategies availability can be accessed immediately (authors)

Japanese Nuclear Accident and U.S. Response

International Nuclear Information System (INIS)

Douet, Randy

2011-01-01

U.S. Government response to the Fukushima accident: • Multi-agency task force (Nuclear Regulatory Commission, Department of Energy, Department of Defense) supporting Japan recovery efforts; • President Obama directed the NRC to perform a comprehensive review of U.S. reactors; • NRC established agency task force to develop lessons learned from Fukushima Daiichi accident to provide short-term and long-term analysis of the events

Calculation of fission product behavior in a multiple reactor barriers in case of an accident

International Nuclear Information System (INIS)

Ezzedin, A. A.; Dakhil, A. S.; Elbaden, S. E.

2012-12-01

Radiation protection of the population in case of a reactor accident utilizes reference levels which are based on doses values. Therefore, adequate provisions for effective and timely dose assessment for population in case of accidents at nuclear power plant (NPP) are important. Developing the background for such provisions is the objective of this study. In particular, an exponential model has been developed and utilized to calculate the release rate of the most volatile gaseous materials from different reactor barriers. Calculation has been performed for noble gases (1 33X e, 1 35X e, 1 38X e, 8 5K r, 8 7K r, 8 8K r) and the halogens(1'3 1I , 1 32I , 1 33I , 1'3 4I , 1 35I ). The effective dose rate equivalent is calculations in the nearly stage of a reactor accident. Calculations are performed using the MCNP-4C code. The results are comparable with the final analysis report which utilizes different codes. Results of our calculation shows no excessive dose in populated regions and it is recommended to use secondary containment barrier for highly reduction of the release rate to the environment. (Author)

Analysis of Construction Accidents in Turkey and Responsible Parties

Science.gov (United States)

GÜRCANLI, G. Emre; MÜNGEN, Uğur

2013-01-01

Construction is one of the world’s biggest industry that includes jobs as diverse as building, civil engineering, demolition, renovation, repair and maintenance. Construction workers are exposed to a wide variety of hazards. This study analyzes 1,117 expert witness reports which were submitted to criminal and labour courts. These reports are from all regions of the country and cover the period 1972–2008. Accidents were classified by the consequence of the incident, time and main causes of the accident, construction type, occupation of the victim, activity at time of the accident and party responsible for the accident. Falls (54.1%), struck by thrown/falling object (12.9%), structural collapses (9.9%) and electrocutions (7.5%) rank first four places. The accidents were most likely between the hours 15:00 and 17:00 (22.6%), 10:00–12:00 (18.7%) and just after the lunchtime (9.9%). Additionally, the most common accidents were further divided into sub-types. Expert-witness assessments were used to identify the parties at fault and what acts of negligence typically lead to accidents. Nearly two thirds of the faulty and negligent acts are carried out by the employers and employees are responsible for almost one third of all cases. PMID:24077446

Analysis of construction accidents in Turkey and responsible parties.

Science.gov (United States)

Gürcanli, G Emre; Müngen, Uğur

2013-01-01

Construction is one of the world's biggest industry that includes jobs as diverse as building, civil engineering, demolition, renovation, repair and maintenance. Construction workers are exposed to a wide variety of hazards. This study analyzes 1,117 expert witness reports which were submitted to criminal and labour courts. These reports are from all regions of the country and cover the period 1972-2008. Accidents were classified by the consequence of the incident, time and main causes of the accident, construction type, occupation of the victim, activity at time of the accident and party responsible for the accident. Falls (54.1%), struck by thrown/falling object (12.9%), structural collapses (9.9%) and electrocutions (7.5%) rank first four places. The accidents were most likely between the hours 15:00 and 17:00 (22.6%), 10:00-12:00 (18.7%) and just after the lunchtime (9.9%). Additionally, the most common accidents were further divided into sub-types. Expert-witness assessments were used to identify the parties at fault and what acts of negligence typically lead to accidents. Nearly two thirds of the faulty and negligent acts are carried out by the employers and employees are responsible for almost one third of all cases.

Burden of road traffic accidents in Nepal by calculating disability-adjusted life years

Directory of Open Access Journals (Sweden)

Ling Huang

2017-10-01

Full Text Available Objective: To calculate the burden of road traffic accidents in Kathmandu Valley and then extrapolate this to the national level. Methods: A prospective study was performed to compute the burden of road traffic accidents by quantification of disability-adjusted life years (DALYs using the Global Burden of Disease Study method on the basis of 1-year data from nine hospitals in Nepal and the Department of Forensic Medicine and cross-checked with the Nepal Traffic Directorate. Multiple methods were applied to the extrapolated population metrics of the burden of road traffic accidents in Nepal. Results: The total number of years of life lived in disability, years of life lost, and DALYs in Nepal were 38,848±194, 119,935±1464, and 158,783±1658 (95% confidence interval respectively. The number of years lost because of morbidity and death was similar in Kathmandu Valley. Most (75% of the DALYs resulted from years of life lost in Nepal. Males accounted for 73% of DALYs. Almost half (44% of the DALYs were contributed by the group aged 15–29 years. Conclusion: This study is the first to calculate the burden of road traffic accidents in Nepal using Nepal’s own data. Nepal needs to develop and enhance its own system to identify significant public health issues so as to set national priorities for prevention of road traffic accidents.

Response of HEPA filters to simulated-accident conditions

International Nuclear Information System (INIS)

Gregory, W.S.; Martin, R.A.; Smith, P.R.; Fenton, D.E.

1982-01-01

High-efficiency particulate air (HEPA) filters have been subjected to simulated accident conditions to determine their response to abnormal operating events. Both domestic and European standard and high-capacity filters have been evaluated to determine their response to simulated fire, explosion, and tornado conditions. The HEPA filter structural limitations for tornado and explosive loadings are discussed. In addition, filtration efficiencies during these accident conditions are reported for the first time. Our data indicate efficiencies between 80% and 90% for shock loadings below the structural limit level. We describe two types of testing for ineffective filtration - clean filters exposed to pulse-entrained aerosol and dirty filters exposed to tornado and shock pulses. Efficiency and material loss data are described. Also, the resonse of standard HEPA filters to simulated fire conditions is presented. We describe a unique method of measuring accumulated combustion products on the filter. Additionally, data relating to pressure drop vs accumulated mass during plugging are reported for simulated combustion aerosols. The effects of concentration and moisture levels on filter plugging were evaluated. We are obtaining all of the above data so that mathematical models can be developed for fire, explosion, and tornado accident analysis computer codes. These computer codes can be used to assess the response of nuclear air cleaning systems to accident conditions

Planning and Preparing for Emergency Response to Transport Accidents Involving Radioactive Material. Safety Guide

International Nuclear Information System (INIS)

2009-01-01

This Safety Guide provides guidance on various aspects of emergency planning and preparedness for dealing effectively and safely with transport accidents involving radioactive material, including the assignment of responsibilities. It reflects the requirements specified in Safety Standards Series No. TS-R-1, Regulations for the Safe Transport of Radioactive Material, and those of Safety Series No. 115, International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. Contents: 1. Introduction; 2. Framework for planning and preparing for response to accidents in the transport of radioactive material; 3. Responsibilities for planning and preparing for response to accidents in the transport of radioactive material; 4. Planning for response to accidents in the transport of radioactive material; 5. Preparing for response to accidents in the transport of radioactive material; Appendix I: Features of the transport regulations influencing emergency response to transport accidents; Appendix II: Preliminary emergency response reference matrix; Appendix III: Guide to suitable instrumentation; Appendix IV: Overview of emergency management for a transport accident involving radioactive material; Appendix V: Examples of response to transport accidents; Appendix VI: Example equipment kit for a radiation protection team; Annex I: Example of guidance on emergency response to carriers; Annex II: Emergency response guide.

A calculational methodology for comparing the accident, occupational, and waste-disposal hazards of fusion reactor designs

International Nuclear Information System (INIS)

Fetter, S.

1985-01-01

A methodology has been developed for calculating indices of three classes of radiological hazards: reactor accidents, occupational exposures, and waste-disposal hazards. Radionuclide inventories, biological hazard potentials (BHP), and various dose-related indices are calculated. In the case of reactor accidents, the critical, 50-year and chronic dose are computed, as well as the number of early deaths and illnesses and late cancer fatalities. For occupational exposure, the contact dose rate is calculated for several times after reactor shutdown. In the case of waste-disposal hazards, the intruder dose and the intruder hazard potential (IHP) are calculated. Sample calculations for the MARS reactor design show the usefulness of the methodology in exploring design improvements

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)

System response of a DOE Defense Program package in a transportation accident environment

International Nuclear Information System (INIS)

Chen, T.F.; Hovingh, J.; Kimura, C.Y.

1992-01-01

The system response in a transportation accident environment is an element to be considered in an overall Transportation System Risk Assessment (TSRA) framework. The system response analysis uses the accident conditions and the subsequent accident progression analysis to develop the accident source term, which in turn, is used in the consequence analysis. This paper proposes a methodology for the preparation of the system response aspect of the TSRA

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)

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

International Nuclear Information System (INIS)

Adomavicius, A.; Belousov, A.; Ognerubov, V.

2004-01-01

Background of thermo hydraulic processes audit calculations in the frame of RSR-2 project is presented. Assumptions for the design based accident - RBMK-1500 group distributor header break analysis and modeling are presented. Audit calculations by ATHLET code and evaluation of results were provided. (author)

National response plan - Major nuclear or radiological accidents

International Nuclear Information System (INIS)

2014-02-01

France has been implementing stringent radiation protection and nuclear safety and security measures for many years. However, this does not mean that the country is exempt from having to be prepared to deal with an emergency. Changes in France, Europe and other parts of the globe have made it necessary for France to reconsider how it responds to nuclear and radiological emergencies. As the potential impact of a nuclear or radiological accident can affect a wide range of activities, the plan described herein is based on a cross-sector and inter-ministerial approach to emergency response. The Chernobyl and Fukushima-Daiichi disasters are proof that the consequences of a major nuclear or radiological accident can affect all levels of society. These challenges are substantial and relate to: public health: An uncontrolled nuclear accident can have immediate consequences (death, injury, irradiation) as well as long-term consequences that can lead to increased risk of developing radiation-induced diseases (such as certain types of cancer); environmental quality: Radiation contamination can last for several decades and, in some cases, can result in an area being closed off permanently to the public; economic and social continuity: Nuclear accidents bring human activity to a halt in contaminated areas, disrupting the economic and social order of the entire country. It may therefore be necessary to adapt economic and social systems and carry out clean-up operations if people and businesses have been displaced; quality of international relations: Related to fulfillment of obligations to alert and inform European and international partners. This international dimension also covers the protection of French nationals present in countries stricken by a nuclear accident. This national plan provides reference information on how to prepare for a nuclear or radiological emergency and make the appropriate decisions in the event of an emergency. It covers the emergency phase (including

Soviet medical response to the Chernobyl nuclear accident

International Nuclear Information System (INIS)

Linnemann, R.E.

1987-01-01

The nuclear accident at Chernobyl was the worst in the history of nuclear power. It tested the organized medical response to mass radiation casualties. This article reviews the Soviet response as reported at the 1986 postaccident review meeting in Vienna and as determined from interviews. The Soviets used three levels of care: rescue and first aid at the plant site; emergency treatment at regional hospitals; and definitive evaluation and treatment in Moscow. Diagnosis, triage, patient disposition, attendant exposure, and preventive actions are detailed. The United States would be well advised to organize its resources definitively to cope with future nonmilitary nuclear accidents

Perception of risk and the attribution of responsibility for accidents.

Science.gov (United States)

Rickard, Laura N

2014-03-01

Accidents, one often hears, "happen"; we accept, and even expect, that they will be part of daily life. But in situations in which injury or death result, judgments of responsibility become critical. How might our perceptions of risk influence the ways in which we allocate responsibility for an accident? Drawing from attribution and risk perception theory, this study investigates how perceived controllability and desirability of risk, in addition to perceived danger and recreational risk-taking, relate to attributions of responsibility for the cause of unintentional injury in a unique setting: U.S. national parks. Three parks, Mount Rainier, Olympic, and Delaware Water Gap, provide the setting for this survey-based study, which considers how park visitors (N = 447) attribute responsibility for the cause of a hypothetical visitor accident. Results suggest that respondents tended to make more internal (i.e., related to characteristics of the victim), rather than external (i.e., related to characteristics of the park, or park management) attributions. As respondents viewed park-related risk as controllable, they were more likely to attribute the cause of the accident to the victim. Moreover, among other significant variables, having experienced a similar accident predicted lower internal causal attribution. Opportunities for future research linking risk perception and attribution variables, as well as practical implications for the management of public outdoor settings, are presented. © 2013 Society for Risk Analysis.

Conservatism in effective dose calculations for accident events involving fuel reprocessing waste tanks.

Science.gov (United States)

Bevelacqua, J J

2011-07-01

Conservatism in the calculation of the effective dose following an airborne release from an accident involving a fuel reprocessing waste tank is examined. Within the regulatory constraints at the Hanford Site, deterministic effective dose calculations are conservative by at least an order of magnitude. Deterministic calculations should be used with caution in reaching decisions associated with required safety systems and mitigation philosophy related to the accidental release of airborne radioactive material to the environment.

Supporting system in emergency response plan for nuclear material transport accidents

International Nuclear Information System (INIS)

Nakagome, Y.; Aoki, S.

1993-01-01

As aiming to provide the detailed information concerning nuclear material transport accidents and to supply it to the concerned organizations by an online computer, the Emergency Response Supporting System has been constructed in the Nuclear Safety Technology Center, Japan. The system consists of four subsystems and four data bases. By inputting initial information such as name of package and date of accident, one can obtain the appropriate initial response procedures and related information for the accident immediately. The system must be useful for protecting the public safety from nuclear material transport accidents. But, it is not expected that the system shall be used in future. (J.P.N.)

Cognitive modeling and dynamic probabilistic simulation of operating crew response to complex system accidents

International Nuclear Information System (INIS)

Chang, Y.H.J.; Mosleh, A.

2007-01-01

This is the third in a series of five papers describing the IDAC (Information, Decision, and Action in Crew context) model for human reliability analysis. An example application of this modeling technique is also discussed in this series. The model is developed to probabilistically predict the responses of the nuclear power plant control room operating crew in accident conditions. The operator response spectrum includes cognitive, emotional, and physical activities during the course of an accident. This paper discusses the modeling components and their process rules. An operator's problem-solving process is divided into three types: information pre-processing (I), diagnosis and decision-making (D), and action execution (A). Explicit and context-dependent behavior rules for each type of operator are developed in the form of tables, and logical or mathematical relations. These regulate the process and activities of each of the three types of response. The behavior rules are developed for three generic types of operator: Decision Maker, Action Taker, and Consultant. This paper also provides a simple approach to calculating normalized probabilities of alternative behaviors given a context

DEFORM-4: fuel pin characterization and transient response in the SAS4A accident analysis code system

International Nuclear Information System (INIS)

Miles, K.J.; Hill, D.J.

1986-01-01

The DEFORM-4 module is the segment of the SAS4A Accident Analysis Code System that calculates the fuel pin characterization in response to a steady state irradiation history, thereby providing the initial conditions for the transient calculation. The various phenomena considered include fuel porosity migration, fission gas bubble induced swelling, fuel cracking and healing, fission gas release, cladding swelling, and the thermal-mechanical state of the fuel and cladding. In the transient state, the module continues the thermal-mechanical response calculation, including fuel melting and central cavity pressurization, until cladding failure is predicted and one of the failed fuel modules is initiated. Comparisons with experimental data have demonstrated the validity of the modeling approach

The Fukushima Daiichi Accident. Technical Volume 3/5. Emergency Preparedness and Response

International Nuclear Information System (INIS)

2015-08-01

This volume describes the key events and response actions from the onset of the accident at the Fukushima Daiichi nuclear power plant (NPP), operated by the Tokyo Electric Power Company (TEPCO), on 11 March 2011. It also describes the national emergency preparedness and response (EPR) system in place in Japan and the international EPR framework prior to the accident. It is divided into five sections. Section 3.1 describes the initial actions taken by Japan in response to the accident, involving: identification of the accident, notification of off-site authorities and activation of the response; mitigatory actions taken on-site; and initial off-site response. Section 3.2 describes the protective measures taken for personnel in response to the natural disaster, protection of emergency workers, medical management of emergency workers and the voluntary involvement of members of the public in the emergency response. Section 3.3 describes the protective actions and other response actions taken by Japan to protect the public. It addresses urgent and early protective actions; the use of a dose projection model, the System for Prediction of Environmental Emergency Dose Information (SPEEDI), as a basis for decisions on protective actions during the accident; environmental monitoring; provision of information to the public and international community; and issues related to international trade and waste management. Section 3.4 describes the transition from the emergency phase to the recovery phase. It also addresses the national analysis of the accident and the emergency response. Section 3.5 describes the response by the IAEA, other international organizations within the Inter- Agency Committee on Radiological and Nuclear Emergencies (IACRNE), the actions of IAEA Member States with regard to protective actions recommended to their nationals in Japan and the provision of international assistance. A summary, observations and lessons conclude each section. There are three

Emergency response plan for accidents in Saudi Arabia

International Nuclear Information System (INIS)

Al-Solaiman, K.M.; Al-Arfaj, A.M.; Farouk, M.A.

2000-01-01

This paper presents a brief description of the general emergency plan for accidents involving radioactive materials in the Kingdom of Saudi Arabia. Uses of radioactive materials and radiation sources and their associated potential accident are specified. Most general accident scenarios of various levels have been determined. Protective measures have been specified to reduce individual and collective doses arising during accident situations. Intervention levels for temporary exposure situations, as established in the IAEA's basic safety standards for protection against ionising radiation and for the safety of radiation sources, are adopted as national intervention levels. General procedures for implementation of the response plan, including notification and radiological monitoring instrumentation and equipment, are described and radiation monitoring teams are nominated. Training programs for the different parties which may be called upon to respond are studied and will be started. (author)

The risk of a major nuclear accident: calculation and perception of probabilities

International Nuclear Information System (INIS)

Leveque, Francois

2013-07-01

The accident at Fukushima Daiichi, Japan, occurred on 11 March 2011. This nuclear disaster, the third on such a scale, left a lasting mark in the minds of hundreds of millions of people. Much as Three Mile Island or Chernobyl, yet another place will be permanently associated with a nuclear power plant which went out of control. Fukushima Daiichi revived the issue of the hazards of civil nuclear power, stirring up all the associated passion and emotion. The whole of this paper is devoted to the risk of a major nuclear accident. By this we mean a failure initiating core meltdown, a situation in which the fuel rods melt and mix with the metal in their cladding. Such accidents are classified as at least level 5 on the International Nuclear Event Scale. The Three Mile Island accident, which occurred in 1979 in the United States, reached this level of severity. The explosion of reactor 4 at the Chernobyl plant in Ukraine in 1986 and the recent accident in Japan were classified as class 7, the highest grade on this logarithmic scale. The main difference between the top two levels and level 5 relates to a significant or major release of radioactive material to the environment. In the event of a level-5 accident, damage is restricted to the inside of the plant, whereas, in the case of level-7 accidents, huge areas of land, above or below the surface, and/or sea may be contaminated. Before the meltdown of reactors 1, 2 and 3 at Fukushima Daiichi, eight major accidents affecting nuclear power plants had occurred worldwide. This is a high figure compared with the one calculated by the experts. Observations in the field do not appear to fit the results of the probabilistic models of nuclear accidents produced since the 1970's. Oddly enough the number of major accidents is closer to the risk as perceived by the general public. In general we tend to overestimate any risk relating to rare, fearsome accidents. What are we to make of this divergence? How are we to reconcile

Radiation protection issues on preparedness and response for a severe nuclear accident: experiences of the Fukushima accident.

Science.gov (United States)

Homma, T; Takahara, S; Kimura, M; Kinase, S

2015-06-01

Radiation protection issues on preparedness and response for a severe nuclear accident are discussed in this paper based on the experiences following the accident at Fukushima Daiichi nuclear power plant. The criteria for use in nuclear emergencies in the Japanese emergency preparedness guide were based on the recommendations of International Commission of Radiological Protection (ICRP) Publications 60 and 63. Although the decision-making process for implementing protective actions relied heavily on computer-based predictive models prior to the accident, urgent protective actions, such as evacuation and sheltering, were implemented effectively based on the plant conditions. As there were no recommendations and criteria for long-term protective actions in the emergency preparedness guide, the recommendations of ICRP Publications 103, 109, and 111 were taken into consideration in determining the temporary relocation of inhabitants of heavily contaminated areas. These recommendations were very useful in deciding the emergency protective actions to take in the early stages of the Fukushima accident. However, some suggestions have been made for improving emergency preparedness and response in the early stages of a severe nuclear accident. © The Chartered Institution of Building Services Engineers 2014.

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

Coupled hydro-neutronic calculations for fast burst reactor accidents

International Nuclear Information System (INIS)

Paternoster, R.; Kimpland, R.; Jaegers, P.; McGhee, J.

1994-01-01

Methods are described for determining the fully coupled neutronic/hydrodynamic response of fast burst reactors (FBR) under disruptive accident conditions. Two code systems, PAD (1 -D Lagrangian) and NIKE-PAGOSA (3-D Eulerian) were used to accomplish this. This is in contrast to the typical methodology that computes these responses by either single point kinetics or in a decoupled manner. This methodology is enabled by the use of modem supercomputers (CM-200). Two examples of this capability are presented: an unreflected metal fast burst assembly, and a reflected fast burst assembly typical of the Skua or SPR-III class of fast burst reactor

Accidents in family forestry's firewood production.

Science.gov (United States)

Lindroos, Ola; Aspman, Emma Wilhelmson; Lidestav, Gun; Neely, Gregory

2008-05-01

Firewood is commonly used around the world, but little is known about the work involved in its production and associated accidents. The objectives were to identify relationships between accidents and time exposure, workers' age and sex, equipment used and work activities in family forestry's firewood production. Data from a postal survey in Northern Sweden were compared to a database of injuries in the same region. Most accidents occurred to 50-69 year old men, who also worked most hours. No significant differences in sex and age were found between expected and recorded accident frequencies when calculated from total work hours; however, when calculated using numbers of active persons significant differences were found for both age and sex. Frequency of accidents per unit worked time was higher for machine involving activities than for other activities. Accidents that occurred when using wedge splitter machines were responsible for most of this overrepresentation. Fingers were the most commonly injured body parts. Mean accident rate for the equipment used was 87 accidents per million work hours, and the rate was highest for wedge splitters (122 accidents per million work hours). Exposure to elevated risks due to violation of safety procedures is discussed, as well as possible preventative measures.

Dose calculation for accident situations at WWR-S type spent nuclear fuel repository

International Nuclear Information System (INIS)

Margeanu, S.; Florescu, G.

2006-01-01

Full text: The Spent Nuclear Fuel Repository at IFIN-HH Bucharest (SNFR IFIN-HH) consists in four pools, repository hall, radiological monitoring system, ventilation system and auxiliary systems. At the moment the remaining activity in the repository is about 3500 Ci. Despite of the small activity, for emergency preparedness purposes, several accident scenarios, with a non zero probability of occurrence during the repository lifetime, have been postulated. Evaluations of radiological consequences to personnel, general public and environment, for each accident scenario have been performed. The radioactive inventory was evaluated with ORIGEN code from SCALE computer code system and radiological consequences were evaluated with COSYMA computer code. Assumptions for the source term determination, meteorological conditions and release, are presented. The calculated values of doses and risk are also presented. The impact of these accident scenarios on population and environment is also discussed. (authors)

Model for calculation of concentration and load on behalf of accidents with radioactive materials

International Nuclear Information System (INIS)

Janssen, L.A.M.; Heugten, W.H.H. van

1987-04-01

In the project 'Information- and calculation-system for disaster combatment', by order of the Dutch government, a demonstration model has been developed for a diagnosis system for accidents. In this demonstration a model is used to calculate the concentration- and dose-distributions caused by incidental emissions of limited time. This model is described in this report. 4 refs.; 2 figs.; 3 tabs

Calculation of individual and population doses on Danish territory resulting from hypothetical core-melt accidents at the Barsebaeck reactor

International Nuclear Information System (INIS)

1977-01-01

Individual and population doses within Danish territory are calculated from hypothetical, severe core-melt accidents at the Swedish nuclear plant at Barsebaeck. The fission product inventory of the Barsebaeck reactor is calculated. The release fractions for the accidents are taken from WASH-1400. Based on parametric studies, doses are calculated for very unfavourable, but not incredible weather conditions. The probability of such conditions in combination with wind direction towards Danish territory is estimated. Doses to bone marrow, lungs, GI-tract and thyroid are calculated based on dose models developed at Risoe. These doses are found to be consistent with doses calculated with the models used in WASH-1400. (author)

The Influence of atmospheric conditions to probabilistic calculation of impact of radiology accident on PWR 1000 MWe

International Nuclear Information System (INIS)

Pande Made Udiyani; Sri Kuntjoro

2015-01-01

The calculation of the radiological impact of the fission products releases due to potential accidents that may occur in the PWR (Pressurized Water Reactor) is required in a probabilistic. The atmospheric conditions greatly contribute to the dispersion of radionuclides in the environment, so that in this study will be analyzed the influence of atmospheric conditions on probabilistic calculation of the reactor accidents consequences. The objective of this study is to conduct an analysis of the influence of atmospheric conditions based on meteorological input data models on the radiological consequences of PWR 1000 MWe accidents. Simulations using PC-Cosyma code with probabilistic calculations mode, the meteorological data input executed cyclic and stratified, the meteorological input data are executed in the cyclic and stratified, and simulated in Muria Peninsula and Serang Coastal. Meteorological data were taken every hour for the duration of the year. The result showed that the cumulative frequency for the same input models for Serang coastal is higher than the Muria Peninsula. For the same site, cumulative frequency on cyclic input models is higher than stratified models. The cyclic models provide flexibility in determining the level of accuracy of calculations and do not require reference data compared to stratified models. The use of cyclic and stratified models involving large amounts of data and calculation repetition will improve the accuracy of statistical calculation values. (author)

National plan of response to a major nuclear or radiological accident

International Nuclear Information System (INIS)

2014-02-01

The first part of this document presents the response strategies and principles to be applied in the case of a major nuclear or radiological accident. It presents the general framework and the 8 reference situations which are used as references for the plan. It presents the general organisation of crisis management by the State (initial organisation, organisation at the national level, communication channel, international channels, case of transport of radioactive materials, responsibility of the various actors). Then, it presents the strategies of response, i.e., a global strategy and more specific strategies applicable in different sectors or fields: for the control of the concerned installation or transport, in the case of transport of radioactive materials, for the protection of the population, for the taking into care, for communication, for the continuity of social and economic life, at the European level, for the post-accidental management. The second part is a guide which contains sheets describing reactions in different situations: uncertainty, accident in an installation resulting in an either immediate and short, or immediate and long, or delayed and long release, accident in a transport of radioactive materials with potential release, accident occurring abroad which may have a more or less significant impact in France, and accident at sea

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

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.

Emergency response planning and preparedness for transport accidents involving radioactive material

International Nuclear Information System (INIS)

1988-01-01

The purpose of this Guide is to provide assistance to public authorities and others (including consignors and carriers of radioactive materials) who are responsible for ensuring safety in establishing and developing emergency response arrangements for responding effectively to transport accidents involving radioactive materials. This Guide is concerned mainly with the preparation of emergency response plans. It provides information which will assist those countries whose involvement with radioactive materials is just beginning and those which have already developed their industries involving radioactive materials and attendant emergency plans, but may need to review and improve these plans. The need for emergency response plans and the ways in which they are implemented vary from country to country. In each country, the responsible authorities must decide how best to apply this Guide, taking into account the actual shipments and associated hazards. In this Guide the emergency response planning and response philosophy are outlined, including identification of emergency response organizations and emergency services that would be required during a transport accident. General consequences which could prevail during an accident are described taking into account the IAEA Regulations for the Safe Transport of Radioactive Material. 43 refs, figs and tabs

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

Fission product releases at severe LWR accident conditions: ORNL/CEA measurements versus calculations

Energy Technology Data Exchange (ETDEWEB)

Andre, B.; Ducros, G.; Leveque, J.P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Osborne, M.F.; Lorenz, R.A. [Oak Ridge National Lab., TN (United States); Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Dept. de Protection de l`Environnement et des Installations

1995-12-31

Experimental programs in the United States and France have followed similar paths in supplying much of the data needed to analyze severe accidents. Both the HI/VI program, conducted at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (NRC), and the HEVA/VERCORS program, supported by IPSN-Commissariat a l`Energie Atomique (CEA) and carried out at the Centre d`Etudes Nucleaires de Grenoble, have studied fission product release from light water reactor (LWR) fuel samples during test sequences representative of severe accidents. Recognizing that more accurate data, i.e., a better defined source term, could reduce the safety margins included in the rather conservative source terms originating from WASH-1400, the primary objective of these programs has been to improve the data base concerning fission product release and behavior at high temperatures. To facilitate the comparison, a model based on fission product diffusion mechanisms that was developed at ORNL and adapted with CEA experimental data is proposed. This CEA model is compared with the ORNL experimental data in a blind test. The two experimental programs used similar techniques in out-of-pile studies. Highly irradiated fuel samples were heated in radiofrequency induction furnaces to very high temperatures (up to 2700 K at ORNL and 2750 K at CEA) in oxidizing (H{sub 2}O), reducing (H{sub 2}) or mixed (H{sub 2}O+H{sub 2}) environments. The experimental parameters, which were chosen from calculated accident scenarios, did not duplicate specific accidents, but rather emphasized careful control of test conditions to facilitate extrapolation of the results to a wide variety of accident situations. This paper presents a broad and consistent database from ORNL and CEA release results obtained independently since the early 1980`S. A comparison of CORSOR and CORSOR Booth calculations, currently used in safety analysis, and the experimental results is presented and

Helicopter emergency medical services response to equestrian accidents.

Science.gov (United States)

Lyon, Richard M; Macauley, Ben; Richardson, Sarah; de Coverly, Richard; Russell, Malcolm

2015-04-01

Horse riding is a common leisure activity associated with a significant rate of injury. Helicopter emergency medical services (HEMS) may be called to equestrian accidents. Accurate HEMS tasking is important to ensure appropriate use of this valuable medical resource. We sought to review HEMS response to equestrian accidents and identify factors associated with the need for HEMS intervention or transport of the patient to a major trauma centre. Retrospective case review of all missions flown by Kent, Surrey & Sussex Air Ambulance Trust over a 1-year period (1 July 2011 to 1 July 2012). All missions were screened for accidents involving a horse. Call details, patient demographics, suspected injuries, clinical interventions and patient disposition were all analysed. In the 12-month data collection period there were 47 equestrian accidents, representing ∼3% of the total annual missions. Of the 42 cases HEMS attended, one patient was pronounced life extinct at the scene. In 15 (36%) cases the patient was airlifted to hospital. In four (10%) cases, the patient underwent prehospital anaesthesia. There were no specific predictors of HEMS intervention. Admission to a major trauma centre was associated with the rider not wearing a helmet, a fall onto their head or the horse falling onto the rider. Equestrian accidents represent a significant proportion of HEMS missions. The majority of patients injured in equestrian accidents do not require HEMS intervention, however, a small proportion have life-threatening injuries, requiring immediate critical intervention. Further research is warranted, particularly regarding HEMS dispatch, to further improve accuracy of tasking to equestrian accidents.

Henri Jammet Memorial lecture: The role of dosimetry in radiation accident response

International Nuclear Information System (INIS)

Ricks, Robert C.; Joiner, Eugene; Toohey, Richard E.; Holloway, Elizabeth C.

1997-01-01

This document presents a lecture given on the role of dosimetry in radiation accident response, focusing accidents such as: Vinca, occurred on october 15, 1958, Goiania Cs-137, Hanford Am-241 and Juarez Co-60, Chernobyl nuclear power plant. Other accidents are reported as they are registered in the REAC/TS Registry

Federal Radiological Monitoring and Assessment Center (FRMAC), US response to major radiological accidents

International Nuclear Information System (INIS)

Mueller, P.G.

2000-01-01

During the 1960's and 70's the expanded use of nuclear materials to generate electricity, to provide medical benefits, and for research purposes continued to grow in the United States. While substantial effort went into constructing plants and facilities and providing for a number of redundant backup systems for safety purposes, little effort went into the development of emergency response plans for possible major radiological accidents. Unfortunately, adequate plans and procedures had not been developed to co-ordinate either state or federal emergency response assets and personnel should a major radiological accident occur. This situation became quite evident following the Three Mile Island Nuclear Reactor accident in 1979. An accident of that magnitude had not been adequately prepared for and Pennsylvania's limited emergency radiological resources and capabilities were quickly exhausted. Several federal agencies with statutory responsibilities for emergency response, including the U.S. Environmental Protection Agency (EPA), U.S. Department of Energy (DOE), Federal Emergency Management Agency (FEMA), Nuclear Regulatory Commission (NRC), and others provided extensive assistance and support during the accident. However, the assistance was not fully co-ordinated nor controlled. Following the Three Mile Island incident 13 federal agencies worked co-operatively to develop an agreement called the Federal Radiological Emergency Response Plan (FRERP). Signed in November 1985, this plan delineated the statutory responsibilities and authorities of each federal agency signatory to the FRERP. In the event of a major radiological accident, the FRERP would be activated to ensure that a co-ordinated federal emergency response would be available to respond to any major radiological accident scenario. The FRERP encompasses a wide variety of radiological accidents, not just those stemming from nuclear power plants. Activation of the FRERP could occur from major accidents involving

Neutron-gamma flux and dose calculations for feasibility study of DISCOMS instrumentation in case of severe accident in a GEN 3 reactor

Science.gov (United States)

Brovchenko, Mariya; Duhamel, Isabelle; Dechenaux, Benjamin

2017-09-01

The present paper presents the study carried out in the frame of the DISCOMS project, which stands for "DIstributed Sensing for COrium Monitoring and Safety". This study concerns the calculation of the neutron and gamma radiations received by the considered instrumentation during the normal reactor operation as well as in case of a severe accident for the EPR reactor, outside the reactor pressure vessel and in the containment basemat. This paper summarizes the methods and hypotheses used for the particle transport simulation outside the vessel during normal reactor operation. The results of the simulations are then presented including the responses for distributed Optical Fiber Sensors (OFS), such as the gamma dose and the fast neutron fluence, and for Self Powered Neutron Detectors (SPNDs), namely the neutron and gamma spectra. Same responses are also evaluated for severe accident situations in order to design the SPNDs being sensitive to the both types of received neutron-gamma radiation. By contrast, fibers, involved as transducers in distributed OFS have to resist to the total radiation gamma dose and neutron fluence received during normal operation and the severe accident.

Uncertainty analysis of accident notification time and emergency medical service response time in work zone traffic accidents.

Science.gov (United States)

Meng, Qiang; Weng, Jinxian

2013-01-01

Taking into account the uncertainty caused by exogenous factors, the accident notification time (ANT) and emergency medical service (EMS) response time were modeled as 2 random variables following the lognormal distribution. Their mean values and standard deviations were respectively formulated as the functions of environmental variables including crash time, road type, weekend, holiday, light condition, weather, and work zone type. Work zone traffic accident data from the Fatality Analysis Report System between 2002 and 2009 were utilized to determine the distributions of the ANT and the EMS arrival time in the United States. A mixed logistic regression model, taking into account the uncertainty associated with the ANT and the EMS response time, was developed to estimate the risk of death. The results showed that the uncertainty of the ANT was primarily influenced by crash time and road type, whereas the uncertainty of EMS response time is greatly affected by road type, weather, and light conditions. In addition, work zone accidents occurring during a holiday and in poor light conditions were found to be statistically associated with a longer mean ANT and longer EMS response time. The results also show that shortening the ANT was a more effective approach in reducing the risk of death than the EMS response time in work zones. To shorten the ANT and the EMS response time, work zone activities are suggested to be undertaken during non-holidays, during the daytime, and in good weather and light conditions.

Design of a High Power Robotic Manipulator for Emergency Response to the Nuclear Accidents

International Nuclear Information System (INIS)

Park, Jongwon; Bae, Yeong-Geol; Kim, Myoung Ho; Choi, Young Soo

2016-01-01

An accident in a nuclear facility causes a great social cost. To prevent an unexpected nuclear accident from spreading to the catastrophic disaster, emergency response action in early stage is required. However, high radiation environment has been proved as a challenging obstacle for human workers to access to the accident site and take an action in previous accident cases. Therefore, emergency response robotic technology to be used in a nuclear accident site instead of human workers are actively conducted in domestically and internationally. Robots in an accident situation are required to carry out a variety of tasks depend on the types and patterns of accidents. An emergency response usually includes removing of debris, make an access road to a certain place and handling valves. These tasks normally involve high payload handling. A small sized high power robotic manipulator can be an appropriate candidate to deal with a wide spectrum of tasks in an emergency situation. In this paper, we discuss about the design of a high power robotic manipulator, which is capable of handling high payloads for an initial response action to the nuclear facility accident. In this paper, we presented a small sized high power robotic manipulator design. Actuator types of manipulator was selected and mechanical structure was discussed. In the future, the servo valve and hydraulic pump systems will be determined. Furthermore, control algorithms and test bed experiments will be also conducted

Design of a High Power Robotic Manipulator for Emergency Response to the Nuclear Accidents

Energy Technology Data Exchange (ETDEWEB)

Park, Jongwon; Bae, Yeong-Geol; Kim, Myoung Ho; Choi, Young Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

An accident in a nuclear facility causes a great social cost. To prevent an unexpected nuclear accident from spreading to the catastrophic disaster, emergency response action in early stage is required. However, high radiation environment has been proved as a challenging obstacle for human workers to access to the accident site and take an action in previous accident cases. Therefore, emergency response robotic technology to be used in a nuclear accident site instead of human workers are actively conducted in domestically and internationally. Robots in an accident situation are required to carry out a variety of tasks depend on the types and patterns of accidents. An emergency response usually includes removing of debris, make an access road to a certain place and handling valves. These tasks normally involve high payload handling. A small sized high power robotic manipulator can be an appropriate candidate to deal with a wide spectrum of tasks in an emergency situation. In this paper, we discuss about the design of a high power robotic manipulator, which is capable of handling high payloads for an initial response action to the nuclear facility accident. In this paper, we presented a small sized high power robotic manipulator design. Actuator types of manipulator was selected and mechanical structure was discussed. In the future, the servo valve and hydraulic pump systems will be determined. Furthermore, control algorithms and test bed experiments will be also conducted.

Severe accident sequences simulated at the Grand Gulf Nuclear Station

International Nuclear Information System (INIS)

Carbajo, J.J.

1999-01-01

Different severe accident sequences employing the MELCOR code, version 1.8.4 QK, have been simulated at the Grand Gulf Nuclear Station (Grand Gulf). The postulated severe accidents simulated are two low-pressure, short-term, station blackouts; two unmitigated small-break (SB) loss-of-coolant accidents (LOCAs) (SBLOCAs); and one unmitigated large LOCA (LLOCA). The purpose of this study was to calculate best-estimate timings of events and source terms for a wide range of severe accidents and to compare the plant response to these accidents

Atmospheric dispersion calculation for posturated accident of nuclear facilities and the computer code: PANDA

International Nuclear Information System (INIS)

Kitahara, Yoshihisa; Kishimoto, Yoichiro; Narita, Osamu; Shinohara, Kunihiko

1979-01-01

Several Calculation methods for relative concentration (X/Q) and relative cloud-gamma dose (D/Q) of the radioactive materials released from nuclear facilities by posturated accident are presented. The procedure has been formulated as a Computer program PANDA and the usage is explained. (author)

Analysis of emergency response procedures and air traffic accidents ...

African Journals Online (AJOL)

Incessant air transport accidents have been a source of concern to stakeholders and aviation experts in Nigeria, yet the response and process has not been adequately appraised. This study attempts an evaluation of the emergency response procedures in the aviation industry with particular focus on Murtala Muhammed ...

Contain calculations of debris conditions adjacent to the BWR Mark I drywell shell during the later phases of a severe accident

International Nuclear Information System (INIS)

Hyman, C.R.

1988-01-01

Best estimate CONTAIN calculations have recently been performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to predict the primary containment response during the later phases of an unmitigated low-pressure Short Term Station Blackout at the Peach Bottom Atomic Power Station. Debris pour conditions leaving the failed reactor vessel are taken from the results of best estimate BWRSAR analyses that are based upon an assumed metallic debris melting temperature of 2750/degree/F (1783 K) and an oxide debris melting temperature of 4350/degree/F (2672 K). Results of the CONTAIN analysis for the case without sprays indicate failure of the drywell seals due to the extremely hot atmospheric conditions extant in the drywell. The maximum calculated temperature of the debris adjacent to the drywell shell is less than the melting temperature of the shell, yet the sustained temperatures may be sufficient to induce primary containment pressure boundary failure by the mechanism of creep-rupture. It is also predicted that a significant portion of the reactor pedestal wall is ablated during the period of the calculation. Nevertheless, the calculated results are recognized to be influenced by large modeling uncertainties. Several deficiencies in the application of the CORCON module within the CONTAIN code to BWR severe accident sequences are identified and discussed. 5 refs., 9 figs., 4 tabs.,

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

International Nuclear Information System (INIS)

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

1990-12-01

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

Pilot program: NRC severe reactor accident incident response training manual: US Nuclear Regulatory Commission response

International Nuclear Information System (INIS)

Sakenas, C.A.; McKenna, T.J.; Perkins, K.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Giitter, J.G.; Watkins, R.M.

1987-02-01

This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. US Nuclear Regulatory Commission Response is the fifth in a series of volumes that collectively summarize the US Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume describes NRC response modes, organizations, and official positions; roles of other federal agencies are also described briefly. Each volume serves, respectively, as the text for a course of instruction in a series of courses for NRC response personnel. These materials do not provide guidance or license requirements for NRC licensees. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are called out in the text

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.

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.

Assessment of the accident response of a light-water-moderated breeder-reactor system: AWBA development program

International Nuclear Information System (INIS)

High, H.M.

1983-05-01

The predicted accident response for a light water moderated, thorium/U-233 fueled, seed-blanket reactor concept was assessed. The first part of the assessment compared breeder accident response with that of a current commercial pressurized water reactor design for several different types of transients. Based on these comparisons and a review of the various parameter differences between the breeder and a U-235 fueled plant, the second part of the assessment studied the breeder accident behavior in more detail, particularly in areas of potential concern. Based on the two parts of the assessment, it was concluded that the breeder accident response would be very similar to that of present commercial pressurized water reactor plants. The large Doppler and moderator reactivity coefficients of the breeder would significantly reduce the severity of many of the accidents that must be considered. It is expected that the accident response of the breeder can be shown to meet regulatory criteria

Emergency response to a highway accident in Springfield, Massachusetts, on December 16, 1991

International Nuclear Information System (INIS)

1992-06-01

On December 16, 1991, a truck carrying unirradiated (fresh) nuclear fuel was involved in an accident on US Interstate 91, in Springfield, Massachusetts. This report describes the emergency response measures undertaken by local, State, Federal, and private parties. The report also discusses ''lessons learned'' from the response to the accident and suggests areas where improvements might be made

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

International Nuclear Information System (INIS)

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

1988-12-01

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

Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix VI. Calculation of reactor accident consequences

International Nuclear Information System (INIS)

1975-10-01

Information is presented concerning the radioactive releases from the containment following accidents; radioactive inventory of the reactor core; atmospheric dispersion; reactor sites and meteorological data; radioactive decay and deposition from plumes; finite distance of plume travel; dosimetric models; health effects; demographic data; mitigation of radiation exposure; economic model; and calculated results with consequence model

An analysis of reactor structural response to fuel sodium interaction in a hypothetical core disruptive accident

International Nuclear Information System (INIS)

Suzuki, K.; Tashiro, M.; Sasanuma, K.; Nagashima, K.

1976-01-01

This study shows the effect of constraints around FSI zone on FSI phenomena and deformations of reactor structures. SUGAR-PISCES code system has been developed to evaluate the phenomena of FSI and the response of reactor structure. SUGAR calculates the phenomena of FSI. PISCES, developed by Physics International Company in U.S.A., calculates the dynamic response of reactor structure in two-dimensional, time-dependent finite-difference Lagrangian model. The results show that the peak pressure and energy by FSI and the deformation of reactor structures are about twice in case of FSI zone surrounding by blanket than by coolant. The FSI phenomena highly depend on the reactor structure and the realistic configuration around core must be considered for analyzing hypothetical core disruptive accident. This work was supported by a grant from Power Reactor and Nuclear Fuel Development Corporation. (auth.)

An Ontology-Underpinned Emergency Response System for Water Pollution Accidents

Directory of Open Access Journals (Sweden)

Xiaoliang Meng

2018-02-01

Full Text Available With the unceasing development and maturation of environment geographic information system, the response to water pollution accidents has been digitalized through the combination of monitoring sensors, management servers, and application software. However, most of these systems only achieve the basic and general geospatial data management and functional process tasks by adopting mechanistic water-quality models. To satisfy the sustainable monitoring and real-time emergency response application demand of the government and public users, it is a hotspot to study how to make the water pollution information being semantic and make the referred applications intelligent. Thus, the architecture of the ontology-underpinned emergency response system for water pollution accidents is proposed in this paper. This paper also makes a case study for usability testing of the water ontology models, and emergency response rules through an online water pollution emergency response system. The system contributes scientifically to the safety and sustainability of drinking water by providing emergency response and decision-making to the government and public in a timely manner.

Graphic console for analysis of severe accidents visualization of OEs, NAEs and calculation of the source term for the NPP-LV (CoGrAAS)

International Nuclear Information System (INIS)

Sandoval V, S.; Mendoza E, P. R.; Gonzalez C, J. M.; Cecenas F, M.; Tijerina S, F.

2016-09-01

In response to the Fukushima Daiichi nuclear power plant accident, the NRC conducted an analysis and issued recommendations to improve the safety of the nuclear reactors. These include strengthening and integrating emergency response capabilities and emphasizing periodic staff training, the performance of simulation exercises. As a tool to observe these recommendations, the Graphic Console was developed for Analysis of Severe Accidents, Visualization of OEs, NAEs and calculation of the source term for nuclear power plant of Laguna Verde (NPP-LV ; CoGrAAS). The CoGrAAS is a computer system that displays in an integrated, graphic and dynamic way the information of a catalog of previously simulated accident scenarios. Has core mimics, vessel, primary containment and safety systems, trend graph of thermodynamic and radiological variables and the emergency procedures (OEs), chronological list of events, windows with detailed information for the dry-well, among others. The use of CoGrAAS allows that staff to understand and become familiar with the thermo-hydraulic progression of actual scenarios that exceed the design basis including those with core damage as severe accidents. The system enables personnel to develop an integral vision of the scenarios during the exercises and drills by observing and analyzing the evolution of the main reactor, core and primary containment variables, the response of emergency systems and the influence of that progression on OEs and the emergency action levels (NAEs). The CoGrAAS allows o observe the radiological variables and obtain the source term, to make the projection of doses, at any time within the scenario evolution. Thus, not only can the phenomenology of severe accidents be analyzed and understood, it is also possible to exercise, verify and evaluate the performance of critical tasks in the application of procedures, guidelines and emergency management plans. (Author)

Thermal-hydraulic uncertainties affecting severe accident progression

International Nuclear Information System (INIS)

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

1984-01-01

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

Review and assessment of package requirements (yellowcake) and emergency response to transportation accidents

International Nuclear Information System (INIS)

1978-10-01

As a consequence of an accident involving a truck shipment of yellowcake, a joint NRC--DOT study was undertaken to review and assess the regulations and practices related to package integrity and to emergency response to transportation accidents involving low specific activity radioactive materials. Recommendations are made regarding the responsibilities of state and local agencies, carriers, and shippers, and the DOT and NRC regulations

Calculation and experimental study of the RBMK-1500 reactor emergency cooling at maximum designed accident

International Nuclear Information System (INIS)

Cherkashov, Yu.M.; Vasilevskij, V.P.; Labazov, V.H.; Loninov, A.Ya.; Molochnikov, Yu.S.; Novosel'skij, O.Yu.; Podlazov, L.N.; Pavlov, V.B.; Pushkarev, V.I.

1981-01-01

The analysis of thermohydraulic and neutron-physical processes occurring in the RBMK-1500 reactor during the reactor emergency cooling system triggering (RECS) after the maximum designed accident (MDA) is conducted. The MDA means hypothetical instant hilliotine break of the main circulating pump head collector. During the whole cooling down period the RECS should provide the temperature level of the fuel elements not exceeding 1200 deg C and the channel pipe temperature - 600 deg C. The principal flowsheet of the balloon type RECS is described. Calculations of the valve fast response effect on the RECS productivity are carried out. It is concluded that the chosen balloon RECS provides reliable temperature modes of fuel elements naand channel pipes under the MDA conditions. At the same time a momentary splash of neutron power by the value not more than 10% can take place [ru

Nuclear accident/radiological emergency assistance plan. NAREAP - edition 2000. Emergency preparedness and response

International Nuclear Information System (INIS)

2000-01-01

The purpose of the Nuclear Accident/Radiological Emergency Assistance Plan (NAREAP) is to describe the framework for systematic, integrated, co-ordinated, and effective preparedness and response for a nuclear accident or radiological emergency involving facilities or practices that may give rise to a threat to health, the environment or property. The purpose of the NAREAP is: to define the emergency response objectives of the Agency's staff in a nuclear accident or a radiological emergency; to assign responsibilities for performing the tasks and authorities for making the decisions that comprise the Agency staff's response to a nuclear accident or radiological emergency; to guide the Agency managers who must ensure that all necessary tasks are given the necessary support in discharging the Agency staff responsibilities and fulfilling its obligations in response to an emergency; to ensure that the development and maintenance of detailed and coherent response procedures are well founded; to act as a point of reference for individual Agency staff members on their responsibilities (as an individual or a team member) throughout a response; to identify interrelationships with other international intergovernmental Organizations; and to serve as a training aid to maintain readiness of personnel. The NAREAP refers to the arrangements of the International Atomic Energy Agency and of the United Nations Security and Safety Section at the Vienna International Centre (UNSSS-VIC) that may be necessary for the IAEA to respond to a nuclear accident or radiological emergency, as defined in the Early Notification and Assistance Conventions. It covers response arrangements for any situation that may have actual, potential or perceived radiological consequences and that could require a response from the IAEA, as well as the arrangements for developing, maintaining and exercising preparedness. The implementing procedures themselves are not included in the NAREAP, but they are required

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

Containment response to a severe accident (TMLB sequence) with and without mitigation strategies

International Nuclear Information System (INIS)

Passalacqua, R.

2004-01-01

A loss of SG feed-water (TMLB sequence) for a prototypic PWR 900 MWe with a multi-compartment configuration (with 11 and 16 cells nodalization) has been calculated by the author using the ASTEC code in the frame of the EVITA project (5th Framework Programme, FWP). A variety of hypothesis (e.g. activation of sprays and hydrogen recombiners) and possible consequences of these assumptions (cavity flooding, hydrogen combustion, etc.) have been made in order to evaluate the global reactor containment building response (pressure, aerosol/FP concentration, etc.). The need to dispose of severe accident management guidelines (SAMGs) is increasing. These guidelines are meant for nuclear plants' operators in order to allow them to apply mitigation strategies all along a severe accident, which, only in its initial phase, may last several days. The purpose of this paper is to outline the influence on the containment load of most common accident occurrences and operators actions, which is essential in establishing SAMGs. ASTEC (Accident Source Term Evaluation Code) is a computer code for the evaluation of the consequences of a postulated nuclear plant severe accident sequence. ASTEC is a computer tool currently under joint development by the Institut de Radioprotection et de Surete Nucleaire (IRSN), France, and Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS), Germany. The aim of the development is to create a fast running integral code package, reliable in all simulations of a severe accident, to be used for level-2 PSA analysis. It must be said that several recent developments have significantly improved the best-estimate models of ASTEC and a new version (ASTEC V1.0) has been released mid-2002. Nevertheless, the somehow obsolete ASTECv0.3 version here used, has given results very useful for the estimation of the global risk of a nuclear plant. Moreover, under the current 6th FWP (Sustainable Integration of EU Research on Severe Accident Phenomenology and Management), the

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

ARGX-87: Accident Response Group Exercise, 1987: A Broken Arrow mini exercise

International Nuclear Information System (INIS)

Schuld, E.P.; Cruff, D.F.

1987-07-01

A Broken Arrow mini exercise dubbed ''Accident Response Group Exercise - 1987'' (ARGX-87) was conducted on June 1, 1987 at the Lawrence Livermore National Laboratory (LLNL) and Sandia National Laboratories, Livermore (SNLL). The exercise started at 0445 PDT with a call from the Department of Energy (DOE) - EOC in Washington, DC, to the Albuquerque Operations (AL - ) - EOC. AL, in turn, called the Laboratory off-hour emergency number (Fire Dispatcher), who called the Laboratory Emergency Duty Officer (LEDO). The LEDO then contacted the Accident Response Group (ARG) Senior Scientific Advisor. Calls were placed to assemble appropriate members of the ARG in the ALERT Center. No phone number for SNLL was available at the Albuquerque Operations EOC, so a controller injected a message to SNLL to get them involved in the exercise. The messages received at the Laboratory identified the Air Force line item weapon system involved in the accident and the accident location. As people arrived at the ALERT Center they began discussing the details of the accident. They also started working the deployment logistics and other issues. Travel arrangements for the HOT SPOT equipment and ARG personnel were made for immediate deployment to the accident site in North Dakota. The exercise was terminated at 0840 as planned. While certain procedural deficiencies were noted, the exercise was considered a valuable learning experience. The results and observations from this experience will be used to refine the operating procedures and the training program

A model national emergency response plan for radiological accidents

International Nuclear Information System (INIS)

1993-09-01

The IAEA has supported several projects for the development of a national response plan for radiological emergencies. As a results, the IAEA has developed a model National Emergency Response Plan for Radiological Accidents (RAD PLAN), particularly for countries that have no nuclear power plants. This plan can be adapted for use by countries interested in developing their own national radiological emergency response plan, and the IAEA will supply the latest version of the RAD PLAN on computer diskette upon request. 2 tabs

Benchmarking MARS (accident management software) with the Browns Ferry fire

International Nuclear Information System (INIS)

Dawson, S.M.; Liu, L.Y.; Raines, J.C.

1992-01-01

The MAAP Accident Response System (MARS) is a userfriendly computer software developed to provide management and engineering staff with the most needed insights, during actual or simulated accidents, of the current and future conditions of the plant based on current plant data and its trends. To demonstrate the reliability of the MARS code in simulatng a plant transient, MARS is being benchmarked with the available reactor pressure vessel (RPV) pressure and level data from the Browns Ferry fire. The MRS software uses the Modular Accident Analysis Program (MAAP) code as its basis to calculate plant response under accident conditions. MARS uses a limited set of plant data to initialize and track the accidnt progression. To perform this benchmark, a simulated set of plant data was constructed based on actual report data containing the information necessary to initialize MARS and keep track of plant system status throughout the accident progression. The initial Browns Ferry fire data were produced by performing a MAAP run to simulate the accident. The remaining accident simulation used actual plant data

Relocation tabletop exercise: federal radiological response in the post-accident phase

International Nuclear Information System (INIS)

Grant, K.; Adler, M.V.; Wolff, W.F.

1986-01-01

The federal Radiological Emergency Response Plan (FRERP) was developed to provide the framework for coordinating federal radiological assistance to states and to local authorities faced with a large radiological accident. The Relocation Tabletop Exercise was conducted on December 9-11, 1985 at the Beaver Valley Power Station, the site of the simulated accident. The exercise scenario had postulated a substantial release of radioactive materials from a fuel handling accident at the Beaver Valley Power Station in Shippingport, Pennsylvania, leaving radioactive materisls deposited over part of the surrounding area. The exercise was structured as a sequential series of nice mini-scenarios, each of which focused on one problems. The exercise was intended to identify issues and problems which needed consideration or procedures which might need to be developed for this post-accident phase. It was a ''no-fault'' excercise

Emergency response strategies

International Nuclear Information System (INIS)

Carrilo, D.; Dias de la Cruz, F.

1984-01-01

In the present study is estimated, on the basis of a release category (PWR4) and several accident scenarios previously set up, the emergency response efficacy obtained in the application of different response strategies on each of the above mentioned scenarios. The studied strategies contemplate the following protective measures: evacuation, shelter and relocation. The radiological response has been obtained by means of CRAC2 (Calculation of Reactor Accident Consequences) code, and calculated in terms of absorbed dose equivalent (Whole body and thyroid), as well as early and latent biological effects. (author)

Improving Research Reactor Accident Response Capability at the Hungarian Nuclear Safety Authority

International Nuclear Information System (INIS)

Vegh, J.; Gajdos, F.; Horvath, Cs.; Matisz, A.; Nyisztor, D.

2013-06-01

The paper describes the design and implementation of an on-line operation monitoring and accident response support system to be used at the CERTA emergency response centre of Hungarian Atomic Energy Authority (HAEA). The monitored facility is the Budapest Research Reactor (BRR), which is a tank-type thermal reactor having 10 MW thermal power. The basic reason for the development of the on-line safety information system is to extend the emergency response capability of the CERTA crisis centre to include emergencies related to BRR, as well. CERTA is operated by HAEA at its Budapest headquarters and the centre already has an on-line system for monitoring the state of the four units of Paks NPP, Hungary. The system is called CERTA VITA and it is able to monitor the four VVER-440/V213 units during their normal operation, and during emergencies (including severe accidents). Ensuring appropriate emergency response capabilities, as well as improving the presently available systems and tools was one of the important recommendations resulting from the analyses following the severe accident at Fukushima. This task is valid not only for the operators of the nuclear facilities but also for the nuclear safety authorities, therefore HAEA decided to launch a project - together with the Centre for Energy Research, the operator of BRR - to establish an on-line information system similar to the CERTA VITA used for monitoring the four units of the Paks NPP. It is believed that by the introduction of this new on-line system the accident response capabilities of HAEA will be further enhanced and the BRR emergencies will be handled at the same professional level as potential emergencies at Paks NPP. (authors)

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...

Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

International Nuclear Information System (INIS)

Sdouz, Gert

2006-01-01

The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the un-tightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the 'Station Blackout'- sequence and the 'Large Break LOCA'. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was

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

Severe accident analysis methodology in support of accident management

International Nuclear Information System (INIS)

Boesmans, B.; Auglaire, M.; Snoeck, J.

1997-01-01

The author addresses the implementation at BELGATOM of a generic severe accident analysis methodology, which is intended to support strategic decisions and to provide quantitative information in support of severe accident management. The analysis methodology is based on a combination of severe accident code calculations, generic phenomenological information (experimental evidence from various test facilities regarding issues beyond present code capabilities) and detailed plant-specific technical information

Radionuclide metrology: traceability and response to a radiological accident

Energy Technology Data Exchange (ETDEWEB)

Tauhata, L.; Cruz, P.A.L. da; Silva, C.J. da; Delgado, J.U.; Oliveira, A.E. de; Oliveira, E.M. de; Poledna, R.; Loureiro, J. dos S.; Ferreira Filho, A.L.; Silva, R.L. da; Filho, O. L.T.; Santos, A.R.L. dos; Veras, E.V. de; Rangel, J. de A.; Quadros, A.L.L.; Araújo, M.T.F. de; Souza, P.S. de; Ruzzarim, A.; Conceição, D.A. da; Iwahara, A., E-mail: palcruz@ird.gov.br [Instituto de Radioproteção e Dosimetria (LNMRI/IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. Nacional de Metrologia das Radiações Ionizantes

2017-07-01

In the case of a radiological accident, there are characteristic phases: discovery and initial assistance with first aid; the triage and monitoring of the affected population; the release of the affected people; forward the victims to medical care; as well as the preparation of the report on the accident. In addition, studies and associated researches performed in the later period. Monitors, dosimeters and measuring systems should be calibrated by contaminating radionuclide standards. The radioactive sources used must be metrologically reliable. In Brazil, this function is performed by LNMRI/IRD/CNEN, designated by INMETRO, which Radionuclide Metrology Laboratory is responsible for the standardization and supply of radioactive sources in diverse geometries and matrices. This laboratory has a stock of radionuclide solutions with controlled environmental variables for the preparation of sources, which are calibrated and standardized by mean of primary and secondary systems. It is also responsible for the dissemination of standards and, in order to establish the metrological traceability of national standards, participates in international key-comparisons promoted by BIPM and regional metrology organizations. Internally, it promotes the National Comparison Programs for laboratories for the analysis of environmental samples and the traceability for producing centers of radiopharmaceuticals and Nuclear Medicine Services in the country. The paper presents the demand for {sup 137}Cs related to the radioactive accident in Goiania/Brazil and the significant results for the main radionuclides standardized by the Radionuclide Metrology Laboratory for international key-comparisons and national comparisons to provide metrological traceability. With the obtained results, the LNMRI of Brazil integrates the international metrology BIPM network and fulfills its function of supplying, with about a hundred of radioactive standards, the country's needs in different applications

Radionuclide metrology: traceability and response to a radiological accident

International Nuclear Information System (INIS)

Tauhata, L.; Cruz, P.A.L. da; Silva, C.J. da; Delgado, J.U.; Oliveira, A.E. de; Oliveira, E.M. de; Poledna, R.; Loureiro, J. dos S.; Ferreira Filho, A.L.; Silva, R.L. da; Filho, O. L.T.; Santos, A.R.L. dos; Veras, E.V. de; Rangel, J. de A.; Quadros, A.L.L.; Araújo, M.T.F. de; Souza, P.S. de; Ruzzarim, A.; Conceição, D.A. da; Iwahara, A.

2017-01-01

In the case of a radiological accident, there are characteristic phases: discovery and initial assistance with first aid; the triage and monitoring of the affected population; the release of the affected people; forward the victims to medical care; as well as the preparation of the report on the accident. In addition, studies and associated researches performed in the later period. Monitors, dosimeters and measuring systems should be calibrated by contaminating radionuclide standards. The radioactive sources used must be metrologically reliable. In Brazil, this function is performed by LNMRI/IRD/CNEN, designated by INMETRO, which Radionuclide Metrology Laboratory is responsible for the standardization and supply of radioactive sources in diverse geometries and matrices. This laboratory has a stock of radionuclide solutions with controlled environmental variables for the preparation of sources, which are calibrated and standardized by mean of primary and secondary systems. It is also responsible for the dissemination of standards and, in order to establish the metrological traceability of national standards, participates in international key-comparisons promoted by BIPM and regional metrology organizations. Internally, it promotes the National Comparison Programs for laboratories for the analysis of environmental samples and the traceability for producing centers of radiopharmaceuticals and Nuclear Medicine Services in the country. The paper presents the demand for 137 Cs related to the radioactive accident in Goiania/Brazil and the significant results for the main radionuclides standardized by the Radionuclide Metrology Laboratory for international key-comparisons and national comparisons to provide metrological traceability. With the obtained results, the LNMRI of Brazil integrates the international metrology BIPM network and fulfills its function of supplying, with about a hundred of radioactive standards, the country's needs in different applications

Investigation of air cleaning system response to accident conditions

International Nuclear Information System (INIS)

Andrae, R.W.; Bolstad, J.W.; Foster, R.D.; Gregory, W.S.; Horak, H.L.; Idar, E.S.; Martin, R.A.; Ricketts, C.I.; Smith, P.R.; Tang, P.K.

1980-01-01

Air cleaning system response to the stress of accident conditions are being investigated. A program overview and hghlight recent results of our investigation are presented. The program includes both analytical and experimental investigations. Computer codes for predicting effects of tornados, explosions, fires, and material transport are described. The test facilities used to obtain supportive experimental data to define structural integrity and confinement effectiveness of ventilation system components are described. Examples of experimental results for code verification, blower response to tornado transients, and filter response to tornado and explosion transients are reported

Analysis of emergency response to fukushima nuclear accident in Japan and suggestions for China's nuclear emergency management

International Nuclear Information System (INIS)

Li Wei; Ding Qihua; Wu Haosong

2014-01-01

On March 11, 2011, the Fukushima Dai-ichi Nuclear Power Station of the Tokyo Electric Power Company ('TEPCO') was hit and damaged by a magnitude 9 earthquake and accompanying tsunami. The accident is determined to be of the highest rating on the International Nuclear Event Scale. The Government of Japan and TEPCO have taken emergency response actions on-site and off-site at the accident. It became clear through the investigation that the accident had been initiated on the occasion of a natural disaster of an earthquake and tsunami, but there have been various complex problems behind this very serious and large scale accident. For an example, the then-available accident preventive measures and disaster preparedness of TEPCO were insufficient against tsunami and severe accidents; inadequate TEPCO emergency responses to the accident at the site were also identified. The accident rang the alarm for the nuclear safety of nuclear power plants. It also taught us a great of lessons in nuclear emergency management. (authors)

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

International Nuclear Information System (INIS)

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

1985-01-01

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

The Radiological Accident in Lia, Georgia

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-12-15

The use of radioactive material offers a wide range of benefits to medicine, research and industry throughout the world. Precautions are necessary, however, to limit the exposure of people to the radiation emitted. Where the amount of radioactive material is substantial, as in the case of radiotherapy or industrial radiography sources, great care is required to prevent accidents which could have severe consequences. Nevertheless, in spite of the precautions taken, serious accidents involving radiation sources continue to occur, albeit infrequently. The IAEA conducts follow-up reviews of such serious accidents to provide an account of their circumstances and consequences, from which organizations with responsibilities for radiation protection, safety of sources and emergency preparedness and response may learn. A serious radiological accident occurred in Georgia on 2 December 2001, when three inhabitants of the village of Lia found two metal objects in the forest while collecting firewood. These objects were {sup 90}Sr sources with an activity of 1295 TBq. The three inhabitants used the objects as heaters when spending the night in the forest. The major cause of the accident was the improper and unauthorized abandonment of radiation sources in Georgia and the absence of clear labels or radiation signs on the sources warning of the potential radiation hazard. Under the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency (Assistance Convention), the Georgian authorities requested assistance from the IAEA to advise on the dose assessment, source recovery and medical management of those involved in the accident. This publication describes the circumstances and events surrounding the accident, its management and the medical treatment of the people exposed. It also describes the dose reconstruction calculations and biodosimetry assessments conducted. A number of uncertainties remain relating to some details of the accident. However

The Radiological Accident in Lia, Georgia

International Nuclear Information System (INIS)

2014-12-01

The use of radioactive material offers a wide range of benefits to medicine, research and industry throughout the world. Precautions are necessary, however, to limit the exposure of people to the radiation emitted. Where the amount of radioactive material is substantial, as in the case of radiotherapy or industrial radiography sources, great care is required to prevent accidents which could have severe consequences. Nevertheless, in spite of the precautions taken, serious accidents involving radiation sources continue to occur, albeit infrequently. The IAEA conducts follow-up reviews of such serious accidents to provide an account of their circumstances and consequences, from which organizations with responsibilities for radiation protection, safety of sources and emergency preparedness and response may learn. A serious radiological accident occurred in Georgia on 2 December 2001, when three inhabitants of the village of Lia found two metal objects in the forest while collecting firewood. These objects were 90 Sr sources with an activity of 1295 TBq. The three inhabitants used the objects as heaters when spending the night in the forest. The major cause of the accident was the improper and unauthorized abandonment of radiation sources in Georgia and the absence of clear labels or radiation signs on the sources warning of the potential radiation hazard. Under the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency (Assistance Convention), the Georgian authorities requested assistance from the IAEA to advise on the dose assessment, source recovery and medical management of those involved in the accident. This publication describes the circumstances and events surrounding the accident, its management and the medical treatment of the people exposed. It also describes the dose reconstruction calculations and biodosimetry assessments conducted. A number of uncertainties remain relating to some details of the accident. However, sufficient

Use of inelastic analysis to determine the response of packages to puncture accidents

International Nuclear Information System (INIS)

Ammerman, D.J.; Ludwigsen, J.S.

1996-01-01

The accurate analytical determination of the response of radioactive material transportation packages to the hypothetical puncture accident requires inelastic analysis techniques. Use of this improved analysis method recudes the reliance on empirical and approximate methods to determine the safety for puncture accidents. This paper will discuss how inelastic analysis techniques can be used to determine the stresses, strains and deformations resulting from puncture accidents for thin skin materials with different backing materials. A method will be discussed to assure safety for all of these types of packages

Conservatism of loss-of-coolant accident licensing analysis compared to experimental results and best-estimate calculation

International Nuclear Information System (INIS)

Winkler, F.; Friedmann, P.

1986-01-01

The paper compares results of loss-of-coolant accident licensing analysis with experimental results and results of best-estimate calculations. The large safety margins resulting from the more realistic best-estimate results are used to show the high conservatism inherent in the licensing process of pressurized water reactors. (orig.) [de

A radioactive waste transportation package monitoring system for normal transport and accident emergency response conditions

International Nuclear Information System (INIS)

Brown, G.S.; Cashwell, J.W.; Apple, M.L.

1993-01-01

This paper addresses spent fuel and high level waste transportation history and prospects, discusses accident histories of radioactive material transport, discusses emergency responder needs and provides a general description of the Transportation Intelligent Monitoring System (TRANSIMS) design. The key objectives of the monitoring system are twofold: (1) to facilitate effective emergency response to accidents involving a radioactive waste transportation package, while minimizing risk to the public and emergency first-response personnel, and (2) to allow remote monitoring of transportation vehicle and payload conditions to enable research into radioactive material transportation for normal and accident conditions. (J.P.N.)

Photon energy dependence and angular response of glass display used in mobile phones for accident dosimetry

International Nuclear Information System (INIS)

Discher, Michael; Greiter, Matthias; Woda, Clemens

2014-01-01

Previous studies have shown that glass displays extracted from mobile phones are suitable as emergency dosimeters in case of an accidental radiation overexposure using the thermoluminescence (TL) method. So far these studies have focused only on recovering the absorbed dose to the material. However, dose in air or dose to the victim carrying the device might be significantly different. Therefore the aim of this work was to investigate photon energy dependence and angular response of glass display used in modern mobile phones. An over-response of about a factor of five is observed for low photon energies compared to the response to Cs-137 (662 keV) which is in reasonable agreement with calculated values mass energy-absorption coefficients of glass and air. Little variation in the energy dependence can be seen for glass displays coming from three different mobile phone models. The angular response for display glass is flat with regard to air kerma within the incident angle of ±60°, independent of the irradiation setup used (with a water phantom or with air kerma reference conditions). For incident angles of 90° the shielding effect of the mobile phones becomes important. With the dosimetric characterization of the photon energy and angular dependencies the absorbed dose in a glass display can be transferred to a reference air kerma dose and provides a useful option for retrospective accident dosimetry. - Highlights: • Determination of the photon energy dependence and angular response for display glass used as an accident dosimeter. • Over-response of about a factor of five for low photon energies. • Flat angular response within incident angles between ±60°

Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

International Nuclear Information System (INIS)

Hsii, Y.H.

1975-01-01

The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The analytical model used for the program is described. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc. 11 references. (U.S.)

The role of chemistry in nuclear accidents

International Nuclear Information System (INIS)

Johnson, C.E.; Johnson, I.

1986-01-01

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

Accidents in radiotherapy: Lack of quality assurance?

International Nuclear Information System (INIS)

Novotny, J.

1997-01-01

About 150 radiological accidents, involving more than 3000 patients with adverse effects, 15 patient's fatalities and about 5000 staff and public exposures have been collected and analysed. Out of 67 analysed accidents in external beam therapy 22% has been caused by wrong calculation of the exposure time or monitor units, 13% by inadequate review of patient's chart, 12% by mistakes in the anatomical area to be treated. The remaining 35% can be attributed to 17 different causes. The most common mistakes in brachytherapy were wrong activities of sources used for treatment (20%), inadequate procedures for placement of sources applicators (14%), mistakes in calculating the treatment time (12%), etc. The direct and contributing causes of radiological accidents have been deduced from each event, when it was possible and categorized into 9 categories: mistakes in procedures (30%), professional mistakes (17%), communication mistakes (15%), lack of training (8.5%), interpretation mistakes (7%), lack of supervision (6%), mistakes in judgement (6%), hardware failures (5%), software and other mistakes (5.5%). Three types of direct and contributing causes responsible for almost 62% of all accidents are directly connected to the quality assurance of treatment. The lessons learnt from the accidents are related to frequencies of direct and contributing factors and show that most of the accident are caused by lack, non-application of quality assurance (QA) procedures or by underestimating of QA procedures. The international system for collection of accidents and dissemination of lessons learnt from the different accidents, proposed by IAEA, can contribute to better practice in many radiotherapy departments. Most of the accidents could have been avoided, had a comprehensive QA programme been established and properly applied in all radiotherapy departments, whatever the size. (author)

Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix VI. Calculation of reactor accident consequences. [PWR and BWR

Energy Technology Data Exchange (ETDEWEB)

1975-10-01

Information is presented concerning the radioactive releases from the containment following accidents; radioactive inventory of the reactor core; atmospheric dispersion; reactor sites and meteorological data; radioactive decay and deposition from plumes; finite distance of plume travel; dosimetric models; health effects; demographic data; mitigation of radiation exposure; economic model; and calculated results with consequence model.

Organizational culture during the accident response process

International Nuclear Information System (INIS)

Shurberg, D.A.; Haber, S.B.

1992-01-01

A large volume of literature hypothesizes a direct relationship between organizational culture and organizational effectiveness. Culture data have been collected by the authors and others at nuclear power plants (NPPs) and other organizations that demand high reliability. In this paper, the literature and data are used to explore a critical dimension of the accident response process in an NPP: the transition from an anticipatory strategy to an ad hoc strategy. In particular, the effect of organizational culture on the implementation of each of these strategies is examined

Prevalence of bronchopulmonary pathology in the participants of Chernobyl Nuclear Power Plant accident response

International Nuclear Information System (INIS)

Antonov, N.S.; Yakushin, S.P.; Stulova, O.Yu.; Zajtseva, O.Yu.; Stroev, E.A.

1998-01-01

Epidemiologic examination of the participants of the Chernobyl accident response is performed. Fact of acute effect of the Chernobyl aerosol inhalation on respiratory organs is found. Prevalence of bronchopulmonary diseases in participants of accident response is almost 2 times higher than that in reference group. Further program of investigations includes the hospital stage and the preventive measures at prehospital stage under ambulatory conditions. Assessments of the efficiency of performed treatment - prophylactic measures and their economic benefit are made [ru

Inventory of programs. Calculation of the isotope inventory after a hypothetical accident at the Cofrentes Nuclear power

International Nuclear Information System (INIS)

Albendea, M.

2014-01-01

Iberdrola is developing a new application to calculate the inventory of radiological material, then of a hypothetical accident, with the name of inventory. This application allows you to calculate the inventory isotopic, analysers and accurate thermal of all or part of the nucleus of the plant of Cofrentes, even of any single element, based on its history of irradiation and specific periods of decay, since the reactor at any time after the shutdown. (Author)

The response to a worst-case scenario - the national emergency plan for nuclear accidents

Energy Technology Data Exchange (ETDEWEB)

Cunningham D, John [Radiological Protection Inst. of Ireland (Ireland)

1996-10-01

The Chernobyl accident in 1986 highlighted many deficiencies in the preparedness of countries to deal with a major accident. It demonstrated how vulnerable countries are to transboundary contamination. Ireland had no emergency plan at the time of the accident and only minimal facilities with which to assess the consequences of the accident. Nonetheless, the then Nuclear Energy Board with the assistance of Government Departments and the Civil Defence organisation reacted quickly to assess the situation despite the complete lack of information about the accident from the then USSR. Even countries with advanced nuclear technologies faced similar difficulties. It was quickly recognised by Government that the national laboratory facilities were totally inadequate. The Nuclear Energy Board was provided with additional resources to assist it to cope in the short term with the very large demand for monitoring. In the longer term a new national radiation laboratory was provided and the Board was formally replaced by the Radiological Protection Institute of Ireland. It was given statutory responsibility to monitor radiation levels, to advise measures to be taken for the protection of the public and to provide information for the public. An emergency plan based on the Chernobyl experience was drafted in 1987, amended and published in 1992. Certain features of this plan were implemented from 1987 onwards, notably the classification of responsibilities and the installation of a national continuous radiation monitoring system. The paper outlines the responsibilities of those who could be involved in a response to a nuclear incident, the procedures used to evaluate its consequences and the provision of information for the public. The plan provides an integrated management system which has sufficient flexibility to enable a rapid response to be made to a major or minor crisis, either foreseen or unforeseen and whatever its cause.

Legal responsibility in case of a nuclear accident

International Nuclear Information System (INIS)

Nabhane, M. F.

1988-01-01

Numerous laws have been elaborated in order to determine the legal responsibility in case of a nuclear accident. These laws were made necessary because of intervention of the factor 'error' in the nuclear accident. The legal definition of 'error' assumes that it results from non-respect or negligence of established norms on the part of the persons who manipulate the instruments of radioactive production. Nuclear research should not be undertaken in a country without the formal engagement of the central authorities to take the necessary dispositions to ensure the security and safety of the populations and their possessions. The world community should not admit a scientific activity in the nuclear field in the absence of guarantees for the safety and the security of man. The state that permits the production of nuclear energy is legally responsible for any failure that might result in radioactive spills. Considering the possibility of error and the dangers attached to the manipulation of radioactive material, the legislators have elaborated a series of laws, which take into consideration two principles: a)The inalienable right of man to life as conceived in the monotheistic religions and proclaimed by positive law; and b)The responsibility of the state for the safety and security of its citizens. Of course, error is human; but if man may make an error of judgement in ordinary normal life, he does not have the right to make the least miscalculation when this might lead to a nuclear disaster. (author)

Application of RASCAL code for multiunit accident in domestic nuclear sites

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Monte Carlo dose reconstruction in case of a radiological accident: application to the accident in Chile in December 2005; Reconstitution de dose par calcul Monte Carlo en cas d'accident radiologique: application a l'accident du Chili de decembre 2005

Energy Technology Data Exchange (ETDEWEB)

Huet, C.; Clairand, I.; Trompier, F.; Bottollier-Depois, J.F. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Dir. de la Radioprotection de l' Homme, 92 - Fontenay aux Roses (France); Bey, E. [Hopital d' Instruction des Armees Percy, 92 - Clamart (France)

2007-10-15

Following a radiological accident caused by a gamma-graphy source in Chile in December 2005 involving one victim, I.R.S.N. was contacted to perform the dosimetric reconstruction of the accident using numerical simulation. Tools developed in the laboratory, associating anthropomorphic mathematic or voxel phantoms with the Monte Carlo calculation code m.c.n.p.x., were used in order to determine the dose distribution on the left buttock and absorbed doses to critical organs. The dosimetric mapping show that the absorbed at the skin surface is very high (1900 Gy) but drops rapidly at deep. At a depth of 5 cm, it is 20 Gy. Calculations performed with a mathematical phantom indicate that average doses to the critical organs are relatively low. Moreover, possible bone marrow sites for puncture are identified. Based on the dosimetric mapping, an excision measuring 5 cm in depth by 10 cm in diameter was performed on the left buttock of the victim. (authors)

A simplified model for calculating atmospheric radionuclide transport and early health effects from nuclear reactor accidents

International Nuclear Information System (INIS)

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

1995-01-01

During certain hypothetical severe accidents in a nuclear power plant, radionuclides could be released to the environment as a plume. Prediction of the atmospheric dispersion and transport of these radionuclides is important for assessment of the risk to the public from such accidents. A simplified PC-based model was developed that predicts time-integrated air concentration of each radionuclide at any location from release as a function of time integrated source strength using the Gaussian plume model. The solution procedure involves direct analytic integration of air concentration equations over time and position, using simplified meteorology. The formulation allows for dry and wet deposition, radioactive decay and daughter buildup, reactor building wake effects, the inversion lid effect, plume rise due to buoyancy or momentum, release duration, and grass height. Based on air and ground concentrations of the radionuclides, the early dose to an individual is calculated via cloudshine, groundshine, and inhalation. The model also calculates early health effects based on the doses. This paper presents aspects of the model that would be of interest to the prediction of environmental flows and their public consequences

Response to a Chemical Incident or Accident -- Who Is In Charge?

National Research Council Canada - National Science Library

Briggs, Darryl J

2007-01-01

.... The thesis of the paper is as follows: Combatant Commanders and the Services must have specific guidance and appropriate authorities to be able to effectively manage a Chemical Accident and Incident Response and Assistance (CAIRA...

Return to driving after severe traumatic brain injury: increased risk of traffic accidents and personal responsibility.

Science.gov (United States)

Bivona, Umberto; DʼIppolito, Mariagrazia; Giustini, Marco; Vignally, Pascal; Longo, Eloise; Taggi, Franco; Formisano, Rita

2012-01-01

To determine the frequency of road traffic accidents among individuals who start or resume driving after severe traumatic brain injury (TBI) and to investigate their responsibility for these accidents. Observational/retrospective study. Sixty adults with severe TBI and their caregivers. Return to Driving Questionnaire and Glasgow Outcome Scale. Thirty of the 60 participants started to drive or resumed driving after TBI. Nineteen (63%) of them were involved in traffic accidents, with personal responsibility in 26 of 36 after return to driving. Participants caused a significantly higher number of accidents after TBI than before. The ability to drive is frequently compromised after severe TBI. Specific rehabilitation of this complex activity should be a main goal of social reintegration programs in this population.

Lessons learned from the Fukushima Dai-ichi accident and responses in NRA regulatory requirements

International Nuclear Information System (INIS)

Fuketa, Toyoshi

2014-01-01

The author would like to present significant lessons learned from the TEPCO’s Fukushima Dai-ichi accident and responses in regulatory requirements developed by the Nuclear Regulation Authority for power-producing light water reactors. The presentation will cover prevention of structures, systems and components failures, measures to prevent common cause failures, prevention of core damage, mitigation of severe accidents, emergency preparedness, continuous improvement of safety, use of probabilistic risk assessment, and post-accident regulation on the Fukushima Dai-ichi. (author)

TRAC calculations of a loss-of-coolant accident in a reactor scale model

International Nuclear Information System (INIS)

Pyun, J.J.

1981-01-01

The TRAC (Transient Reactor Analysis Code) is being developed at the Los Alamos National Laboratory as an advanced best-estimate computer program for analysis of postulated hypothetical accidents in pressurized water reactors. As a part of the TRAC developmental verification efforts, a TRAC posttest analysis of Semiscale Mod-3 Test S-07-6 was conducted. The results of this analysis show that the agreement between TRAC calculations and experimental data is not very good. In particular, TRAC does not predict the long term doncomer and core liquid level oscillations during the reflood phase

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

Pilot program: NRC severe reactor accident incident response training manual. Overview and summary of major points

International Nuclear Information System (INIS)

McKenna, T.J.; Martin, J.A. Jr.; Giitter, J.G.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Watkins

1987-02-01

Overview and Summary of Major Points is the first in a series of volumes that collectively summarize the U.S. Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume describes elementary perspectives on severe accidents and accident assessment. Other volumes in the series are: Volume 2-Severe Reactor Accident Overview; Volume 3- Response of Licensee and State and Local Officials; Volume 4-Public Protective Actions-Predetermined Criteria and Initial Actions; Volume 5 - U.S. Nuclear Regulatory Commission. Each volume serves, respectively, as the text for a course of instruction in a series of courses for NRC response personnel. These materials do not provide guidance or license requirements for NRC licensees. The volumes have been organized into these training modules to accommodate the scheduling and duty needs of participating NRC staff. Each volume is accompanied by an appendix of slides that can be used to present this material

Optimization of emergency response to major nuclear accidents

International Nuclear Information System (INIS)

Papazoglou, I.A.; Christou, M.D.

1991-01-01

A methodology for the optimization of the short-term emergency response in the event of a nuclear accident has been developed. The method aims at an optimum combination of protective actions in the presence of a multitude of conflicting objectives and under uncertainty. Conflicting objectives arise when the minimization of the potential adverse effects of an accident and the simultaneous minimization of the associated socioeconomic impacts is attempted. Additional conflicting objectives appear whenever an emergency plan tends to decrease a particular health effect (e.g. acute deaths) while at the same time it increases another (e.g. latent deaths). The uncertainty is due to the multitude of the possible accident scenarios and their respective probability of occurrence, the stochastic variability in the weather conditions and in the variability and/or lack of knowledge in the parameters of the risk assessment models. A multiobjective optimization approach is adopted in a dynamic programming scheme. An emergency protective plan consists of defining a protective action (e.g. evacuation, sheltering) at each spatial cell around the plant. Three criteria (evaluators) are used as the objective functions of the problem, namely, acute fatalities, latent effects and socioeconomic cost. The optimization procedure defines the efficient frontier, i.e. all emergency plans that are not dominated by another in all three criteria. No value trade-offs are necessary up to this point

Specific features of RBMK severe accidents progression and approach to the accident management

International Nuclear Information System (INIS)

Vasilevskij, V.P.; Nikitin, Yu.M.; Petrov, A.A.; Potapov, A.A.; Cherkashov, Yu.M.

2001-01-01

Fundamental construction features of the LWGR facilities (absence of common external containment shell, disintegrated circulation circuit and multichannel reactor core, positive vapor reactivity coefficient, high mass of thermally capacious graphite moderator) predetermining development of assumed heavy non-projected accidents and handling them are treated. Rating the categories of the reactor core damages for non-projected accidents and accident types producing specific grope of damages is given. Passing standard non-projected accidents, possible methods of attack accident consequences, as well as methods of calculated analysis of non-projected accidents are demonstrated [ru

Accident and Off-Normal Response and Recovery from Multi-Canister Overpack (MCO) Processing Events

International Nuclear Information System (INIS)

ALDERMAN, C.A.

2000-01-01

In the process of removing spent nuclear fuel (SNF) from the K Basins through its subsequent packaging, drymg, transportation and storage steps, the SNF Project must be able to respond to all anticipated or foreseeable off-normal and accident events that may occur. Response procedures and recovery plans need to be in place, personnel training established and implemented to ensure the project will be capable of appropriate actions. To establish suitable project planning, these events must first be identified and analyzed for their expected impact to the project. This document assesses all off-normal and accident events for their potential cross-facility or Multi-Canister Overpack (MCO) process reversal impact. Table 1 provides the methodology for establishing the event planning level and these events are provided in Table 2 along with the general response and recovery planning. Accidents and off-normal events of the SNF Project have been evaluated and are identified in the appropriate facility Safety Analysis Report (SAR) or in the transportation Safety Analysis Report for Packaging (SARP). Hazards and accidents are summarized from these safety analyses and listed in separate tables for each facility and the transportation system in Appendix A, along with identified off-normal events. The tables identify the general response time required to ensure a stable state after the event, governing response documents, and the events with potential cross-facility or SNF process reversal impacts. The event closure is predicated on stable state response time, impact to operations and the mitigated annual occurrence frequency of the event as developed in the hazard analysis process

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

International Nuclear Information System (INIS)

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

1988-07-01

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

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)

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

International Nuclear Information System (INIS)

Ostmeyer, R.M.

1982-01-01

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

Spent nuclear fuel structural response when subject to an end impact accident

Energy Technology Data Exchange (ETDEWEB)

Tang, D.T.; Guttmann, J. [United States Nuclear Regulatory Commission, Rockville, MD (United States)]|[United States Nuclear Regulatory Commission, Washington, DC (United States); Koeppel, B.J.; Adkins, H.E.

2004-07-01

The US Nuclear Regulatory Commission (USNRC) is responsible for licensing spent fuel storage and transportation systems. A subset of this responsibility is to investigate and understand the structural performance of these systems. Studies have shown that the fuel rods of intact spent fuel assemblies with burn-ups up to 45 gigawatt days per metric ton of uranium (Gwd/MTU) are capable of resisting the normally expected impact loads subjected during drop accident conditions. However, effective cladding thickness for intact spent fuel assemblies with burn ups greater than 45 Gwd/MTU can be reduced due to corrosion. The capability of the fuel rod to withstand the expected loads encountered under normal and accident conditions may also be reduced, given degradation of the material properties under extended use, such as decrease in ductility. The USNRC and Pacific Northwest Laboratory (PNNL) performed computational studies to predict the structural response of spent nuclear fuel in a transport system that is subjected to a hypothetical regulatory impact accident, as defined in 10 CFR71.73. This study performs a structural analysis of a typical high burn up Pressurized Water Reactor (PWR) fuel assembly using the ANSYS {sup registered} ANSYS {sup registered} /LS- DYNA {sup registered} finite element analysis (FEA) code. The material properties used in the analyses were based on expert judgment and included uncertainties. Ongoing experimental programs will reduce the uncertainties. The current evaluations include the pins, spacer grids, and tie plates to assess possible cladding failure/rupture under hypothetical impact accident loading. This paper describes the USNRC and PNNL staff's analytical approach, provides details on the single pin model developed for this assessment, and presents the results.

Package of programs for calculating accidents involving melting of the materials in a fast-reactor vessel

International Nuclear Information System (INIS)

Vlasichev, G.N.

1994-01-01

Methods for calculating one-dimensional nonstationary temperature distribution in a system of physically coupled materials are described. Six computer programs developed for calculating accident processes for fast reactor core melt are described in the article. The methods and computer programs take into account melting, solidification, and, in some cases, vaporization of materials. The programs perform calculations for heterogeneous systems consisting of materials with arbitrary but constant composition and heat transfer conditions at material boundaries. Additional modules provide calculations of specific conditions of heat transfer between materials, the change in these conditions and configuration of the materials as a result of coolant boiling, melting and movement of the fuel and structural materials, temperature dependences of thermophysical properties of the materials, and heat release in the fuel. 11 refs., 3 figs

Application of FFTBM to severe accidents

International Nuclear Information System (INIS)

Prosek, A.; Leskovar, M.

2005-01-01

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

Response of a DSNP pressurizer model under accident conditions

International Nuclear Information System (INIS)

Saphier, D.; Kallfelz, J.; Belblidia, L.

1986-01-01

Recently a new pressurizer model was developed for the DSNP simulation language. The model was connected to a simulation of the Trojan pressurized water reactor (PWR) and tested by simulating a loss-of-off-site power (LOSP) anticipated transient without scram. The results compare well to a similar study performed using the RELAP code. The pressurizer model and its response to the LOSP accident are presented

Final disposal room structural response calculations

International Nuclear Information System (INIS)

Stone, C.M.

1997-08-01

Finite element calculations have been performed to determine the structural response of waste-filled disposal rooms at the WIPP for a period of 10,000 years after emplacement of the waste. The calculations were performed to generate the porosity surface data for the final set of compliance calculations. The most recent reference data for the stratigraphy, waste characterization, gas generation potential, and nonlinear material response have been brought together for this final set of calculations

Cognitive modeling and dynamic probabilistic simulation of operating crew response to complex system accidents

International Nuclear Information System (INIS)

Chang, Y.H.J.; Mosleh, A.

2007-01-01

This is the last in a series of five papers that discuss the Information Decision and Action in Crew (IDAC) context for human reliability analysis (HRA) and example application. The model is developed to probabilistically predict the responses of the control room operating crew in nuclear power plants during an accident, for use in probabilistic risk assessments (PRA). The operator response spectrum includes cognitive, emotional, and physical activities during the course of an accident. This paper describes a dynamic PRA computer simulation program, accident dynamics simulator (ADS), developed in part to implement the IDAC model. This paper also provides a detailed example of implementing a simpler version of IDAC, compared with the IDAC model discussed in the first four papers of this series, to demonstrate the practicality of integrating a detailed cognitive HRA model within a dynamic PRA framework

Emergency preparedness and response to 'Not-in-a-Facility' radiological accidents

International Nuclear Information System (INIS)

Grlicarev, Igor

2008-01-01

The paper provides an overview of lessons learned from the past radiological accidents, which have not occurred in an operating facility, i.e. 'not-in-a-facility' radiological emergencies. A method to analyze status of prevention of accidents is proposed taking into account the experiences and findings from the past events. The main emergency planning items are discussed, which would render effective response in case of such emergencies. Although the IAEA has published many documents about establishing an adequate emergency response capability, it is not an easy task to bring these recommendations into life. This paper gives some hints how to overcome the most obvious difficulties while users of these documents trying to adapt the guidance to their own needs. The special cases of alpha emitters and radiological dispersal devices were considered separately. The balanced approach to emergency response is promoted throughout the text, which means that a level of preparedness should be commensurate to the threat and the existing resources should be used to the extent possible. (author)

Microbial aerosol generation during laboratory accidents and subsequent risk assessment.

Science.gov (United States)

Bennett, A; Parks, S

2006-04-01

To quantify microbial aerosols generated by a series of laboratory accidents and to use these data in risk assessment. A series of laboratory accident scenarios have been devised and the microbial aerosol generated by them has been measured using a range of microbial air samplers. The accident scenarios generating the highest aerosol concentrations were, dropping a fungal plate, dropping a large bottle, centrifuge rotor leaks and a blocked syringe filter. Many of these accidents generated low particle size aerosols, which would be inhaled into the lungs of any exposed laboratory staff. Spray factors (SFs) have been calculated using the results of these experiments as an indicator of the potential for accidents to generate microbial aerosols. Model risk assessments have been described using the SF data. Quantitative risk assessment of laboratory accidents can provide data that can aid the design of containment laboratories and the response to laboratory accidents. A methodology has been described and supporting data provided to allow microbiological safety officers to carry out quantitative risk assessment of laboratory accidents.

Detection, diagnosis, and treatment of accident conditions using response trees

International Nuclear Information System (INIS)

Nelson, W.R.

1980-01-01

Response Trees were developed at the LOFT facility in 1978 and included in the Plant Operating Manual (POM) to assist reactor operators in selecting emergency procedures. In an emergency situation the operator would manually gather data and evaluate the trees to select the appropriate procedures. As a portion of the LOFT Augmented Operator Capability (AOC) Program, the response tree methodology has been extended so that a computer can be used to evaluate the trees and recommend an appropriate response for an accident. Techniques for diagnosing failures within a cooling mode have also been investigated. This paper summarizes these additions to the response tree methodology

Modular telerobot control system for accident response

Science.gov (United States)

Anderson, Richard J. M.; Shirey, David L.

1999-08-01

The Accident Response Mobile Manipulator System (ARMMS) is a teleoperated emergency response vehicle that deploys two hydraulic manipulators, five cameras, and an array of sensors to the scene of an incident. It is operated from a remote base station that can be situated up to four kilometers away from the site. Recently, a modular telerobot control architecture called SMART was applied to ARMMS to improve the precision, safety, and operability of the manipulators on board. Using SMART, a prototype manipulator control system was developed in a couple of days, and an integrated working system was demonstrated within a couple of months. New capabilities such as camera-frame teleoperation, autonomous tool changeout and dual manipulator control have been incorporated. The final system incorporates twenty-two separate modules and implements seven different behavior modes. This paper describes the integration of SMART into the ARMMS system.

Economic and social impacts of nuclear accidents on the agricultural sector

International Nuclear Information System (INIS)

Brenot, J.; Hubert, P.

1997-01-01

The economic and social impact of a major nuclear accident on the agricultural sector are reviewed. The associated costs are evaluated by more or less proper methods depending on the duration and severity of the post accident situation. Calculating such costs is necessary in order to allow farmers, farm-food enterprises, and public authorities to define the indemnification levels as well as to identify means of minimizing the accident consequences. The indemnification procedures are described in a section dedicated to liability issues and the costs due to Chernobyl accident. Concerning the limitation of accident consequences the responsibility falls upon the public authorities. In regard for decision making the existent methods vary according to the situation complexity and proposed objectives. Examples are given to point out the costs and social impact

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

International Nuclear Information System (INIS)

Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

2013-01-01

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

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

International Nuclear Information System (INIS)

Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

2014-01-01

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

Accident progression event tree analysis for postulated severe accidents at N Reactor

International Nuclear Information System (INIS)

Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M.; Medford, G.T.

1990-06-01

A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied

Experiments on the lower plenum response during a severe accident

International Nuclear Information System (INIS)

Henry, Robert E.; Hammersley, Robert J.; Klopp, George T.; Merilo, Mati

2004-01-01

Severe accident evaluations for nuclear reactors consider the response when the core materials have been overheated sufficient to melt and change geometry. One possible consequence of this is that molten core debris could drain into the lower plenum, as occurred in the TMI-2 accident. Given this state, several physical processes need to be analyzed, i.e. the extent of debris particulation and cooling, the potential for thermal attack of lower plenum structures, the thermal transient of the RPV and the potential for external cooling of the RPV lower head. These are important and complex processes, the evaluations of which need to be guided by well founded experiments. To support the development of the MAAP codes, recent experiments have been performed on specific issues such as: 1. the response of lower head penetrations submerged in a high temperature melt, 2. the net steam generation rate when molten debris drains into the lower plenum, 3. the formation of a contact resistance when molten debris drains through water and contacts the RPV wall and 4. the potential for external cooling of the RPV lower head. This paper discusses these experiments and their results. More importantly, it discusses how these are used in formulating models to represent the lower plenum response in the MAAP codes. (author)

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)

Guidance on accidents involving radioactivity

International Nuclear Information System (INIS)

1989-01-01

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

Normal accidents

International Nuclear Information System (INIS)

Perrow, C.

1989-01-01

The author has chosen numerous concrete examples to illustrate the hazardousness inherent in high-risk technologies. Starting with the TMI reactor accident in 1979, he shows that it is not only the nuclear energy sector that bears the risk of 'normal accidents', but also quite a number of other technologies and industrial sectors, or research fields. The author refers to the petrochemical industry, shipping, air traffic, large dams, mining activities, and genetic engineering, showing that due to the complexity of the systems and their manifold, rapidly interacting processes, accidents happen that cannot be thoroughly calculated, and hence are unavoidable. (orig./HP) [de

Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

International Nuclear Information System (INIS)

Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

2000-01-01

Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

An Analysis of Reactor Structural Response to Fuel Sodium Interaction in a Hypothetical Core Disruptive Accident

International Nuclear Information System (INIS)

Suzuki, K.; Tashiro, M.; Sasanuma, K.; Nagashima, K.

1976-01-01

This study shows the effect of constraints around FSI zone on FSI phenomena and deformations of reactor structures. SUGAR-PISCES code system has been developed to evaluate the phenomena of FSI and the response of reactor structure. SUGAR calculates the phenomena of FSI. PISCES, developed by Physics International Company in U.S.A, calculates the dynamic response of reactor structure in two-dimensional, time-dependent finite-difference Lagrangian model. The results show that the peak pressure and energy by FSI and the deformation of reactor structures are about twice in case of FSI zone surrounding by blanket than by coolant. The FSI phenomena highly depend on the reactor structure and the realistic configuration around core must be considered for analyzing hypothetical core disruptive accident. In conclusion: FSI phenomena depend highly on constraints around FSI zone, so that the constraints must be dealt with realistically in analytical models. Although a two-dimensional model is superior to a quasi-two-dimensional model. The former needs long calculation time, so it is very expensive using in parametric study. Therefore, it is desirable that the two-dimensional model is used in the final study of reactor design and the quasi-two-dimensional model is used in parametric study. The blanket affects on the acoustic pressure and the deformations of radial structures, but affects scarcely on the upper vessel deformation. The blanket also affects on the mechanical work largely. The core barrel gives scarcely the effects on pressure in single phase but gives highly the effects on pressure in two-phase and deformation of reactor structures in this study. For studying the more realistic phenomena of FSI in the reactor design, the following works should be needed. (i) Spatial Distribution of FSI Region Spatial and time-dependent distribution of fuel temperature and molten fuel fraction must be taken in realistic simulation of accident condition. To this purpose, the code will

Accident at Chernobyl and the medical response

International Nuclear Information System (INIS)

Geiger, H.J.

1986-01-01

The author was in the Soviet Union in early June 1986, leading a medical lecture tour under an exchange program sponsored by Physicians for Social Responsibility. This provided an opportunity for extensive discussions with the Soviet physicians in charge of the medical response to Chernobyl, for a visit to Moscow Hospital number 6, the center of care for those acutely injured for observation of seven acutely irradiated patients and reviews of their clinical courses, and for discussion with the medical teams providing the acute care and planning the necessary long term epidemiologic and environmental investigations. This report is based on information provided by these sources and on data released in Moscow by Robert P. Gale, MD, the American physician from UCLA who, with his associates, flew to the Soviet Union within days to join the team already caring for irradiated victims of the accident

Estimate of the economy-wide financial risk associated with a serious nuclear reactor accident

International Nuclear Information System (INIS)

Heising, C.D.; George, V.P.

1985-01-01

Since the accident at Three Mile Island, considerable attention has focused on the financial risks associated with nuclear reactor accidents. Previous studies have calculated the plant-specific financial risk due to primarily off-site consequences, including most probabilistic risk assessments (PRAs) beginning first with the Reactor Safety Study. Previous studies, however, have not studied the possible economy-wide costs of a severe reactor accident due to such considerations as adverse public opinion, enhanced concern in the financial community, and, particularly, regulatory constraints placed on the industry subsequent to an accident. These considerations could lead to a spectrum of scenarios, including partial or complete nuclear moratoria, which may have far reaching economy-wide effects. It is the purpose of this study to provide an estimate of these effects, and to compare these estimates with the plant-specific financial risks calculated earlier by others. First, a model is presented for including the impact of various regulatory response options on the nuclear industry, followed by a discussion of how the economy-wide costs associated with these options may be calculated. Then, results of series of calculations are presented, and finally, conclusions are given

Crisis exercises are necessary for an efficient response in case of accident

International Nuclear Information System (INIS)

Loyen, J.

2016-01-01

In France an average of 10 emergency exercises simulating reactor or radiation accidents are organized every year at a national level. They are the only opportunity to test material and humane resources and most importantly they test the reactivity and the coordination of all the actors involved in the response to an accidental situation. Accident simulation scenarios are established by IRSN (Institute for Radiation Protection and Nuclear Safety) and the Ministry of the Interior and they cover both aspects of a severe nuclear accident: the emergency situation and the post-accidental situation. The feedback experience of these exercises but also real situations have led to a better taking into account of some issues like the necessary consistency of the information delivered to the media and the public, the involvement of the civil society or trans-national cooperation. (A.C.)

Dose-Response Calculator for ArcGIS

Science.gov (United States)

Hanser, Steven E.; Aldridge, Cameron L.; Leu, Matthias; Nielsen, Scott E.

2011-01-01

The Dose-Response Calculator for ArcGIS is a tool that extends the Environmental Systems Research Institute (ESRI) ArcGIS 10 Desktop application to aid with the visualization of relationships between two raster GIS datasets. A dose-response curve is a line graph commonly used in medical research to examine the effects of different dosage rates of a drug or chemical (for example, carcinogen) on an outcome of interest (for example, cell mutations) (Russell and others, 1982). Dose-response curves have recently been used in ecological studies to examine the influence of an explanatory dose variable (for example, percentage of habitat cover, distance to disturbance) on a predicted response (for example, survival, probability of occurrence, abundance) (Aldridge and others, 2008). These dose curves have been created by calculating the predicted response value from a statistical model at different levels of the explanatory dose variable while holding values of other explanatory variables constant. Curves (plots) developed using the Dose-Response Calculator overcome the need to hold variables constant by using values extracted from the predicted response surface of a spatially explicit statistical model fit in a GIS, which include the variation of all explanatory variables, to visualize the univariate response to the dose variable. Application of the Dose-Response Calculator can be extended beyond the assessment of statistical model predictions and may be used to visualize the relationship between any two raster GIS datasets (see example in tool instructions). This tool generates tabular data for use in further exploration of dose-response relationships and a graph of the dose-response curve.

The sensitivity of calculated doses to critical assumptions for the offsite consequences of nuclear power reactor accidents

International Nuclear Information System (INIS)

Moeller, M.P.; Scherpelz, R.I.; Desrosiers, A.E.

1982-01-01

This work analyzes the sensitivity of calculated doses to critical assumptions for offsite consequences following a PWR-2 accident at a nuclear power reactor. The calculations include three radiation dose pathways: internal dose resulting from inhalation, external doses from exposure to the plume, and external doses from exposure to contaminated ground. The critical parameters are the time period of integration for internal dose commitment and the duration of residence on contaminated ground. The data indicate the calculated offsite whole body dose will vary by as much as 600% depending upon the parameters assumed. When offsite radiation doses determine the size of emergency planning zones, this uncertainty has significant effect upon the resources allocated to emergency preparedness

Legislative framework on establishing emergency response plan in the case of a nuclear accident

International Nuclear Information System (INIS)

Novosel, N.; Valcic, I.; Biscan, R.

2000-01-01

To give an overview of the legislative framework, which defined emergency planning in Croatia in the case of a nuclear accident, it's necessary to look at all international recommendations and obligations and the national legislation, acts and regulations. It has to be emphasized that Croatia signed three international conventions in this field, and by that took over some responsibilities and obligations. Beside that, it is also in Croatian interest to follow the recommendations of international institutions such as International Atomic Energy Agency (IAEA standards and technical documents). On the other hand, national legislation in this field consists of several laws, which cover nuclear safety measures, governmental organization, natural disasters and acts (decree, decisions) of responsible authority for emergency planning in the case of a nuclear accident (Ministry of Economy). This paper presents an overview of the international and Croatian legislation which influenced the emergency planning in the case of a nuclear accident. (author)

Support calculations for management of PRISE leakage accidents

Energy Technology Data Exchange (ETDEWEB)

Matejovic, P.; Vranka, L. [Nuclear Power Plants Research Inst. Vuje, Trnava (Slovakia)

1997-12-31

Accidents involving primary-to-secondary leakage (PRISE) caused by rupture of one or a few tubes are well known design basis events in both, western and VVER NPPs. Operating experience and in-service inspections of VVER-440 units have demonstrated also the potential for large PRISE leaks in the case of the steam generator (SG) primary collector cover lift-up (Rovno NPP). Without performing any countermeasure for limitation of SG collector cover lift-up, a full opening results in PRISE leak with an equivalent diameter 107 mm. Although this accident was not considered in the original design, this event is usually analysed as DBA too. Different means are available for detection and mitigation of PRISE leakage in NPPs currently in operation (J.Bohunice V-1 and V-2) or under construction (Mochovce) in Slovakia. 8 refs.

Support calculations for management of PRISE leakage accidents

Energy Technology Data Exchange (ETDEWEB)

Matejovic, P; Vranka, L [Nuclear Power Plants Research Inst. Vuje, Trnava (Slovakia)

1998-12-31

Accidents involving primary-to-secondary leakage (PRISE) caused by rupture of one or a few tubes are well known design basis events in both, western and VVER NPPs. Operating experience and in-service inspections of VVER-440 units have demonstrated also the potential for large PRISE leaks in the case of the steam generator (SG) primary collector cover lift-up (Rovno NPP). Without performing any countermeasure for limitation of SG collector cover lift-up, a full opening results in PRISE leak with an equivalent diameter 107 mm. Although this accident was not considered in the original design, this event is usually analysed as DBA too. Different means are available for detection and mitigation of PRISE leakage in NPPs currently in operation (J.Bohunice V-1 and V-2) or under construction (Mochovce) in Slovakia. 8 refs.

NPP Krsko Severe Accident Management Guidelines Implementation

International Nuclear Information System (INIS)

Basic, I.; Krajnc, B.; Bilic-Zabric, T.; Spiler, J.

2002-01-01

Severe Accident Management is a framework to identify and implement the Emergency Response Capabilities that can be used to prevent or mitigate severe accidents and their consequences. The USA NRC has indicated that the development of a licensee plant specific accident management program will be required in order to close out the severe accident regulatory issue (Ref. SECY-88-147). Generic Letter 88-20 ties the Accident management Program to IPE for each plant. The SECY-89-012 defines those actions taken during the course of an accident by the plant operating and technical staff to: 1) prevent core damage, 2) terminate the progress of core damage if it begins and retain the core within the reactor vessel, 3) maintain containment integrity as long as possible, and 4) minimize offsite releases. The subject of this paper is to document the severe accident management activities, which resulted in a plant specific Severe Accident Management Guidelines implementation. They have been developed based on the Krsko IPE (Individual Plant Examination) insights, Generic WOG SAMGs (Westinghouse Owners Group Severe Accident Management Guidances) and plant specific documents developed within this effort. Among the required plant specific actions the following are the most important ones: Identification and documentation of those Krsko plant specific severe accident management features (which also resulted from the IPE investigations). The development of the Krsko plant specific background documents (Severe Accident Plant Specific Strategies and SAMG Setpoint Calculation). Also, paper discusses effort done in the areas of NPP Krsko SAMG review (internal and external ), validation on Krsko Full Scope Simulator (Severe Accident sequences are simulated by MAAP4 in real time) and world 1st IAEA Review of Accident Management Programmes (RAMP). (author)

Accident analysis for PRC-II reactor

International Nuclear Information System (INIS)

Wei Yongren; Tang Gang; Wu Qing; Lu Yili; Liu Zhifeng

1997-12-01

The computer codes, calculation models, transient results, sensitivity research, design improvement, and safety evaluation used in accident analysis for PRC-II Reactor (The Second Pulsed Reactor in China) are introduced. PRC-II Reactor is built in big populous city, so the public pay close attention to reactor safety. Consequently, Some hypothetical accidents are analyzed. They include an uncontrolled control rod withdrawal at rated power, a pulse rod ejection at rated power, and loss of coolant accident. Calculation model which completely depict the principle and process for each accident is established and the relevant analysis code is developed. This work also includes comprehensive computing and analyzing transients for each accident of PRC-II Reactor; the influences in the reactor safety of all kind of sensitive parameters; evaluating the function of engineered safety feature. The measures to alleviate the consequence of accident are suggested and taken in the construction design of PRC-II Reactor. The properties of reactor safety are comprehensively evaluated. A new advanced calculation model (True Core Uncovered Model) of LOCA of PRC-II Reactor and the relevant code (MCRLOCA) are first put forward

The Fukushima Daiichi Accident. Technical Volume 3/5. Emergency Preparedness and Response. Annexes

International Nuclear Information System (INIS)

2015-08-01

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

Development of an anthropomorphic model and a Monte Carlo calculation code devoted to the physical reconstruction of a radiological accident

International Nuclear Information System (INIS)

Roux, A.

2001-01-01

The diversity of radiological accidents makes difficult the medical prognosis and the therapy choice from only clinical observations. To complete this information, it is important to know the global dose received by the organism and the dose distributions in depth in tissues. The dose estimation can be made by a physical reconstruction of the accident with the help of tools based on experimental techniques or on calculation. The software of the geometry construction (M.G.E.D.), associated to the Monte-Carlo code of photons and neutrons transport (M.O.R.S.E.) replies these constraints. An important result of this work is to determine the principal parameters to know in function of the accident type, as well as the precision level required for these parameters. (N.C.)

CONTEMPT: computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

International Nuclear Information System (INIS)

Hsii, Y.H.

1978-04-01

The CONTEMPT code is used by Babcock and Wilcox for containment analysis following a postulated loss of coolant accident. An additional model is described which is used for the calculation of long term post reflood mass and energy releases to the containment that is used for the containment design basis LOCA calculations. These calculations maximize the rate of energy flow to the containment. The mass and energy data are given to the containment designer for use in calculating the containment building design pressure and temperature and in sizing containment heat removal equipment

Planning for large-scale accidents: learning from the Three Mile Island accident

International Nuclear Information System (INIS)

Fischer, D.W.

1981-01-01

Decision-making issues raised at the Three Mile Island nuclear accident in Pennsylvania are explored. The organizations involved, their interconnections, and decisions are described. The underlying issues bearing on allocation of effort to pre-accident planning and actual accident responses are also noted. Finally, a framework from this effort is used for guiding the planning of operations for future accidents. (author)

HANARO thermal hydraulic accident analysis

Energy Technology Data Exchange (ETDEWEB)

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

1996-06-01

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

Safety analysis of accident localization system

International Nuclear Information System (INIS)

1999-01-01

A complex safety analysis of accident localization system of Ignalina NPP was performed. Calculation results obtained, results of non-destruct ing testing and experimental data of reinforced concrete testing of buildings does not revealed deficiencies of buildings of accident localization system at unit 1 of Ignalina NPP. Calculations were performed using codes NEPTUNE, ALGOR, CONTAIN

CONTAIN calculations

International Nuclear Information System (INIS)

Scholtyssek, W.

1995-01-01

In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident 'medium-sized leak in the cold leg', especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)

Planning for off-site response to radiation accidents in nuclear facilities

International Nuclear Information System (INIS)

1981-01-01

The purpose of this publication is to give guidance to those who are responsible for the protection of the public in the event of an accident occurring at a land-based nuclear facility. This guidance should assist in the advance preparation of emergency response plans and implementing procedures. Basic principles of protective measures along with their advantages and disadvantages are discussed. Other principles related to emergency planning and the operational response to an emergency are outlined. Although the guidance is primarily oriented towards land-based nuclear power facilities, the guidance does have general application to other types of nuclear facility

Planning for off-site response to radiation accidents in nuclear facilities

International Nuclear Information System (INIS)

1979-01-01

The purpose of this manual is to give guidance to those who are responsible for the protection of the public in the event of an accident occurring at a land-based nuclear facility. This guidance should assist in the advance preparation of emergency response plans and implementing procedures. Basic principles of protective measures along with their advantages and disadvantages are discussed. Other principles related to emergency planning and the operational response to an emergency are outlined. Although the guidance is primarily oriented toward land-based nuclear power facilities, the guidance does have general application to other types of nuclear facilities

Emergency response planning for transport accidents involving radioactive materials

International Nuclear Information System (INIS)

1982-03-01

The document presents a basic discussion of the various aspects and philosophies of emergency planning and preparedness along with a consideration of the problems which might be encountered in a transportation accident involving a release of radioactive materials. Readers who are responsible for preparing emergency plans and procedures will have to decide on how best to apply this guidance to their own organizational structures and will also have to decide on an emergency planning and preparedness philosophy suitable to their own situations

On the contribution of external cost calculations to energy system governance: The case of a potential large-scale nuclear accident

International Nuclear Information System (INIS)

Laes, Erik; Meskens, Gaston; Sluijs, Jeroen P. van der

2011-01-01

The contribution of nuclear power to a sustainable energy future is a contested issue. This paper presents a critical review of an attempt to objectify this debate through the calculation of the external costs of a potential large-scale nuclear accident in the ExternE project. A careful dissection of the ExternE approach resulted in a list of 30 calculation steps and assumptions, from which the 6 most contentious ones were selected through a stakeholder internet survey. The policy robustness and relevance of these key assumptions were then assessed in a workshop using the concept of a 'pedigree of knowledge'. Overall, the workshop outcomes revealed the stakeholder and expert panel's scepticism about the assumptions made: generally these were considered not very plausible, subjected to disagreement, and to a large extent inspired by contextual factors. Such criticism indicates a limited validity and useability of the calculated nuclear accident externality as a trustworthy sustainability indicator. Furthermore, it is our contention that the ExternE project could benefit greatly - in terms of gaining public trust - from employing highly visible procedures of extended peer review such as the pedigree assessment applied to our specific case of the external costs of a potential large-scale nuclear accident. - Highlights: → Six most contentious assumptions were selected through a stakeholder internet survey. → Policy robustness of these assumptions was assessed in a pedigree assessment workshop. → Assumptions were considered implausible, controversial, and inspired by contextual factors. → This indicates a limited validity and useability as a trustworthy sustainability indicator.

Accident management insights after the Fukushima Daiichi NPP accident

International Nuclear Information System (INIS)

Degueldre, Didier; Viktorov, Alexandre; Tuomainen, Minna; Ducamp, Francois; Chevalier, Sophie; Guigueno, Yves; Tasset, Daniel; Heinrich, Marcus; Schneider, Matthias; Funahashi, Toshihiro; Hotta, Akitoshi; Kajimoto, Mitsuhiro; Chung, Dae-Wook; Kuriene, Laima; Kozlova, Nadezhda; Zivko, Tomi; Aleza, Santiago; Jones, John; McHale, Jack; Nieh, Ho; Pascal, Ghislain; ); Nakoski, John; Neretin, Victor; Nezuka, Takayoshi; )

2014-01-01

The Fukushima Daiichi nuclear power plant (NPP) accident, that took place on 11 March 2011, initiated a significant number of activities at the national and international levels to reassess the safety of existing NPPs, evaluate the sufficiency of technical means and administrative measures available for emergency response, and develop recommendations for increasing the robustness of NPPs to withstand extreme external events and beyond design basis accidents. The OECD Nuclear Energy Agency (NEA) is working closely with its member and partner countries to examine the causes of the accident and to identify lessons learnt with a view to the appropriate follow-up actions to be taken by the nuclear safety community. Accident management is a priority area of work for the NEA to address lessons being learnt from the accident at the Fukushima Daiichi NPP following the recommendations of Committee on Nuclear Regulatory Activities (CNRA), Committee on the Safety of Nuclear Installations (CSNI), and Committee on Radiation Protection and Public Health (CRPPH). Considering the importance of these issues, the CNRA authorised the formation of a task group on accident management (TGAM) in June 2012 to review the regulatory framework for accident management following the Fukushima Daiichi NPP accident. The task group was requested to assess the NEA member countries needs and challenges in light of the accident from a regulatory point of view. The general objectives of the TGAM review were to consider: - enhancements of on-site accident management procedures and guidelines based on lessons learnt from the Fukushima Daiichi NPP accident; - decision-making and guiding principles in emergency situations; - guidance for instrumentation, equipment and supplies for addressing long-term aspects of accident management; - guidance and implementation when taking extreme measures for accident management. The report is built on the existing bases for capabilities to respond to design basis

Emergency preparedness and response in case of a fire accident with (UF6) packages tracking Suez Canal

International Nuclear Information System (INIS)

Salama, M.

2004-01-01

Egypt has a unique problem - the Suez Canal. Radioactive cargo passing regularly through the canal carrying new and spent reactor fuel. Moreover there are also about 1000 metric tons of uranium hexaflouride (UF6) passing through the canal every year. In spite of all precautions taken in the transportation, accidents with packages containing (UF 6 ) and shipped through the Suez Canal, accidents may arise even though the probability is minimal. These accidents, may be accompanied by injuries or death of persons and damage to property. Due to the radiation and criticality hazards of (UF 6 ) and its high risk of chemical toxicity. The probability of a fire accident with a cargo carrying (UF 6 ) during its crossing the Suez Canal can cause serious chemical toxic and radiological hazards, particularly if the accident occurred close or near to one of the three densely populated cities (Port-Said, Ismailia, and Suez), which are located along the Suez Canal, west bank. The government of Egypt has elaborated a national radiological emergency plan inorder to face probable radiological accidents, which may be arised inside the country. Arrangements have been also elaborated for the medical care of any persons who, might be injured or contaminated, or who, have been exposed to severe radiation doses. The motivation of the present paper was undertaken to visualize a fire accident scenario occurring in industrial packages containing UF6 on board of a Cargo crossing the Suez Canal near Port-Said City. The accident scenario and emergency response actions taken during the different phases of the accident are going to be presented and discussed. The proposed emergency response actions taken to face the accident are going to be also presented. The work presented had revealed the importance of public awareness will be needed for populations located in densely populated areas along Suez Canal bank inorder to react timely and effectively to avoid the toxic and radiological hazards

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

International Nuclear Information System (INIS)

Margeanu, Sorin; Tatiana, Angelescu

2005-01-01

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

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Guidelines for calculating radiation doses to the public from a release of airborne radioactive material under hypothetical accident conditions in nuclear reactors

International Nuclear Information System (INIS)

1991-04-01

This standard provides guidelines and a methodology for calculating effective doses and thyroid doses to people (either individually or collectively) in the path of airborne radioactive material released from a nuclear facility following a hypothetical accident. The radionuclides considered are those associated with substances having the greatest potential for becoming airborne in reactor accidents: tritium (HTO), noble gases and their daughters, radioiodines, and certain radioactive particulates (Cs, Ru, Sr, Te). The standard focuses on the calculation of radiation doses for external exposures from radioactive material in the cloud; internal exposures for inhalation of radioactive material in the cloud and skin penetration of tritium; and external exposures from radionuclides deposited on the ground. It uses as modified Gaussian plume model to evaluate the time-integrated concentration downwind. (52 refs., 12 tabs., 21 figs.)

Accident consequence calculations for project W-058 safetyanalysis

Energy Technology Data Exchange (ETDEWEB)

Van Keuren, J.C.

1997-06-10

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

Radiological health risks from accidents during transportation of spent nuclear fuels

International Nuclear Information System (INIS)

Chen, S.Y.; Yuan, Y.C.

1988-01-01

Potential radiological health risks from severe accident scenarios during the transportation of spent nuclear fuels are estimated. These extremely low probability, but potentially credible, scenarios are characterized by the U.S. Nuclear Regulatory Commission's Modal Study in terms of the maximum credible structural responses and/or the maximum credible cask temperature responses. In some accident scenarios, the spent nuclear fuel casks are assumed to be breached, resulting in the release of radioactivity to the atmosphere. Models have been developed to estimate radiological health consequences, including potential short-term exposures and health effects to individuals and potential long-term environmental dose commitments and health effects to the population. The population risks are calculated using state-level data, and the resulting overall health risks are compared for several levels of cleanup effort to determine the relative effects on long-term risks to the population in the event of an accident. 4 refs., 3 figs., 3 tabs

A Modular Telerobot Control System for Accident Response

International Nuclear Information System (INIS)

Anderson, Robert J.; Shirey, David L.

1999-01-01

The Accident Response Mobile Manipulator System (ARMMS) is a teleoperated emergency response vehicle that deploys two hydraulic manipulators, five cameras, and an array of sensors to the scene of an incident. It is operated from a remote base station that can be situated up to four kilometers away from the site. Recently, a modular telerobot control architecture called SMART (Sandia's Modular Architecture for Robotic and Teleoperation) was applied to ARMMS to improve the precision, safety, and operability of the manipulators on board. Using SMART, a prototype manipulator control system was developed in a couple of days, and an integrated working system was demonstrated within a couple of months. New capabilities such as camera teleoperation, autonomous tool changeout and dual manipulator control have been incorporated. The final system incorporates twenty-two separate modules and implements eight different behavior modes. This paper describes the integration of SMART into the ARMMS system

Applicability of coupled code RELAP5/GOTHIC to NPP Krsko MSLB calculation

International Nuclear Information System (INIS)

Keco, M.; Debrecin, N.; Grgic, D.

2005-01-01

Usual way to analyze Main Steam Line Break (MSLB) accident in PWR plants is to calculate core and containment responses in two separate calculations. In first calculation system code is used to address behaviour of nuclear steam supply system and containment is modelled mainly as a boundary condition. In second calculation mass and energy release data are used to perform containment analysis. Coupled code R5G realized by direct explicit coupling of system code RELAP5/MOD3.3 and containment code GOTHIC is able to perform both calculations simultaneously. In this paper R5G is applied to calculation of MSLB accident in large dry containment of NPP Krsko. Standard separate calculation is performed first and then both core and containment responses are compared against corresponding coupled code results. Two versions of GOTHIC code are used, one old ver 3.4e and the last one ver 7.2. As expected, differences between standard procedure and coupled calculations are small. The performed analyses showed that classical uncoupled approach is applicable in case of large dry containment calculation, but that new approach can bring some additional insight in understanding of the transient and that can be used as simple and reliable procedure in performing MSLB calculation without any significant calculation overhead. (author)

Medical response for radiological accidents from Regional cooperation : Latin-American radiopathology networks

International Nuclear Information System (INIS)

Valverde, Nelson; Cardenas, Juan; Perez, Maria del Rosario; Trano, Jose Luiz Di; Gisone, Pablo

2001-01-01

The objective of this program is to have a system, in order to assure an appropriate medical response in the case of radiological accidents and to offer medical advice in aspects related to the biological effects of ionizing radiations in risk assessment of radiation workers, medical exposures, potential effects of prenatal irradiation

Guidelines for calculating radiation doses to the public from a release of airborne radioactive material under hypothetical accident conditions in nuclear reactors

International Nuclear Information System (INIS)

1989-09-01

This Standard provides guidelines and a methodology for calculating effective doses and thyroid doses to people (either individually or collectively) in the path of airborne radioactive material released from a nuclear facility following a hypothetical accident. The specific radionuclides considered in the Standard are those associated with substances having the greatest potential for becoming airborne in reactor accidents (eg, tritium (HTO), noble gases and their daughters (Kr-Rb, Xe-Cs), and radioiodines (I)); and certain radioactive particulates (eg, Cs, Ru, Sr, Te) that may become airborne under exceptional circumstances

Accident response -- X-ray to virtual environment

International Nuclear Information System (INIS)

Hefele, J.; Stupin, D.; Kelley, T.; Sheats, M.; Tsai, C.

1999-01-01

The Engineering Sciences and Applications (ESA) Division of Los Alamos National Laboratory (LANL) has been working to develop a process to extract topographical information from digital x-ray data for modeling in a Computer Aided Design (CAD) environment and translation into a virtual environment. The application for this process is the evolution of a field deployable tool for use by the Accident Response Group (ARG) at the Laboratory. The authors have used both CT Scan and radiography data in their process development. The data is translated into a format recognizable by Pro/ENGINEER trademark and then into a virtual environment that can be operated on by dVISE trademark. They have successfully taken both CT Scan and radiograph data of single components and created solid and virtual environment models for interrogation

Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Preliminary design report

International Nuclear Information System (INIS)

Siefken, L.J.

1999-01-01

Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ''Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents.''

Lessons learned from on-site safety assessments performed by DOE in response to the Tomsk accident

International Nuclear Information System (INIS)

Witmer, F.E.

1995-01-01

In response to the accident, in April 1993, at the nuclear fuel reprocessing plant of the Siberian chemical Combine, Tomsk, Russia, the U.S. Department of Energy (DOE) initiated concurrent efforts to understand the causes for the accident and to review potential vulnerabilities for similar occurrences across the DOE radiochemical complex. Because the accident occurred in the feed adjustment stage of a Purex type process, US facilities which contained significant inventories of TBP, organic diluent and nitric acid were evaluated by expert teams. From accident conditions, prior experience, modeling and experimental programs and confirmatory dialogue with the Russians, enhanced understanding was achieved and vulnerabilities (e.g., lack of safety analysis, organic layering, inadvertent acid addition, use of aromatic diluents, uncertain venting capability, no mitigative/emergency procedures, etc.) were identified and corrected

Emergency preparedness and response in case of a fire accident with (UF{sub 6}) packages tracking Suez Canal

Energy Technology Data Exchange (ETDEWEB)

Salama, M. [National Center for Nuclear Safety and Radiation Control (NCNSRC), Nasr City, Cairo (Egypt)

2004-07-01

Egypt has a unique problem - the Suez Canal. Radioactive cargo passing regularly through the canal carrying new and spent reactor fuel. Moreover there are also about 1000 metric tons of uranium hexaflouride (UF6) passing through the canal every year. In spite of all precautions taken in the transportation, accidents with packages containing (UF{sub 6}) and shipped through the Suez Canal, accidents may arise even though the probability is minimal. These accidents, may be accompanied by injuries or death of persons and damage to property. Due to the radiation and criticality hazards of (UF{sub 6}) and its high risk of chemical toxicity. The probability of a fire accident with a cargo carrying (UF{sub 6}) during its crossing the Suez Canal can cause serious chemical toxic and radiological hazards, particularly if the accident occurred close or near to one of the three densely populated cities (Port-Said, Ismailia, and Suez), which are located along the Suez Canal, west bank. The government of Egypt has elaborated a national radiological emergency plan inorder to face probable radiological accidents, which may be arised inside the country. Arrangements have been also elaborated for the medical care of any persons who, might be injured or contaminated, or who, have been exposed to severe radiation doses. The motivation of the present paper was undertaken to visualize a fire accident scenario occurring in industrial packages containing UF6 on board of a Cargo crossing the Suez Canal near Port-Said City. The accident scenario and emergency response actions taken during the different phases of the accident are going to be presented and discussed. The proposed emergency response actions taken to face the accident are going to be also presented. The work presented had revealed the importance of public awareness will be needed for populations located in densely populated areas along Suez Canal bank inorder to react timely and effectively to avoid the toxic and radiological

Cognitive modeling and dynamic probabilistic simulation of operating crew response to complex system accidents. Part 4: IDAC causal model of operator problem-solving response

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.H.J. [Center for Risk and Reliability, University of Maryland, College Park, MD 20742 (United States) and Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)]. E-mail: yhc@umd.edu; Mosleh, A. [Center for Risk and Reliability, University of Maryland, College Park, MD 20742 (United States)

2007-08-15

This is the fourth in a series of five papers describing the Information, Decision, and Action in Crew context (IDAC) operator response model for human reliability analysis. An example application of this modeling technique is also discussed in this series. The model has been developed to probabilistically predicts the responses of a nuclear power plant control room operating crew in accident conditions. The operator response spectrum includes cognitive, emotional, and physical activities during the course of an accident. This paper assesses the effects of the performance-influencing factors (PIFs) affecting the operators' problem-solving responses including information pre-processing (I), diagnosis and decision making (D), and action execution (A). Literature support and justifications are provided for the assessment on the influences of PIFs.

Cognitive modeling and dynamic probabilistic simulation of operating crew response to complex system accidents. Part 4: IDAC causal model of operator problem-solving response

International Nuclear Information System (INIS)

Chang, Y.H.J.; Mosleh, A.

2007-01-01

This is the fourth in a series of five papers describing the Information, Decision, and Action in Crew context (IDAC) operator response model for human reliability analysis. An example application of this modeling technique is also discussed in this series. The model has been developed to probabilistically predicts the responses of a nuclear power plant control room operating crew in accident conditions. The operator response spectrum includes cognitive, emotional, and physical activities during the course of an accident. This paper assesses the effects of the performance-influencing factors (PIFs) affecting the operators' problem-solving responses including information pre-processing (I), diagnosis and decision making (D), and action execution (A). Literature support and justifications are provided for the assessment on the influences of PIFs

Nuclear accident response and population protection in The Netherlands: philosophy, solutions, and problems

International Nuclear Information System (INIS)

Baas, J.L.; Bosnjakovic, B.F.M.

1981-01-01

The subject is dealt with under the headings: nuclear energy in the Netherlands; the philosophy underlying the existing organization of response to nuclear accidents and population protection; practical experiences and recent developments; dose estimates; data transmission and lines of communication; international cooperation. (U.K.)

Planning the medical response to radiological accidents

International Nuclear Information System (INIS)

1998-01-01

Radioactive substances and other sources of ionizing radiation are used to assist in diagnosing and treating diseases, improving agricultural yields, producing electricity and expanding scientific knowledge. The application of sources of radiation is growing daily, and consequently the need to plan for radiological accidents is growing. While the risk of such accidents cannot be entirely eliminated, experience shows that most of the rare cases that have occurred could have been prevented, as they are often caused by human error. Recent radiological accidents such as those at Chernobyl (Ukraine 1986), Goiania (Brazil 1987), San Salvador (El Salvador 1989), Sor-Van (Israel 1990), Hanoi (Viet Nam 1992) and Tammiku (Estonia 1994) have demonstrated the importance of adequate preparation for dealing with such emergencies. Medical preparedness for radiological accidents must be considered an integral part of general emergency planning and preparedness and established within the national framework for radiation protection and safety. An IAEA Technical Committee meeting held in Istanbul in 1988 produced some initial guidance on the subject, which was subsequently developed, reviewed and updated by groups of consultants in 1989, 1992 and 1996. Special comments were provided by WHO, as co-sponsor of this publication, in 1997. This Safety Report outlines the roles and tasks of health authorities and hospital administrators in emergency preparedness for radiological accidents. Health authorities may use this document as the basis for their medical management in a radiological emergency, bearing in mind that adaptations will almost certainly be necessary to take into account the local conditions. This publication also provides information relevant to the integration of medical preparedness into emergency plans

Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

International Nuclear Information System (INIS)

Zhang Longfei; Zhang Dafa; Wang Shaoming

2007-01-01

Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

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

Occupational accidents aboard merchant ships

DEFF Research Database (Denmark)

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

2002-01-01

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

Analysis of an AP600 intermediate-size loss-of-coolant accident

Energy Technology Data Exchange (ETDEWEB)

Boyack, B.E.; Lime, J.F. [Los Alamos National Lab., NM (United States)

1995-09-01

A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

Accident dynamics of LR-0 reactor

International Nuclear Information System (INIS)

Vorisek, M.; Tinka, I.

1981-01-01

The results are given of calculating the accident dynamics of the LR-0 light water experimental zero power reactor. Calculations of the time dependence of power, the total released energy, the temperature of fuel and its cladding were made using program FATRAP for different values of the total inserted reactivity. Using the results, an analysis is made of hypothetic accident states of the LR-0 reactor. The results are shown graphically. (J.B.)

Underreporting of maritime accidents to vessel accident databases.

Science.gov (United States)

Hassel, Martin; Asbjørnslett, Bjørn Egil; Hole, Lars Petter

2011-11-01

Underreporting of maritime accidents is a problem not only for authorities trying to improve maritime safety through legislation, but also to risk management companies and other entities using maritime casualty statistics in risk and accident analysis. This study collected and compared casualty data from 01.01.2005 to 31.12.2009, from IHS Fairplay and the maritime authorities from a set of nations. The data was compared to find common records, and estimation of the true number of occurred accidents was performed using conditional probability given positive dependency between data sources, several variations of the capture-recapture method, calculation of best case scenario assuming perfect reporting, and scaling up a subset of casualty information from a marine insurance statistics database. The estimated upper limit reporting performance for the selected flag states ranged from 14% to 74%, while the corresponding estimated coverage of IHS Fairplay ranges from 4% to 62%. On average the study results document that the number of unreported accidents makes up roughly 50% of all occurred accidents. Even in a best case scenario, only a few flag states come close to perfect reporting (94%). The considerable scope of underreporting uncovered in the study, indicates that users of statistical vessel accident data should assume a certain degree of underreporting, and adjust their analyses accordingly. Whether to use correction factors, a safety margin, or rely on expert judgment, should be decided on a case by case basis. Copyright © 2011 Elsevier Ltd. All rights reserved.

Comparative analysis of a hypothetical coolant loss accident in an LMFB reactor with the use of various calculation models for a common reference problem

International Nuclear Information System (INIS)

Royl, P.

1979-01-01

The results of a comparative analysis of the initial and dismantling stages of a hypothetical loss of flow accident in an LMFB reactor are presented. The analyses were made for a common reference problem with four different calculation models (CARMEN/KADIS, SURDYN, CAPRI/KADIS and FRAX). The reference core is described specifically, as are the differences in its geometrical disposition in the models, the static and transient conditions before and after the start of boiling and during dismantling. The differences in the models used for simulating the boiling and the dismantling are compared. The structure of the core, as well as the calculation conditions and hypotheses, were intentionally designed so that the accident would culminate, in all cases, in an energetic hydrodynamic dismantling stage

Nuclear fuel cycle facility accident analysis handbook

International Nuclear Information System (INIS)

Ayer, J.E.; Clark, A.T.; Loysen, P.; Ballinger, M.Y.; Mishima, J.; Owczarski, P.C.; Gregory, W.S.; Nichols, B.D.

1988-05-01

The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH

Developing Off-site Emergency Preparedness and Response Model (OEPRM) for Severe Accident of NPP in a Densely Populated Country Using System Dynamics Approach

Energy Technology Data Exchange (ETDEWEB)

Hossena, Muhammed Mufazzal; Kang, Kyoung Ho; Song, Jin Ho [KAERI, Deajeon (Korea, Republic of)

2016-05-15

The main objectives of this study are to find the influencing factors of systems and sub-systems of OEPRM, to find the interdependency among the influencing factors, and to develop a conceptual qualitative OEPRM for densely populated NPP country in case of SA using system dynamics (SD). NPP accidents are classified as nuclear accidents and incidents depending on the severity. Severe accident (SA) is certain low probability accident that are beyond design basis accident which may arise due to multiple failures of safety systems leading to significant core degradation and jeopardize the integrity of many or all of the barriers to the release of radioactive material. The weakness to the off-site emergency response in the time of Fukushima accident was observed. So, it is crucial to develop an off-site emergency preparedness and responses model (OEPRM) for radiological emergency in densely populated country from the Fukushima emergency response lesson. In this study, an OEPRM is developed for densely populated NPP country to mitigate radiological effects in case of SA using SD approach. Besides, this study focuses the weakness of emergency response in Fukushima accident and proposed solution approach. The development of OEPRM in case of SA of NPP is very complex because of the involvement of various organization and it requires highly specialized agencies and technical experts. Moreover, if the country is agriculture based, it will make completely sophisticated.

Developing Off-site Emergency Preparedness and Response Model (OEPRM) for Severe Accident of NPP in a Densely Populated Country Using System Dynamics Approach

International Nuclear Information System (INIS)

Hossena, Muhammed Mufazzal; Kang, Kyoung Ho; Song, Jin Ho

2016-01-01

The main objectives of this study are to find the influencing factors of systems and sub-systems of OEPRM, to find the interdependency among the influencing factors, and to develop a conceptual qualitative OEPRM for densely populated NPP country in case of SA using system dynamics (SD). NPP accidents are classified as nuclear accidents and incidents depending on the severity. Severe accident (SA) is certain low probability accident that are beyond design basis accident which may arise due to multiple failures of safety systems leading to significant core degradation and jeopardize the integrity of many or all of the barriers to the release of radioactive material. The weakness to the off-site emergency response in the time of Fukushima accident was observed. So, it is crucial to develop an off-site emergency preparedness and responses model (OEPRM) for radiological emergency in densely populated country from the Fukushima emergency response lesson. In this study, an OEPRM is developed for densely populated NPP country to mitigate radiological effects in case of SA using SD approach. Besides, this study focuses the weakness of emergency response in Fukushima accident and proposed solution approach. The development of OEPRM in case of SA of NPP is very complex because of the involvement of various organization and it requires highly specialized agencies and technical experts. Moreover, if the country is agriculture based, it will make completely sophisticated

Analysis of emergency response after the Chernobyl accident in Belarus: observed and prevented medical consequences, lessons learned

International Nuclear Information System (INIS)

Buglova, E.; Kenigsberg, J.

1997-01-01

Belarus is one of the most contaminated Republic due to the Chernobyl accident. 23% of the entire area of Belarus was contaminated with radionuclides. To protect the population after the accident different types of protective actions were performed during all phases, based on various temporary dose limits. An analysis of conducted protective actions and lessons obtained during the emergency response is briefly presented

Research investigation report on Fukushima Daiichi nuclear accident

International Nuclear Information System (INIS)

2012-03-01

This report was issued in February 2012 by Rebuild Japan Initiative Foundation's Independent Investigation Commission on the Fukushima Daiichi Nuclear Accident, which consisted of six members from the private sector in independent positions and with no direct interest in the business of promoting nuclear power. Commission aimed to determine the truth behind the accident by clarifying the various problems and reveal systematic problems behind these issues so as to create a new starting point by identifying clear lessons learned. Report composed of four chapters; (1) progression of Fukushima accident and resulting damage (accident management after Fukushima accident, and effects and countermeasure of radioactive materials discharged into the environment), (2) response against Fukushima accident (emergency response of cabinet office against nuclear disaster, risk communication and on-site response against nuclear disaster), (3) analysis of historical and structural factors (technical philosophy of nuclear safety, problems of nuclear safety regulation of Fukushima accident, safety regulatory governance and social background of 'Safety Myth'), (4) Global Context (implication in nuclear security, Japan in nuclear safety regime, U.S.-Japan relations for response against Fukushima accident, lessons learned from Fukushima accident - aiming at creation of resilience). Report could identify causes of Fukushima accident and factors related to resulting damages, show the realities behind failure to prevent the spread of damage, and analyze the overall structural and historical background behind the accidents. (T. Tanaka)

State of reproductive system glands in males participated in the Chernobyl accident response

International Nuclear Information System (INIS)

Evdokimov, V.V.; Erasova, V.I.; Demin, A.I.; Dubinina, E.B.; Lyubchenko, P.N.

1993-01-01

State of reproductive system glands in males participated in the Chernobyl accident response and exposed to external irradiation at the dose up to 25 cGy was studied. 164 men at the age of 22-50 y.o. were examinated. Percentages of the various reproductive disorders were presented

Another 60Co hot cell accident

International Nuclear Information System (INIS)

Steidley, K.D.; Zeik, G.S.; Ouellette, R.

1979-01-01

An accident at a commercial irradiation facility is described. Because of inadequate safety arrangements a worker entered the irradiation cell while the 500,000 Ci 60 Co source was exposed. The duration of the exposure was established by re-enactment of the worker's movements, and dose rates at various locations at the accident site were measured. From the results a total absorbed dose of 210 rad to the surface of the victim was calculated. This and results of chromosomal aberration studies (200 rad) and the worker's film badge reading (126 rad) lead to the assignment of a nominal 200 rad absorbed dose. The worker was taken to two local hospital emergency departments, both of which examined and discharged him. Eventually he was admitted to a medical centre with experience of radiation effects. His symptoms and treatment during his 7 week stay in hospital are described and his blood counts vs time are shown graphically. As a result of the unsatisfactory response of the emergency departments the NRC has modified its licensing procedures to include specific arrangements with a local hospital. It is stressed that accident dosage estimates should always be supplied to the patient or his personal physician. (author)

Psychical and social effects related to post-accident situations: some training of Chernobyl accident

International Nuclear Information System (INIS)

Lochand, J.

1995-01-01

Some preliminary considerations on the psychic and societal dimensions related to post-accident situations connected to large scale and heavy land contamination are presented. This is done with the objective of exploring the role that these dimensions could play in the elaboration of new radiological protection principles and concepts in order to restore confidence among affected populations after a nuclear accident. It is important to facilitate the return to normal or, at least, acceptable living conditions, as soon as reasonably achievable, and to prevent the possible emergence of a post-accident crisis. A scheme is proposed for understanding the dynamics of the various phases after an accident, taking into account the collective response to the consequences as well as, the response to the countermeasures. (Author)

Reflections on the emergency preparations and responses of China to Fukushima nuclear accident in Japan

International Nuclear Information System (INIS)

Chen Xiaoqiu; Li Bing; Yu Shaoqing

2012-01-01

This paper reviewed the emergency response of Japan in Fukushima nuclear accident, provided and discussed the issues should be of concern on emergency preparedness and response in future: (1) modifying the existing emergency preparedness and response system; (2) consolidating the concept of emergency preparedness as the ultimate level of defense-in-depth; (3) promoting the emergency response decision-making support capabilities; (4) valuing the information opening of involving nuclear news and radiation environmental information. (authors)

Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET

Energy Technology Data Exchange (ETDEWEB)

Kliem, S

1998-10-01

Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)

Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET

International Nuclear Information System (INIS)

Kliem, S.

1998-01-01

Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)

Vaporization of structural materials in severe accidents

International Nuclear Information System (INIS)

Lorenz, R.A.

1982-01-01

Vaporized structural materials form the bulk of aerosol particles that can transport fission products in severe LWR accidents. As part of the Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory, a model has been developed based on a mass transport coefficient to describe the transport of materials from the surface of a molten pool. In many accident scenarios, the coefficient can be calculated from existing correlations for mass transfer by natural convection. Data from SASCHA fuel melting tests (Karlsruhe, Germany) show that the partial pressures of many of the melt components (Fe, Cr, Co, Mn, Sn) required for the model can be calculated from the vapor pressures of the pure species and Raoult's law. These calculations indicate much lower aerosol concentrations than reported in previous studies

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

Analysis of emergency response after the Chernobyl accident in Belarus: observed and prevented medical consequences, lessons learned

Energy Technology Data Exchange (ETDEWEB)

Buglova, E.; Kenigsberg, J. [Research Clinical Inst. of Radiation Medicine and Endocrinology, Minsk (Belarus)

1997-12-31

Belarus is one of the most contaminated Republic due to the Chernobyl accident. 23% of the entire area of Belarus was contaminated with radionuclides. To protect the population after the accident different types of protective actions were performed during all phases, based on various temporary dose limits. An analysis of conducted protective actions and lessons obtained during the emergency response is briefly presented 9 refs.

History of aerial surveys in response to radiological incidents and accidents

International Nuclear Information System (INIS)

Jobst, J.E.

1986-01-01

EG and G Energy Measurements Inc., operates the Remote Sensing Laboratory for the US Department of Energy (DOE). The Laboratory plays a key role in the federal response to a radiological incident or accident. It assists the DOE in the establishment of a Federal Radiological Monitoring and Assessment Center (FRMAC). The Remote Sensing Laboratory has played a major role in more than 13 incidents, including lost sources, accidental dispersions, and nuclear reactor incidents

Severe accident recriticality analyses (SARA)

Energy Technology Data Exchange (ETDEWEB)

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

2001-11-01

Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the following issues have been investigated in the SARA project: (1) the energy deposition in the fuel during super-prompt power burst; (2) the quasi steady-state reactor power following the initial power burst; and (3) containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality--both super-prompt power bursts and quasi steady-state power generation--for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45-2000 kg s{sup -1} injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g{sup -1}, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s{sup -1}. In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated

Severe accident recriticality analyses (SARA)

International Nuclear Information System (INIS)

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

2001-01-01

Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the following issues have been investigated in the SARA project: (1) the energy deposition in the fuel during super-prompt power burst; (2) the quasi steady-state reactor power following the initial power burst; and (3) containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality--both super-prompt power bursts and quasi steady-state power generation--for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45-2000 kg s -1 injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g -1 , was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s -1 . In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated quasi steady

Study of containment air cooler capacity in steam air environment during accident conditions

International Nuclear Information System (INIS)

Kansal, M.; Mohan, N.; Bhawal, R.N.; Bajaj, S.S.

2002-01-01

Full text: The air coolers are provided for controlling the temperature in the reactor building during normal operation. These air coolers also serve as the main heat sink for the removal of energy from high enthalpy air-steam mixture expected in reactor building under accident conditions. A subroutine COOLER has been developed to estimate the heat removal rate of the air coolers at high temperature and steam conditions. The subroutine COOLER has been attached with the code PACSR (post accident containment system response) used for containment pressure temperature calculation. The subroutine was validated using design parameters at normal operating condition. A study was done to estimate the heat removal rate for some postulated accident conditions. The study reveals that, under accident conditions, the heat removal rate of air coolers increases several times compared with normal operating conditions

Overdose problem associated with treatment planning software for high energy photons in response of Panama's accident.

Science.gov (United States)

Attalla, Ehab M; Lotayef, Mohamed M; Khalil, Ehab M; El-Hosiny, Hesham A; Nazmy, Mohamed S

2007-06-01

The purpose of this study was to quantify dose distribution errors by comparing actual dose measurements with the calculated values done by the software. To evaluate the outcome of radiation overexposure related to Panama's accident and in response to ensure that the treatment planning systems (T.P.S.) are being operated in accordance with the appropriate quality assurance programme, we studied the central axis and pripheral depth dose data using complex field shaped with blocks to quantify dose distribution errors. Multidata T.P.S. software versions 2.35 and 2.40 and Helax T.P.S. software version 5.1 B were assesed. The calculated data of the software treatment planning systems were verified by comparing these data with the actual dose measurements for open and blocked high energy photon fields (Co-60, 6MV & 18MV photons). Close calculated and measured results were obtained for the 2-D (Multidata) and 3-D treatment planning (TMS Helax). These results were correct within 1 to 2% for open fields and 0.5 to 2.5% for peripheral blocked fields. Discrepancies between calculated and measured data ranged between 13. to 36% along the central axis of complex blocked fields when normalisation point was selected at the Dmax, when the normalisation point was selected near or under the blocks, the variation between the calculated and the measured data was up to 500% difference. The present results emphasize the importance of the proper selection of the normalization point in the radiation field, as this facilitates detection of aberrant dose distribution (over exposure or under exposure).

EPR response characterization of drugs excipients for applying in accident dosimetry

International Nuclear Information System (INIS)

Marczewski, Barbara S.; Rodrigues Junior, Orlando; Galante, Ocimar L.; Costa, Zelia M. da; Campos, Leticia L.

2002-01-01

Some drugs are widely used by the population and can be employed to dose retrospective. The carbohydrates (saccharides), commonly used as excipients in the pharmaceutical industry, produce a quantity of free radicals after gamma irradiation, making them useful for dosimetry in emergency or accident situations that imply in dose evaluation from the materials found nearly or in contact with victims. In general, EPR signal from pulverized pills of some drugs are very complex due to the variety of components in the formulation. Because of this fact, some pharmaceutical excipients identified in the pill composition were also analysed by EPR spectrometry. On the counter drugs were studied: Cebion glucose, AAS, Aspirina, Conmel, Lacto-Purga and sugar substitutive ZeroCal. The excipients were: lactose, amide, anhydrous glucose and magnesium stearate. In some samples the number of radicals produced increased with the dose, showing a linear response for a dose range of interest and an adequate sensibility for dosimetry in accident cases

Modeling the consequences of hypothetical accidents for the Titan II system

International Nuclear Information System (INIS)

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

1981-11-01

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

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

Psychological response of accident

International Nuclear Information System (INIS)

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

1996-01-01

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

PWR accident management realated tests: some Bethsy results

International Nuclear Information System (INIS)

Clement, P.; Chataing, T.; Deruaz, R.

1993-01-01

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

Accident consequence calculations for project W-058 safety analysis

International Nuclear Information System (INIS)

Van Keuren, J.C.

1997-01-01

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

Importance of LWR best-estimate safety calculations for analysis of Fukushima-like accidents

International Nuclear Information System (INIS)

Sanchez Espinoza, V.; Ivanov, K.

2011-01-01

The safety assessment of nuclear power plants relies heavily on numerical simulations, which must include the most important physical models that are representative for the reactor type of interest. The current trends in nuclear power generation and regulation are to perform safety studies by 'best-estimate' codes that allow a realistic modeling of nuclear and thermal-hydraulic processes of the reactor core and the entire plant behavior including control and protection functions. Realistic methods are referred to as 'best-estimate' calculations, implying that they use a set of data, correlations, and methods designed to represent the phenomena, using the best available techniques. The application of best-estimate methodologies in the licensing process requires the quantification of the embedded uncertainties of the used codes. In this field many international initiatives are underway under the umbrella of the OECD such as the Light Water Reactor Uncertainty Analysis in Modeling benchmark, Oskarshamn 2 Boiling Water Reactor (BWR) Stability benchmark, Kalinin-3 VVER-1000 benchmark, etc. that underlies the importance of these issues. The Fukushima accident has shown the importance of the knowledge of the initial phase of the accident regarding the state of the core, in-vessel structures, and containment as well as the amount of fissile material inventories that potentially can be released if the safety barriers fail. For the development of mitigation and prevention measures modeling of the sequence of the events along with understanding of the key physical phenomena driving the accident progression is important. The paper presents the best-estimate coupled methodologies implemented, validated and applied at the Karlsruhe Institute Technology (KIT) for both types of LWRs - Pressurized Water Reactors (PWRs) and BWRs. Example are given with a BWR steady state and transient simulations along with corresponding uncertainty quantification. The on-going development of high

Assessment of accident risks in the CRBRP. Volume 2. Appendices

Energy Technology Data Exchange (ETDEWEB)

None

1977-03-01

Appendices to Volume I include core-related accident-sequence definition, CRBRP risk-assessment sequence-probability determinations, failure-probability data, accident scenario evaluation, radioactive material release analysis, ex-core accident analysis, safety philosophy and design features, calculation of reactor accident consequences, sensitivity study, and risk from fires.

A review on contamination and emergency response actions in USSR, European countries and Japan in the reactor accident at Chernobyl

International Nuclear Information System (INIS)

Akaishi, Jun; Ohhata, Tsutomu

1987-01-01

The accident occurred in the Chernobyl No.4 reactor on April 26, 1986 became the largest accident of nuclear reactors so far as the quantity of release of radioactive substances was very much, and the wide area contamination over the whole Northern Hemisphere was brought about. Consequently, the revision over wide ranges, such as the desigh and operation of nuclear reactors and the response at the time of emergency, was to be urgently demanded. As to this accident, many reports have been already made. In this report, the contamination and the response carried out at the time of emergency including the state in Japan are mainly described. The contents of this report are based on the data sent from the international organizations, the results of measurement by the organizations in respective countries, the publication by USSR at the IAEA expert conference in August, 1986, and the data published in respective countries by research institutes. The state in respective countries after the accident, the radioactivity contamination due to the accident in USSR, European countries and Japan, the diffusion and move of radioactive substances, the countermeasures to the accident such as the estimation of exposure dose, scientific investigation, the evacuation of people and so on are reported. (Kako, I.)

CONTAIN calculations; CONTAIN-Rechnungen

Energy Technology Data Exchange (ETDEWEB)

Scholtyssek, W.

1995-08-01

In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident `medium-sized leak in the cold leg`, especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)

Review of specific radiological accident considerations

International Nuclear Information System (INIS)

Elder, J.

1984-01-01

Specific points of guidance provided in the forthcoming document A Guide to Radiological Accident Considerations for Siting and Design of Nonreactor Nuclear Facilities are discussed. Of these, the following are considered of particular interest to analysts of hypothetical accidents: onsite dose limits; population dose, public health effects, and environmental contamination as accident consequences which should be addressed; risk analysis; natural phenomena as accident initiators; recommended dose models; multiple organ equivalent dose; and recommended methods and parameters for source terms and release amount calculations. Comments are being invited on this document, which is undergoing rewrite after the first stage of peer review

[Accidents on Iceland's most dangerous roads].

Science.gov (United States)

Bjarnason, Thóroddur; Arnarsson, Sveinn

2012-02-01

The objective of this paper was to identify the most dangerous segments of the Icelandic road system in terms of the number of accidents pr km and the rate of accidents pr million km travelled. First to identify the segments where the number of accidents is highest and where the risk of the individual traveller is the greatest. Second to evaluate if the association between the number and the rate of accidents is positive or negative. Third to identify the road segments that are the most dangerous in the sense of many accidents and great risk to individual travellers. Main roads outside urban centers were divided into 45 segments that were on average 78 km in length. Infrequently travelled roads and roads within urban centers were omitted. Information on the length of roads, traffic density and number of accidents was used to calculate the number of accidents per km and the rate of accidents per million km travelled. The correlation between the number and rate of accidents was calculated and the most dangerous road segments were identified by the average rank order on both dimensions. Most accidents pr km occurred on the main roads to and from the capital region, but also east towards Hvolsvöllur, north towards Akureyri and in the Mideast region of the country. The rate of accidents pr million km travelled was highest in the northeast region, in northern Snæfellsnes and in the Westfjords. The most dangerous roads on both dimensions were in Mideast, northern Westfjords, in the north between Blönduós and Akureyri and in northern Snæfellsnes. Most accidents pr km occurred on roads with a low accident rate pr million km travelled. It is therefore possible to reduce accidents the most by increasing road safety where it is already the greatest but that would however increase inequalities in road safety. Policy development in transportation is therefore in part a question of priorities in healthcare. Individual equality in safety and health are not always fully

NPP Krsko containment environmental conditions during postulated accident

International Nuclear Information System (INIS)

Kozaric, M.; Cavlina, N.; Spalj, S.

1989-01-01

This paper presents NPP Krsko containment pressure and temperature increase during Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB). Containment environmental condition calculation was performed by CONTEMPT4/MOD4 computer code. Design accident calculations were performed by RELAP4/MOD6 and RELAP5/MOD1 computer codes. Calculational abilities and application methodology of these codes are presented. The CONTEMPT code is described in more detail. The containment pressure and temperature time distribution are presented as well. (author)

Bulgarian Emergency Response System (BERS) in case of nuclear accident with exposure doses estimation

Energy Technology Data Exchange (ETDEWEB)

Syrakov, D.; Prodanova, M.; Slavov, K.; Veleva, B.

2015-07-01

A PC-oriented Emergency Response System in case of nuclear accident (BERS) is developed and works operationally in the National Institute of Meteorology and Hydrology (NIMH). The creation and development of BERS was highly stimulated by the ETEX (European Tracer Experiment) project. BERS comprises two main parts - the operational and the accidental ones. The operational part, run automatically every 12 hours, prepares the input meteorological file used by both trajectory and dispersion models, runs the trajectory models, visualizes the results and uploads the maps of trajectories to a dedicated web-site. The accidental part is activated manually when a real radioactive releases occur or during emergency exercises. Its core is the Bulgarian dispersion models EMAP. Outputs are concentration, accumulated deposition and selected doses fields. In the paper, the BERS overall structure is described and examples of its products are presented. Key words: nuclear accident, emergency response, early warning system, air dispersion models, radioactive exposure dose. (Author)

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

International Nuclear Information System (INIS)

Megahed, M.M.

1992-01-01

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

Accident analysis of railway transportation of low-level radioactive and hazardous chemical wastes: Application of the /open quotes/Maximum Credible Accident/close quotes/ concept

Energy Technology Data Exchange (ETDEWEB)

Ricci, E.; McLean, R.B.

1988-09-01

The maximum credible accident (MCA) approach to accident analysis places an upper bound on the potential adverse effects of a proposed action by using conservative but simplifying assumptions. It is often used when data are lacking to support a more realistic scenario or when MCA calculations result in acceptable consequences. The MCA approach can also be combined with realistic scenarios to assess potential adverse effects. This report presents a guide for the preparation of transportation accident analyses based on the use of the MCA concept. Rail transportation of contaminated wastes is used as an example. The example is the analysis of the environmental impact of the potential derailment of a train transporting a large shipment of wastes. The shipment is assumed to be contaminated with polychlorinated biphenyls and low-level radioactivities of uranium and technetium. The train is assumed to plunge into a river used as a source of drinking water. The conclusions from the example accident analysis are based on the calculation of the number of foreseeable premature cancer deaths the might result as a consequence of this accident. These calculations are presented, and the reference material forming the basis for all assumptions and calculations is also provided.

Accident analysis of railway transportation of low-level radioactive and hazardous chemical wastes: Application of the /open quotes/Maximum Credible Accident/close quotes/ concept

International Nuclear Information System (INIS)

Ricci, E.; McLean, R.B.

1988-09-01

The maximum credible accident (MCA) approach to accident analysis places an upper bound on the potential adverse effects of a proposed action by using conservative but simplifying assumptions. It is often used when data are lacking to support a more realistic scenario or when MCA calculations result in acceptable consequences. The MCA approach can also be combined with realistic scenarios to assess potential adverse effects. This report presents a guide for the preparation of transportation accident analyses based on the use of the MCA concept. Rail transportation of contaminated wastes is used as an example. The example is the analysis of the environmental impact of the potential derailment of a train transporting a large shipment of wastes. The shipment is assumed to be contaminated with polychlorinated biphenyls and low-level radioactivities of uranium and technetium. The train is assumed to plunge into a river used as a source of drinking water. The conclusions from the example accident analysis are based on the calculation of the number of foreseeable premature cancer deaths the might result as a consequence of this accident. These calculations are presented, and the reference material forming the basis for all assumptions and calculations is also provided

Site response calculations for nuclear power plants

International Nuclear Information System (INIS)

Wight, L.H.

1975-01-01

Six typical sites consisting of three soil profiles with average shear wave velocities of 800, 1800, and 5000 ft/sec as well as two soil depths of 200 and 400 ft were considered. Seismic input to these sites was a synthetic accelerogram applied at the surface and corresponding to a statistically representative response spectrum. The response of each of these six sites to this input was calculated with the SHAKE program. The results of these calculations are presented

Corporate Cost of Occupational Accidents

DEFF Research Database (Denmark)

Rikhardsson, Pall M.; Impgaard, M.

2004-01-01

method could be used in all of the companies without revisions. The evaluation of accident cost showed that 2/3 of the costs of occupational accidents are visible in the Danish corporate accounting systems reviewed while 1/3 is hidden from management view. The highest cost of occupational accidents......The systematic accident cost analysis (SACA) project was carried out during 2001 by The Aarhus School of Business and PricewaterhouseCoopers Denmark with financial support from The Danish National Working Environment Authority. Its focused on developing and testing a method for evaluating...... occupational costs of companies for use by occupational health and safety professionals. The method was tested in nine Danish companies within three different industry sectors and the costs of 27 selected occupational accidents in these companies were calculated. One of the main conclusions is that the SACA...

Occupational accidents among mototaxi drivers.

Science.gov (United States)

Amorim, Camila Rego; de Araújo, Edna Maria; de Araújo, Tânia Maria; de Oliveira, Nelson Fernandes

2012-03-01

The use of motorcycles as a means of work has contributed to the increase in traffic accidents, in particular, mototaxi accidents. The aim of this study was to estimate and characterize the incidence of occupational accidents among the mototaxis registered in Feira de Santana, BA. This is a cross-sectional study with descriptive and census data. Of the 300 professionals registered at the Municipal Transportation Service, 267 professionals were interviewed through a structured questionnaire. Then, a descriptive analysis was conducted and the incidence of accidents was estimated based on the variables studied. Relative risks were calculated and statistical significance was determined using the chi-square test and Fisher's exact test, considering p accidents were observed in 10.5% of mototaxis. There were mainly minor injuries (48.7%), 27% of them requiring leaves of absence from work. There was an association between the days of work per week, fatigue in lower limbs and musculoskeletal complaints, and accidents. Knowledge of the working conditions and accidents involved in this activity can be of great importance for the adoption of traffic education policies, and to help prevent accidents by improving the working conditions and lives of these professionals.

Response of an ESBWR to an accident like Fukushima

International Nuclear Information System (INIS)

Pena G, C. A.; Salazar S, E.

2014-10-01

The Fukushima nuclear accident occurred on March 11, 2011 was caused by an earthquake measuring 9.0 on the seismic scale, damaging the nuclear power plant (NPP) containing 6 nuclear reactors of boiling water, some of whom suffered a meltdown due to failure in the cooling systems of these. Electrical wiring was damaged and emergency diesel generators stopped with the arrival of the tsunami, caused by the earthquake. The presence of many critical systems in flood areas provided that a cascade of technological failures occurred, culminating in the complete loss of control over the central and reactors. This had important political consequences and several countries have revised their nuclear programs, considering dangerous and unworkable to the nuclear energy to meet the specific energy needs of each country. However, progress in the design of new reactors raise the possibility of continuing to use nuclear energy for much safer than before. The NPP of Fukushima contained BWR reactors of the first commercial generation designed in the 50. The new passive safety systems raise the idea of simplifying the systems, lowering costs while they are safer systems. New designs meet a these needs in both BWRs and PWRs. In this paper the response of an ESBWR reactor to similar conditions to those suffered by the reactors of Fukushima NPP during the accident is analyzed. In this new reactor of III + generation, the recirculation pumps are removed and emergency cooling systems that can maintain the stable core for several hours without electricity are used. To analyze the responses of this new reactor the simulator with basic principles that IAEA makes available to from Member States was used to better understand and analyze the performance of this type of reactors. (Author)

Emotional reactivity: Beware its involvement in traffic accidents.

Science.gov (United States)

M'bailara, Katia; Atzeni, Thierry; Contrand, Benjamin; Derguy, Cyrielle; Bouvard, Manuel-Pierre; Lagarde, Emmanuel; Galéra, Cédric

2018-04-01

Reducing risk attributable to traffic accidents is a public health challenge. Research into risk factors in the area is now moving towards identification of the psychological factors involved, particularly emotional states. The aim of this study was to evaluate the link between emotional reactivity and responsibility in road traffic accidents. We hypothesized that the more one's emotional reactivity is disturbed, the greater the likelihood of being responsible for a traffic accident. This case-control study was based on a sample of 955 drivers injured in a motor vehicle crash. Responsibility levels were determined with a standardized method adapted from the quantitative Robertson and Drummer crash responsibility instrument. Emotional reactivity was assessed with the MATHYS. Hierarchical cluster analysis discriminated four distinctive driver's emotional reactivity profiles: basic emotional reactivity (54%), mild emotional hyper-reactivity (29%), emotional hyper-reactivity (11%) and emotional hypo-reactivity (6%). Drivers who demonstrated emotional hypo-reactivity had a 2.3-fold greater risk of being responsible for a traffic accident than those with basic emotional reactivity. Drivers' responsibility in traffic accidents depends on their emotional status. The latter can change the ability of drivers, modifying their behavior and thus increasing their propensity to exhibit risk behavior and to cause traffic accidents. Copyright © 2017 Elsevier B.V. All rights reserved.

Calculation of Excore Detector Responses upon Control Rods Movement in PGSFR

Energy Technology Data Exchange (ETDEWEB)

Ha, Pham Nhu Viet; Lee, Min Jae; Kang, Chang Moo; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) safety design concept, which aims at achieving IAEA's safety objectives and GIF's safety goals for Generation-IV reactor systems, is mainly focused on the defense in depth for accident detection, prevention, control, mitigation and termination. In practice, excore neutron detectors are widely used to determine the spatial power distribution and power level in a nuclear reactor core. Based on the excore detector signals, the reactor control and protection systems infer the corresponding core power and then provide appropriate actions for safe and reliable reactor operation. To this end, robust reactor power monitoring, control and core protection systems are indispensable to prevent accidents and reduce its detrimental effect should one occur. To design such power monitoring and control systems, numerical investigation of excore neutron detector responses upon various changes in the core power level/distribution and reactor conditions is required in advance. In this study, numerical analysis of excore neutron detector responses (DRs) upon control rods (CRs) movement in PGSFR was carried out. The objective is to examine the sensitivity of excore neutron detectors to the core power change induced by moving CRs and thereby recommend appropriate locations to locate excore neutron detectors for the designing process of the PGSFR power monitoring systems. Section 2 describes the PGSFR core model and calculation method as well as the numerical results for the excore detector spatial weighting functions, core power changes and detector responses upon various scenarios of moving CRs in PGSFR. The top detector is conservatively safe because it overestimated the core power level. However, the lower and bottom detectors still functioned well in this case because they exhibited a minor underestimation of core power of less than ∼0.5%. As a secondary CR was dropped into the core, the lower detector was

Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

[Accidents and injuries at work].

Science.gov (United States)

Standke, W

2014-06-01

In the case of an accident at work, the person concerned is insured by law according to the guidelines of the Sozialgesetzbuch VII as far as the injuries have been caused by this accident. The most important source of information on the incident in question is the accident report that has to be sent to the responsible institution for statutory accident insurance and prevention by the employer, if the accident of the injured person is fatal or leads to an incapacity to work for more than 3 days (= reportable accident). Data concerning accidents like these are sent to the Deutsche Gesetzliche Unfallversicherung (DGUV) as part of a random sample survey by the institutions for statutory accident insurance and prevention and are analyzed statistically. Thus the key issues of accidents can be established and used for effective prevention. Although the success of effective accident prevention is undisputed, there were still 919,025 occupational accidents in 2011, with clear gender-related differences. Most occupational accidents involve the upper and lower extremities. Accidents are analyzed comprehensively and the results are published and made available to all interested parties in an effort to improve public awareness of possible accidents. Apart from reportable accidents, data on the new occupational accident pensions are also gathered and analyzed statistically. Thus, additional information is gained on accidents with extremely serious consequences and partly permanent injuries for the accident victims.

Pilot program: NRC severe reactor accident incident response training manual: Public protective actions: Predetermined criteria and initial actions

International Nuclear Information System (INIS)

Martin, J.A. Jr.; McKenna, T.J.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Giitter, J.G.; Watkins, R.M.

1987-02-01

This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. Public Protective Actions - Predetermined Criteria and Initial Actions is the fourth in a series of volumes that collectively summarize the US Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume reviews public protective action criteria and objectives, their bases and implementation, and the expected public response. Each volume serves, respectively, as the text for a course of instruction in a series of courses for NRC response personnel. These materials do not provide guidance or license requirements for NRC licensees. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are called out in the text

Multicomponent risk inherent to nuclear or radiation accidents

International Nuclear Information System (INIS)

Poyarkov, V.A.

1997-01-01

A nuclear or radiation emergency response planning is based on expected avertable doses for a short (4 hours, 2 days, 1 week) and a long (50 or 70 years) periods of time. On calculating the doses one should take into account not only the sources of ionizing radiation with their foreseen characteristics, e.g. a half-life, but also the possibility of their probabilistic transformations. It can be illustrated by the chronicle of Chernobyl accident, where for nine days it was impossible to envisage with reliability the quantities of released radionuclides due to many uncontrolled factors related to the scope and character of the zone wreckage. Any other developments of the accident, such as destruction of the under reactor support plate, sinking of the melted zone into a highly radioactive water down at the reactor cavity or changing of the wind could have resulted in a substantial contamination of 3 mln. Kiev and other densely populated area of Ukraine. Another factor to be taken into account in the emergency response is a way of an actual risk perception by the population. The article suggests an approach to take into account the time-dependent secondary sources of exposure and the disparity in accepting the hazard of real risk of an accident by the trained workers and population. (author)

A decision support system for emergency response to major nuclear accidents

International Nuclear Information System (INIS)

Papazoglou, I.A.; Christou, M.D.

1997-01-01

A methodology for the optimization of the short-term emergency response in the event of a nuclear accident is presented. The method seeks an optimum combination of protective actions in the presence of a multitude of conflicting objectives and under uncertainty. Conflicting objectives arise in the attempt to minimize simultaneously the potential adverse effects of an accident and the associated socioeconomic impacts. Additional conflicting objectives arise whenever an emergency plan tends to decrease a particular health effect, such as acute deaths, while it increases another, such as latent deaths. The uncertainty is due to the multitude of possible accident scenarios and their respective probability of occurrence, the stochastic variability in the weather conditions, and the variability and/or lack of knowledge of the parameters of the risk assessment models. A multiobjective optimization approach is adopted. An emergency protection plan consists of defining a protective action at each spatial cell around the plant. Three criteria are used as the objective functions of the problem, namely, acute fatalities, latent effects, and socioeconomic cost. The optimization procedure defines the efficient frontier, i.e., all emergency plans that are not dominated by another in all three criteria. No value trade-offs are necessary up to this point. The most preferred emergency plan is then chosen among the set of efficient plans. Finally, the methodology is integrated into a computerized decision support system, and its use is demonstrated in a realistic application

Analysis of accidents at the LPR (Radiochemical Processes Laboratory)

International Nuclear Information System (INIS)

Kaufmann, F.; Boutet, L.I.

1987-01-01

Accidents are defined as not planned events that may result in the emission of significative quantities of radioactive materials to the environment. The pilot plant has been specifically designed to prevent this type of accidents but there still exists the possibility that one or more accidents can be produced during the plant life. In a first phase, the emission of radionuclides to the environment were evaluated for 13 credible accidents. In a second phase, by means of the calculation program SEDA, specially adapted to this purpose, the critical doses of critical group were calculated for each accident. Due to the small capacity of the pilot plant and the long cooling period of treated fuel, it is concluded that the radiological consequences for the external environment are of very small magnitude. In this way, without need of developing complex fault- or event-trees, it is shown that any of the accidents falls into the non acceptable zone of Farmer diagram. (Author)

Severe Accident Recriticality Analyses (SARA)

Energy Technology Data Exchange (ETDEWEB)

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

1999-11-01

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

Severe Accident Recriticality Analyses (SARA)

International Nuclear Information System (INIS)

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

1999-11-01

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

Estimating cost ratio distribution between fatal and non-fatal road accidents in Malaysia

Science.gov (United States)

Hamdan, Nurhidayah; Daud, Noorizam

2014-07-01

Road traffic crashes are a global major problem, and should be treated as a shared responsibility. In Malaysia, road accident tragedies kill 6,917 people and injure or disable 17,522 people in year 2012, and government spent about RM9.3 billion in 2009 which cost the nation approximately 1 to 2 percent loss of gross domestic product (GDP) reported annually. The current cost ratio for fatal and non-fatal accident used by Ministry of Works Malaysia simply based on arbitrary value of 6:4 or equivalent 1.5:1 depends on the fact that there are six factors involved in the calculation accident cost for fatal accident while four factors for non-fatal accident. The simple indication used by the authority to calculate the cost ratio is doubted since there is lack of mathematical and conceptual evidence to explain how this ratio is determined. The main aim of this study is to determine the new accident cost ratio for fatal and non-fatal accident in Malaysia based on quantitative statistical approach. The cost ratio distributions will be estimated based on Weibull distribution. Due to the unavailability of official accident cost data, insurance claim data both for fatal and non-fatal accident have been used as proxy information for the actual accident cost. There are two types of parameter estimates used in this study, which are maximum likelihood (MLE) and robust estimation. The findings of this study reveal that accident cost ratio for fatal and non-fatal claim when using MLE is 1.33, while, for robust estimates, the cost ratio is slightly higher which is 1.51. This study will help the authority to determine a more accurate cost ratio between fatal and non-fatal accident as compared to the official ratio set by the government, since cost ratio is an important element to be used as a weightage in modeling road accident related data. Therefore, this study provides some guidance tips to revise the insurance claim set by the Malaysia road authority, hence the appropriate method

Overdose Problem Associated with Treatment Planning Software for High Energy Photons in Response of Panama's Accident

International Nuclear Information System (INIS)

Attalla, E.M.; Lotayef, M.M.; Khalil, E.M.; El-Hosiny, H.A.

2007-01-01

The purpose of this study was to quantify dose distribution errors by comparing actual dose measurements with the calculated values done by the software. To evaluate the outcome of radiation overexposure related to Panama's accident and in response to ensure that the treatment planning systems (T.P.S.) are being operated in accordance with the appropriate quality assurance programme, we studied the central axis and pripheral depth dose data using complex field shaped with blocks to quantify dose distribution errors. Material and Methods: Multi data T.P.S. software versions 2.35 and 2.40 and Helax T.P.S. software version 5.1 B were assesed. The calculated data of the software treatment planning systems were verified by comparing these data with the actual dose measurements for open and blocked high energy photon fields (Co-60, 6MV and 18MV photons). Results: Close calculated and measured results were obtained for the 2-D (Multi data) and 3-D treatment planning (TMS Helax). These results were correct within 1 to 2% for open fields and 0.5 to 2.5% for peripheral blocked fields. Discrepancies between calculated and measured data ranged between 13. to 36% along the central axis of complex blocked fields when normalisation point was selected at the Dmax, when the normalisation point was selected near or under the blocks, the variation between the calculated and the measured data was up to 500% difference. Conclusions: The present results emphasize the importance of the proper selection of the normalization point in the radiation field, as this facilitates detection of aberrant dose distribution (over exposure or under exposure)

Analysis of hypothetical LMFBR whole-core accidents in the USA

International Nuclear Information System (INIS)

Ferguson, D.R.; Deitrich, L.W.; Brown, N.W.; Waltar, A.E.

1978-01-01

Methods used for analysis of material behaviour, accident phenomenology and integrated accident calculations are reviewed. Applications of these methods to hypothetical LOF and TOP accidents are discussed. Recent results obtained from applications to FFTF and CRBRP are presented. (author)

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

International Nuclear Information System (INIS)

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

1987-02-01

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

Build-up forces at Military Institute of Medical Radiology and Oncology for emergency medical response to some eventualities of radiological accidents - some suggestions

International Nuclear Information System (INIS)

Ho Van Cu; Nguyen Huu Nghia

2011-01-01

Nowadays, the use of various nuclear sources in some fields of the life has brought many practical advantages in general; especially in the next several years, our country will begin construction of the first nuclear plant. However, if there were user carelessness or objective disadvantageous factors (earthquake, tsunami, etc.), that disadvantages could lead to a radiation accident or nuclear accident which causes damages not only for economy but also for public health. Therefore, the emergency response to radiation accident, especially the emergency medical response that has a great important position. To satisfy this real demand, in 1996, Vietnam Ministry of Defence made the decision to establish Center for Nuclear Medicine and Radiation Protecting (now becomes Military Institute of Medical Radiology and Oncology) with the main missions are research, applying radiation protecting methods and organizing treatments to radiation injured victims. To fulfill above main missions, with the help of Vietnam Atomic Energy Institute (VAEI), Vietnam Agency for Radiation and Nuclear Safety (VARANS), the doctors and staffs of our Institute have been participated in the international training courses and workshops that organized in Vietnam or in regional countries about emergency medical response to radiation accidents, they get valuable information, knowledge and documents from these courses and workshops. Depending on the principles of radiation emergency medical response to nuclear/ radiation accidents that International Atomic Energy Agency (IAEA) guided, and with the experience learned from other countries in Asia region, our Institute have been gradually improving on organization and curing processes for the radiation victims and also setting the preparedness for emergency medical response to radiation accidents if maybe they could occur. (author)

Phenomenological analyses and their application to the Defense Waste Processing Facility probabilistic safety analysis accident progression event tree. Revision 1

International Nuclear Information System (INIS)

Kalinich, D.A.; Thomas, J.K.; Gough, S.T.; Bailey, R.T.; Kearnaghan, D.P.

1995-01-01

In the Defense Waste Processing Facility (DWPF) Safety Analysis Reports (SARs) for the Savannah River Site (SRS), risk-based perspectives have been included per US Department of Energy (DOE) Order 5480.23. The NUREG-1150 Level 2/3 Probabilistic Risk Assessment (PRA) methodology was selected as the basis for calculating facility risk. The backbone of this methodology is the generation of an Accident Progression Event Tree (APET), which is solved using the EVNTRE computer code. To support the development of the DWPF APET, deterministic modeling of accident phenomena was necessary. From these analyses, (1) accident progressions were identified for inclusion into the APET; (2) branch point probabilities and any attendant parameters were quantified; and (3) the radionuclide releases to the environment from accidents were determined. The phenomena of interest for accident progressions included explosions, fires, a molten glass spill, and the response of the facility confinement system during such challenges. A variety of methodologies, from hand calculations to large system-model codes, were used in the evaluation of these phenomena

Regulatory perspective on accident management issues

International Nuclear Information System (INIS)

Barrett, R.J.

1988-01-01

Effective response to reactor accidents requires a combination of emergency operations, technical support and emergency response. The NRC and industry have actively pursued programs to assure the adequacy of emergency operations and emergency response. These programs will continue to receive high priority. By contrast, the technical support function has received relatively little attention from NRC and the industry. The results from numerous PRA studies and the severe accident programs of NRC and the industry have yielded a wealth of insights on prevention and mitigation of severe accidents. The NRC intends to work with the industry to make these insights available to the technical support staffs through a combination of guidance, training and periodic drills

Accident source terms for boiling water reactors with high burnup cores.

Energy Technology Data Exchange (ETDEWEB)

Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

2007-11-01

The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

Impact of reducing sodium void worth on the severe accident response of metallic-fueled sodium-cooled reactors

International Nuclear Information System (INIS)

Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

1994-01-01

Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept

CARNSORE: Hypothetical reactor accident study

International Nuclear Information System (INIS)

Walmod-Larsen, O.; Jensen, N.O.; Kristensen, L.; Meide, A.; Nedergaard, K.L.; Nielsen, F.; Lundtang Petersen, E.; Petersen, T.; Thykier-Nielsen, S.

1984-06-01

Two types of design-basis accident and a series of hypothetical core-melt accidents to a 600 MWe reactor are described and their consequences assessed. The PLUCON 2 model was used to calculate the consequences which are presented in terms of individual and collective doses, as well as early and late health consequences. The site proposed for the nucelar power station is Carnsore Point, County Wexford, south-east Ireland. The release fractions for the accidents described are those given in WASH-1400. The analyses are based on the resident population as given in the 1979 census and on 20 years of data from the meteorological stations at Rosslare Harbour, 8.5 km north of the site. The consequences of one of the hypothetical core-melt accidents are described in detail in a meteorological parametric study. Likewise the consequences of the worst conceivable combination of situations are described. Finally, the release fraction in one accident is varied and the consequences of a proposed, more probable ''Class 9 accident'' are presented. (author)

Radionuclide release calculations for selected severe accident scenarios. PWR, ice condenser design

Energy Technology Data Exchange (ETDEWEB)

Denning, R S; Gieseke, J A; Cybulskis, P; Lee, K W; Jordan, H; Curtis, L A; Kelly, R F; Kogan, V; Schumacher, P M

1986-07-01

This report presents results of analyses of the environmental releases of fission products (source terms) for severe accident scenarios in a pressurized water reactor with an ice-condenser containment. The analyses were performed to support the Severe Accident Risk Reduction/Risk Rebaselining Program (SARRP) which is being undertaken for the U.S. Nuclear Regulatory Commission by Sandia National Laboratories. In the SARRP program, risk estimates are being generated for a number of reference plant designs. The Sequoyah Plant has been used in this study as an example of a PWR ice-condenser plant. (author)

Dose calculation for atmospheric releases from a nuclear accident using RAMS/HYPACT

International Nuclear Information System (INIS)

Tamura, Junji; Tomita, Kenichi; Homma, Toshimitsu

2004-01-01

This paper describes the investigation of uncertainties in the structure of the atmospheric dispersion/deposition model used in the probabilistic accident consequence assessment code, OSCAAR. To investigate these uncertainties, we have introduced the more sophisticated computer codes, RAMS and HYPACT, which were widely used in the research field of atmospheric phenomena. In this work, the capabilities of the HYPACT model were extended for use in accident consequence assessments. The preliminary comparison between the predictions by OSCAAR and those by RAMS/HYPACT were conducted for both individual and collective consequences in terms of probabilistic results. (author)

Application of Latin hypercube sampling to RADTRAN 4 truck accident risk sensitivity analysis

International Nuclear Information System (INIS)

Mills, G.S.; Neuhauser, K.S.; Kanipe, F.L.

1994-01-01

The sensitivity of calculated dose estimates to various RADTRAN 4 inputs is an available output for incident-free analysis because the defining equations are linear and sensitivity to each variable can be calculated in closed mathematical form. However, the necessary linearity is not characteristic of the equations used in calculation of accident dose risk, making a similar tabulation of sensitivity for RADTRAN 4 accident analysis impossible. Therefore, a study of sensitivity of accident risk results to variation of input parameters was performed using representative routes, isotopic inventories, and packagings. It was determined that, of the approximately two dozen RADTRAN 4 input parameters pertinent to accident analysis, only a subset of five or six has significant influence on typical analyses or is subject to random uncertainties. These five or six variables were selected as candidates for Latin Hypercube Sampling applications. To make the effect of input uncertainties on calculated accident risk more explicit, distributions and limits were determined for two variables which had approximately proportional effects on calculated doses: Pasquill Category probability (PSPROB) and link population density (LPOPD). These distributions and limits were used as input parameters to Sandia's Latin Hypercube Sampling code to generate 50 sets of RADTRAN 4 input parameters used together with point estimates of other necessary inputs to calculate 50 observations of estimated accident dose risk.Tabulations of the RADTRAN 4 accident risk input variables and their influence on output plus illustrative examples of the LHS calculations, for truck transport situations that are typical of past experience, will be presented

Accident sequence precursor analysis level 2/3 model development

International Nuclear Information System (INIS)

Lui, C.H.; Galyean, W.J.; Brownson, D.A.

1997-01-01

The US Nuclear Regulatory Commission's Accident Sequence Precursor (ASP) program currently uses simple Level 1 models to assess the conditional core damage probability for operational events occurring in commercial nuclear power plants (NPP). Since not all accident sequences leading to core damage will result in the same radiological consequences, it is necessary to develop simple Level 2/3 models that can be used to analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude of the resulting radioactive releases to the environment, and calculate the consequences associated with these releases. The simple Level 2/3 model development work was initiated in 1995, and several prototype models have been completed. Once developed, these simple Level 2/3 models are linked to the simple Level 1 models to provide risk perspectives for operational events. This paper describes the methods implemented for the development of these simple Level 2/3 ASP models, and the linkage process to the existing Level 1 models

Nuclear law and radiological accidents

International Nuclear Information System (INIS)

Frois, F.

1998-01-01

Nuclear activities in Brazil, and particularly the radiological accident of Goiania, are examined in the light of the environmental and nuclear laws of Brazil and the issue of responsibility. The absence of legislation covering radioactive wastes as well as the restrictions on Brazilian States to issue regulations covering nuclear activities are reviewed. The radiological accident and its consequences, including the protection and compensation of the victims, the responsibility of the shareholders of the Instituto Goiano de Radioterapia, operator of the radioactive source, the provisional storage and the final disposal at Abadia de Goias of the radioactive waste generated by the accident are reviewed. Finally, nuclear responsibility, the inapplicability of the Law 6453/77 which deals with nuclear damages, and the state liability regime are analysed in accordance with the principles of the Brazilian Federal Constitution. (author)

QNPP containment response to MSLB accident

International Nuclear Information System (INIS)

Zhou Jie

1989-01-01

The containment behavior of nuclear power plant under accident conditions plays an important role in nuclear safety analysis. The analysis and parametric studies are performed and some conclusions are obtained for QNPP (Qinshen Nuclear Power Plant) as reported in the paper. The CONTEMPT-LT/028 code is used

Accident sequence quantification with KIRAP

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong

1997-01-01

The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP`s cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs.

Accident sequence quantification with KIRAP

International Nuclear Information System (INIS)

Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong.

1997-01-01

The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP's cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs

The IAEA Accident Management Programme

Energy Technology Data Exchange (ETDEWEB)

Kabanov, L.; Jankowski, M.; Mauersberger, H. (International Atomic Energy Agency, Vienna (Austria))

1993-02-01

Accident prevention and mitigation programmes and the Emergency Response System (ERS) are important elements of the Agency's activities in the area of nuclear power plant (NPP) safety. Safety Codes and Guides on siting, design, quality assurance and the operation of NPPs have been produced and are used by NPP operating organizations. Nuclear safety evaluation services are provided by the IAEA. The Emergency Response System and the International Nuclear Event Scale (INES) have been developed. The framework for the development of an accident management programme has been set up. The main goal is to develop an Accident Management Manual to provide a systematic, structured approach to the development and implementation of an accident management programme at NPPs. An outline of the Manual has been distributed and the first draft is available. The component parts are: Co-ordinated research programmes (CRPs) on severe accident management and containment behaviour; the use of vulnerability analysis; mitigation of the effects of hydrogen, and generic symptom oriented emergency operating procedures. The IAEA provides guidance by the dissemination of information on methods for accident management; collates information on approaches in this field in different organizations and countries; and arranges exchange of experience and the promulgation of knowledge through the training of NPP managers and senior technical staff. (orig.).

The IAEA Accident Management Programme

International Nuclear Information System (INIS)

Kabanov, L.; Jankowski, M.; Mauersberger, H.

1993-01-01

Accident prevention and mitigation programmes and the Emergency Response System (ERS) are important elements of the Agency's activities in the area of nuclear power plant (NPP) safety. Safety Codes and Guides on siting, design, quality assurance and the operation of NPPs have been produced and are used by NPP operating organizations. Nuclear safety evaluation services are provided by the IAEA. The Emergency Response System and the International Nuclear Event Scale (INES) have been developed. The framework for the development of an accident management programme has been set up. The main goal is to develop an Accident Management Manual to provide a systematic, structured approach to the development and implementation of an accident management programme at NPPs. An outline of the Manual has been distributed and the first draft is available. The component parts are: Co-ordinated research programmes (CRPs) on severe accident management and containment behaviour; the use of vulnerability analysis; mitigation of the effects of hydrogen, and generic symptom oriented emergency operating procedures. The IAEA provides guidance by the dissemination of information on methods for accident management; collates information on approaches in this field in different organizations and countries; and arranges exchange of experience and the promulgation of knowledge through the training of NPP managers and senior technical staff. (orig.)

Reactivity accident analysis in MTR cores

International Nuclear Information System (INIS)

Waldman, R.M.; Vertullo, A.C.

1987-01-01

The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

Transportation accident response of a high-capacity truck cask for spent fuel

International Nuclear Information System (INIS)

O'Connell, W.J.; Glaser, R.E.; Johnson, G.L.; Perfect, S.A.; McGuinn, E.J.; Lake, W.H.

1995-11-01

Two of the primary goals of this study were (i) to check the structural and thermal performance of the GA-4 cask in a broad range of accidents and (ii) to carry out a severe-accidents analysis as had been addressed in the Modal Study but now using a specific recent cask design and using current-generation computer models and capabilities. At the same time, it was desired to compare the accident performance of the Ga-4 cask to that of the generic truck cask analyzed in the Modal Study. The same range of impact and fire accidents developed in the Modal Study was adopted for this study. The accident-description data base of the Modal Study categorizes accidents into types of collisions with mobile or fixed objects, non-collision accidents, and fires. The mechanical modes of damage may be via crushing, impact, or puncture. The fire occurrences in the Modal Study data are based on truck accident statistics. The fire types are taken to be pool fires of petroleum products from fuel tanks and/or cargoes

Cannabis, alcohol and fatal road accidents.

Science.gov (United States)

Martin, Jean-Louis; Gadegbeku, Blandine; Wu, Dan; Viallon, Vivian; Laumon, Bernard

2017-01-01

This research aims to estimate the relative risks of responsibility for a fatal accident linked to driving under the influence of cannabis or alcohol, the prevalence of these influences among drivers and the corresponding attributable risk ratios. A secondary goal is to estimate the same items for three other groups of illicit drugs (amphetamines, cocaine and opiates), and to compare the results to a similar study carried out in France between 2001 and 2003. Police procedures for fatal accidents in Metropolitan France during 2011 were analyzed and 300 characteristics encoded to provide a database of 4,059 drivers. Information on alcohol and four groups of illicit drugs derived from tests for positivity and potential confirmation through blood analysis. The study compares drivers responsible for causing the accident, that is to say having directly contributed to its occurrence, to drivers involved in an accident for which they were not responsible, and who can be assimilated to drivers in general. The proportion of persons driving under the influence of alcohol is estimated at 2.1% (95% CI: 1.4-2.8) and under the influence of cannabis at 3.4% (2.9%-3.9%). Drivers under the influence of alcohol are 17.8 times (12.1-26.1) more likely to be responsible for a fatal accident, and the proportion of fatal accidents which would be prevented if no drivers ever exceeded the legal limit for alcohol is estimated at 27.7% (26.0%-29.4%). Drivers under the influence of cannabis multiply their risk of being responsible for causing a fatal accident by 1.65 (1.16-2.34), and the proportion of fatal accidents which would be prevented if no drivers ever drove under the influence of cannabis is estimated at 4.2% (3.7%-4.8%). An increased risk linked to opiate use has also been found to be significant, but with low prevalence, requiring caution in interpreting this finding. Other groups of narcotics have even lower prevalence, and the associated extra risks cannot be assessed. Almost a

Cannabis, alcohol and fatal road accidents.

Directory of Open Access Journals (Sweden)

Jean-Louis Martin

Full Text Available This research aims to estimate the relative risks of responsibility for a fatal accident linked to driving under the influence of cannabis or alcohol, the prevalence of these influences among drivers and the corresponding attributable risk ratios. A secondary goal is to estimate the same items for three other groups of illicit drugs (amphetamines, cocaine and opiates, and to compare the results to a similar study carried out in France between 2001 and 2003.Police procedures for fatal accidents in Metropolitan France during 2011 were analyzed and 300 characteristics encoded to provide a database of 4,059 drivers. Information on alcohol and four groups of illicit drugs derived from tests for positivity and potential confirmation through blood analysis. The study compares drivers responsible for causing the accident, that is to say having directly contributed to its occurrence, to drivers involved in an accident for which they were not responsible, and who can be assimilated to drivers in general.The proportion of persons driving under the influence of alcohol is estimated at 2.1% (95% CI: 1.4-2.8 and under the influence of cannabis at 3.4% (2.9%-3.9%. Drivers under the influence of alcohol are 17.8 times (12.1-26.1 more likely to be responsible for a fatal accident, and the proportion of fatal accidents which would be prevented if no drivers ever exceeded the legal limit for alcohol is estimated at 27.7% (26.0%-29.4%. Drivers under the influence of cannabis multiply their risk of being responsible for causing a fatal accident by 1.65 (1.16-2.34, and the proportion of fatal accidents which would be prevented if no drivers ever drove under the influence of cannabis is estimated at 4.2% (3.7%-4.8%. An increased risk linked to opiate use has also been found to be significant, but with low prevalence, requiring caution in interpreting this finding. Other groups of narcotics have even lower prevalence, and the associated extra risks cannot be assessed

Confinement of airborne particulate radioactivity in the case of an accident

International Nuclear Information System (INIS)

Ruedinger, V.; Wilhelm, J.G.

1984-01-01

In the case of an accident, the filter elements on the inlets and exhausts of the air-cleaning systems of a nuclear facility may become a part of the remaining fission product barrier. Among others, the Project Nuclear Safety is pursuing the information necessary to insure safe operation of air-cleaning systems under accident conditions. Experimental investigations into the response of HEPA filters to differential presssures involving both dry and moist air have demonstrated the occurrence of structural failure with subsequent loss of efficiency at low values of differential pressures. Contributions are being made to the development and verification of computer codes used to calculate those fluid-dynamic and thermodynamic conditions expected to prevail in an air-cleaning system as a result of potential accident situations. With regard to further investigations, a new test facility was put into operation for the realization of superimposed challenges and a new method for testing particulate removal efficiency under high temperature or high humidity was developed. (orig./HP)

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)

Utilization of the atmospheric release advisory capability (ARAC) services during and after the Three Mile Island accident

International Nuclear Information System (INIS)

Knox, J.B.; Dickerson, M.H.; Greenly, G.D.; Gudiksen, P.H.; Sullivan, T.J.

1980-01-01

At 0820 PST on 28 March 1979, the Department of Energy's Emergency Operations Center advised the Atmospheric Release Advisory Capability (ARAC) that the Three Mile Island nuclear power plant in Harrisburg, Pennsylvania, had experienced an accident some four hours earlier, resulting in the atmospheric release of xenon-133 and krypton-88. This report describes ARAC's response to the Three Mile Island accident, including the role ARAC played throughout the 20 days that real-time assessments were made available to the Department of Energy on-scene commander. It also describes the follow-up population-dose calculations performed for the President's Commission on Three Mile Island. At the request of the Nuclear Regulatory Commission, a questionnaire addressing the usefulness of ARAC products during the accident was sent to ARAC-product users. A summary of the findings from this questionnaire, along with recommendations for improving ARAC service, is also presented. The accident at Mississauga, Ontario, Canada, is discussed in the context of a well-planned emergency response by local and Federal officials

Development of Krsko Severe Accident Management Database (SAMD)

International Nuclear Information System (INIS)

Basic, I.; Kocnar, R.

1996-01-01

Severe Accident Management is a framework to identify and implement the Emergency Response Capabilities that can be used to prevent or mitigate severe accidents and their consequences. Krsko Severe Accident Management Database documents the severe accident management activities which are developed in the NPP Krsko, based on the Krsko IPE (Individual Plant Examination) insights and Generic WOG SAMGs (Westinghouse Owners Group Severe Accident Management Guidance). (author)

Comparison of containment response by model difference between GOTHIC and CONTEMPT

Energy Technology Data Exchange (ETDEWEB)

Byun, Choong Sup; Song, Dong Soo; Park, Young Chan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2005-07-01

GOTHIC and CONTEMPT are containment performance analysis codes, which are used in calculation of containment pressure and temperature for design basis accidents such as a loss of coolant accident(LOCA) and main steam line break(MSLB) accident. CONTEMPT models vapor and liquid phases, but GOTHIC has one more phase, drop phase. The containment responses between two codes are different each other during initial stage of accident due to this model difference. This paper evaluates the model difference between GOTHIC and CONTEMPT.

Comparison of containment response by model difference between GOTHIC and CONTEMPT

International Nuclear Information System (INIS)

Byun, Choong Sup; Song, Dong Soo; Park, Young Chan

2005-01-01

GOTHIC and CONTEMPT are containment performance analysis codes, which are used in calculation of containment pressure and temperature for design basis accidents such as a loss of coolant accident(LOCA) and main steam line break(MSLB) accident. CONTEMPT models vapor and liquid phases, but GOTHIC has one more phase, drop phase. The containment responses between two codes are different each other during initial stage of accident due to this model difference. This paper evaluates the model difference between GOTHIC and CONTEMPT

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

OpenAIRE

FABBRI LUCIANO; BINDA MASSIMO; BRUINEN DE BRUIN YURI

2017-01-01

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

Visualization of Traffic Accidents

Science.gov (United States)

Wang, Jie; Shen, Yuzhong; Khattak, Asad

2010-01-01

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

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)

Emergency preparedness and response in case of a fire accident with UF6 packages traversing the Suez Canal

International Nuclear Information System (INIS)

Salama, M.

2004-01-01

Egypt has a unique problem, the Suez Canal. Radioactive cargo passes regularly through the canal carrying new and spent reactor fuel. There are also about 1000 metric tonnes of uranium hexafluoride (UF 6 ) passing through the canal every year. In spite of all the precautions taken in the transport, accidents with packages containing UF 6 shipped through the Suez Canal may arise, even though the probability is minimal. Such accidents may be accompanied by injuries to or death of persons and damage to property including radiation and criticality hazards and high chemical toxicity, particularly if the accident occurred close to one of the three densely populated cities (Port Said, Ismailia and Suez), which are located along the west bank of the Suez Canal. The government of Egypt has established a national radiological emergency plan in order to deal with any radiological accidents which may arise inside the country. This paper considers the effect of a fire accident to industrial packages containing UF 6 on board a cargo ship passing along the Suez Canal near Port Said City. The accident scenario and emergency response actions taken during the different phases of the accident are presented and discussed. The paper highlights the importance of public awareness for populations located in densely populated areas along the bank of the Suez Canal, in order to react in a timely and effective way to avoid the toxic and radiological hazards resulting from such a type of accident. The possibility of upgrading the capabilities of civil defence and fire-fighting personnel is also discussed (author)

Core loss during a severe accident (COLOSS)

International Nuclear Information System (INIS)

Adroguer, B.; Bertrand, F.; Chatelard, P.; Cocuaud, N.; Van Dorsselaere, J.P.; Bellenfant, L.; Knocke, D.; Bottomley, D.; Vrtilkova, V.; Belovsky, L.; Mueller, K.; Hering, W.; Homann, C.; Krauss, W.; Miassoedov, A.; Schanz, G.; Steinbrueck, M.; Stuckert, J.; Hozer, Z.; Bandini, G.; Birchley, J.; Berlepsch, T. von; Kleinhietpass, I.; Buck, M.; Benitez, J.A.F.; Virtanen, E.; Marguet, S.; Azarian, G.; Caillaux, A.; Plank, H.; Boldyrev, A.; Veshchunov, M.; Kobzar, V.; Zvonarev, Y.; Goryachev, A.

2005-01-01

The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H 2 production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO 2 and MOX by molten Zircaloy (b) simultaneous dissolution of UO 2 and ZrO 2 (c) oxidation of U-O-Zr mixtures (d) degradation-oxidation of B 4 C control rods. Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B 4 C control rods and in the TMI-2 accident. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Breakthroughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO 2 and MOX dissolution and oxidation of U-O-Zr and B 4 C-metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H 2 production observed during the reflooding of degraded cores under severe accident conditions. The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results. Main results and recommendations for future R and D activities are summarized in this paper

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

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

The numerical simulation of the WWER-440/V-213 reactor pressure vessel internals response to maximum hypothetical large break loss of coolant accident

International Nuclear Information System (INIS)

Hermansky, P.; Krajcovic, M.

2012-01-01

The reactor internals are designed to ensure cooling of the fuel, to ensure the movement of emergency control assemblies under all operating conditions including accidents and facilitate removal of the fuel and of the internals following an accident This paper presents results of the numerical simulation of the WWER-440/V213 reactor vessel internals dynamic response to maximum hypothetical Large-Break Loss of Coolant Accident. The purpose of this analysis is to determine the reactor vessel internals response due to rapid depressurization and to prove no such deformations occur in the reactor vessel internals which would prevent timely and proper activation of the emergency control assemblies. (Authors)

Radionuclide release rate inversion of nuclear accidents in nuclear facility based on Kalman filter

International Nuclear Information System (INIS)

Tang Xiuhuan; Bao Lihong; Li Hua; Wan Junsheng

2014-01-01

The rapidly and continually back-calculating source term is important for nuclear emergency response. The Gaussian multi-puff atmospheric dispersion model was used to produce regional environment monitoring data virtually, and then a Kalman filter was designed to inverse radionuclide release rate of nuclear accidents in nuclear facility and the release rate tracking in real time was achieved. The results show that the Kalman filter combined with Gaussian multi-puff atmospheric dispersion model can successfully track the virtually stable, linear or nonlinear release rate after being iterated about 10 times. The standard error of inversion results increases with the true value. Meanwhile extended Kalman filter cannot inverse the height parameter of accident release as interceptive error is too large to converge. Kalman filter constructed from environment monitoring data and Gaussian multi-puff atmospheric dispersion model can be applied to source inversion in nuclear accident which is characterized by static height and position, short and continual release in nuclear facility. Hence it turns out to be an alternative source inversion method in nuclear emergency response. (authors)

Reactor accident calculation models in use in the Nordic countries

International Nuclear Information System (INIS)

Tveten, U.

1984-01-01

The report relates to a subproject under a Nordic project called ''Large reactor accidents - consequences and mitigating actions''. In the first part of the report short descriptions of the various models are given. A systematic list by subject is then given. In the main body of the report chapter and subchapter headings are by subject. (Auth.)

Influence of precipitations, buildings and over increase of radioactive emission in the population dose calculation under accident conditions

International Nuclear Information System (INIS)

Conte, M.C.

1987-01-01

The influence of precipitations is analyzed, as well as of buildings and emission over increase on the dosis produced on the population as a consequence of some postulated accidents in Atucha II nuclear power plant. The following conclusions were achieved: the calculations performed without considering the above mentioned effects are conservative, excluding the case in which the precipitation is very close to the emission source. In this case, the maximal difference observed was 20% for class C and 5% for class D, at 1 km from source, with a decreasing difference according to the distance. The doses calculated without building effect were approximately 25% greater than those calculated considering this effect for class E and 40% for class F, at 1 km from the source. The difference decreases with distance and increases with the stability of atmospheric conditions. This behaviour is also observed with the over increase effect. In this case, the maximal observed differences were of one order of magnitude for class E and three orders for class F, at 1 km from the source. (Author)

Accident tolerant high-pressure helium injection system concept for light water reactors

International Nuclear Information System (INIS)

Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

2016-01-01

Highlights: • Potential helium injection strategy is proposed for LWR accident scenarios. • Multiple injection sites are proposed for current LWR designs. • Proof-of-concept experimentation illustrates potential helium injection benefits. • Computational studies show an increase in pressure vessel blowdown time. • Current LOCA codes have the capability to include helium for feasibility calculations. - Abstract: While the design of advanced accident-tolerant fuels and structural materials continues to remain the primary focus of much research and development pertaining to the integrity of nuclear systems, there is a need for a more immediate, simple, and practical improvement in the severe accident response of current emergency core cooling systems. Current blowdown and reflood methodologies under accident conditions still allow peak cladding temperatures to approach design limits and detrimentally affect the integrity of core components. A high-pressure helium injection concept is presented to enhance accident tolerance by increasing operator response time while maintaining lower peak cladding temperatures under design basis and beyond design basis scenarios. Multiple injection sites are proposed that can be adapted to current light water reactor designs to minimize the need for new infrastructure, and concept feasibility has been investigated through a combination of proof-of-concept experimentation and computational modeling. Proof-of-concept experiments show promising cooling potential using a high-pressure helium injection concept, while the developed choked-flow model shows core depressurization changes with added helium injection. Though the high-pressure helium injection concept shows promise, future research into the evaluation of system feasibility and economics are needed.Classification: L. Safety and risk analysis

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)

Preliminary Assessment of the Loss of Flow Accident for PGSFR

Energy Technology Data Exchange (ETDEWEB)

Shin, Andong; Choi, Yong Won; Bae, Moohoon [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-10-15

TRACE code have being considered as a candidate tool for SFR audit calculation for licensing review since 2012. On the basis of modeling and precalculation experience for the Demonstration Sodium cooled Fast Reactor (DSFR-600), TRACE code model for PGSFR was developed this year. In this paper, one of representing Design Base Event (DBE), Loss of Flow (LOF) accident was pre-calculated and Locked Rotor (LR) case was compared with LOF case since it could be a possible limiting case for LOF representing DBE. Sensitivity calculation for the LR case was implemented for identifying major parameters for the scenario. For the preparation of the review of licensing application for PGSFR, TRACE model for the PGSFR was developed and the loss of flow accident was precalculated. The locked pump rotor case was also calculated as a possible bounding case for the loss of flow scenario. Pre-calculation showed that the locked rotor case was similar or worst case to the loss of flow accident. Therefore, the locked rotor case should take into account in design base accident assessment of PGSFR. Sensitivity calculations for the rocked rotor case also studied for identification of unfixed design parameters influencing to estimation of inner surface temperature. Sensitivity result showed that the first temperature peak was largely influenced by reactor trip delay and second peak mostly influenced by pump coast down characteristic.

The next nuclear power station generation: Beyond-design accident concepts, methods, and action sequence

International Nuclear Information System (INIS)

Asmolov, V.G.; Khakh, O.Ya.; Shashkov, M.G.

1993-01-01

The problem of beyond-design accidents at nuclear stations will not be solved unless a safety culture becomes a basic characteristic of all lines of activity. Only then can the danger of accidents as an objective feature of nuclear stations be eliminated by purposive skilled and responsible activities of those implementing safety. Nuclear-station safety is provided by the following interacting and complementary lines of activity: (1) the design and construction of nuclear stations by properly qualified design and building organizations; (2) monitoring and supervision of safety by special state bodies; (3) control of the station by the exploiting organization; and (4) scientific examination of safety within the above framework and by independent organizations. The distribution of the responsibilities, powers, and right in these lines should be defined by a law on atomic energy, but there is not such law in Russian. The beyond-design accident problem is a key one in nuclear station safety, as it clear from the serious experience with accidents and numerous probabilistic studies. There are four features of the state of this topic in Russia that are of major significance for managing accidents: the lack of an atomic energy law, the inadequacy of the technical standards, the lack of a verified program package for nuclear-station designs in order to calculate the beyond-design accidents and analyze risks, and a lack of approach by designers to such accidents on the basis of international recommendations. This paper gives a brief description of three-forming points in the scientific activity: the general concept of nuclear-station safety, methods of analyzing and providing accident management, and the sequence of actions developed by specialists at this institute in recent years

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

Emergency preparedness and response: compensating victims of a nuclear accident

International Nuclear Information System (INIS)

Schwartz, Julia

2004-01-01

The 1986 tragedy at the Chernobyl Nuclear Power Plant in Ukraine motivated the entire international nuclear community to ensure that countries would, in the future, be well prepared to manage the physical, psychological and financial consequences of a serious nuclear accident. Since that event, numerous nuclear emergency preparedness and post-emergency management programmes have been established at national and international levels to ensure that appropriate mechanisms will respond to the threat, and the aftermath, of a nuclear accident. The INEX 2000 Workshop on the Indemnification of Nuclear Damage, jointly organised by the OECD/Nuclear Energy Agency and the French Government, was the first ever international programme to address the manner in which victims of a nuclear accident with trans-boundary consequences would be compensated for damage suffered before, during and after the accident. The Workshop results revealed striking differences in the compensation principles and practices implemented in the 30 participating countries, in the co-ordination measures between different public authorities within an affected state, and in the co-operative procedures between the accident state and its neighbours. All participants agreed on the need for improvement in these areas, particularly for maintaining public confidence in governments' ability to properly manage nuclear emergencies

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

International Nuclear Information System (INIS)

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

1993-02-01

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

Development of radiation dose assessment system for radiation accident (RADARAC)

International Nuclear Information System (INIS)

Takahashi, Fumiaki; Shigemori, Yuji; Seki, Akiyuki

2009-07-01

The possibility of radiation accident is very rare, but cannot be regarded as zero. Medical treatments are quite essential for a heavily exposed person in an occurrence of a radiation accident. Radiation dose distribution in a human body is useful information to carry out effectively the medical treatments. A radiation transport calculation utilizing the Monte Carlo method has an advantageous in the analysis of radiation dose inside of the body, which cannot be measured. An input file, which describes models for the accident condition and quantities of interest, should be prepared to execute the radiation transport calculation. Since the accident situation, however, cannot be prospected, many complicated procedures are needed to make effectively the input file soon after the occurrence of the accident. In addition, the calculated doses are to be given in output files, which usually include much information concerning the radiation transport calculation. Thus, Radiation Dose Assessment system for Radiation Accident (RADARAC) was developed to derive effectively radiation dose by using the MCNPX or MCNP code. RADARAC mainly consists of two parts. One part is RADARAC - INPUT, which involves three programs. A user can interactively set up necessary resources to make input files for the codes, with graphical user interfaces in a personnel computer. The input file includes information concerning the geometric structure of the radiation source and the exposed person, emission of radiations during the accident, physical quantities of interest and so on. The other part is RADARAC - DOSE, which has one program. The results of radiation doses can be effectively indicated with numerical tables, graphs and color figures visibly depicting dose distribution by using this program. These results are obtained from the outputs of the radiation transport calculations. It is confirmed that the system can effectively make input files with a few thousand lines and indicate more than 20

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

Accident investigation practices in Europe--main responses from a recent study of accidents in industry and transport.

Science.gov (United States)

Roed-Larsen, Sverre; Valvisto, T; Harms-Ringdahl, L; Kirchsteiger, C

2004-07-26

Europe has during recent years been shocked by disasters from natural events and technical breakdowns. The consequences have been comprehensive, measured by lost lives, injuries, and material and environmental damage. ESReDA wanted in 2000--by setting up a special expert group on accident investigation--to clarify the state of art of accident investigation practices and to map the use of thoroughly accident investigation in order to learn lessons from past disasters and prevent new ones. The scope was to cover three sectors in the society: transport, production processes and storage of hazardous materials, and energy production. The main method used was a questionnaire, which was sent in 2001 to about 150 organisations. About 50 replies were analysed. The replies showed great variations but also similarities, among others in definition of accident and incident, the objectives of the investigation team, criteria used to start an investigation, the status of the investigation organisation, the flow of information, the composition of the investigation team, and the use of internal or international procedures or rules. Several methods (in total 14 different methods were mentioned) were used for carrying out accident /incident investigations. Most of the respondents were willing to co-operate in one or another way with ESReDA. Although there are important biases in the material, the results from questionnaire are important inputs to the future work of ESReDA Expert group in this field. 3 safety approaches have been identified.

ESBWR long term containment response to loss of coolant accidents

International Nuclear Information System (INIS)

Alamgir, M. D.; Marquino, W.; Diaz-Quiroz, J.; Tucker, L.

2010-01-01

ESBWR is a 4500 MWt generation III+ natural circulation reactor with an array of robust passive safety systems to keep the reactor safe during postulated transients and accidents. With the submittal of the latest revision of the Design Control Document (DCD) to US Nuclear Regulatory Commission, ESBWR is nearing the completion of the US certification process. This paper focuses on the bounding licensing analysis of the long-term (30-day) response of the ESBWR containment to limiting Loss of Coolant Accident (LOCA) performed with the TRACG code. It is shown that using only passive systems available during the first 72 hours after the limiting Main Steam Line Break LOCA, the predicted peak containment pressure in the ESBWR containment remain well below the design limits with good margin. After 72 hours of LOCA initiation, PCCS Vent Fans (non-safety system) become available that remove non-condensable gases from, and further enhance the effectiveness of, PCCS heat exchangers to reduce the containment pressure and temperature to values substantially below the design limits. During the post- 72 hour period, the beneficial effects of the Vent Fan operation, combined with the available operator action to refill of PCCS pools, continue to maintain the containment pressure to about 30% below the design limit at 30 days after a limiting ESBWR LOCA. (authors)

Key Characteristics of Combined Accident including TLOFW accident for PSA Modeling

Energy Technology Data Exchange (ETDEWEB)

Kim, Bo Gyung; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Technology and Research, Abu Dhabi (United Arab Emirates)

2015-05-15

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

PWR pressure vessel integrity during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.

1981-01-01

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

Criteria for calculating the efficiency of HEPA filters during and after design basis accidents

International Nuclear Information System (INIS)

Bergman, W.; First, M.W.; Anderson, W.L.; Gilbert, H.; Jacox, J.W.

1994-12-01

We have reviewed the literature on the performance of high efficiency particulate air (HEPA) filters under normal and abnormal conditions to establish criteria for calculating the efficiency of HEPA filters in a DOE nonreactor nuclear facility during and after a Design Basis Accident (DBA). The literature review included the performance of new filters and parameters that may cause deterioration in the filter performance such as filter age, radiation, corrosive chemicals, seismic and rough handling, high temperature, moisture, particle clogging, high air flow and pressure pulses. The deterioration of the filter efficiency depends on the exposure parameters; in severe exposure conditions the filter will be structurally damaged and have a residual efficiency of 0%. Despite the many studies on HEPA filter performance under adverse conditions, there are large gaps and limitations in the data that introduce significant error in the estimates of HEPA filter efficiencies under DBA conditions. Because of this limitation, conservative values of filter efficiency were chosen when there was insufficient data

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.)

Uncertainty analysis in calculations of a road accident consequences

International Nuclear Information System (INIS)

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

1995-01-01

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

Occupational Radiation Protection in Severe Accident Management

International Nuclear Information System (INIS)

2015-01-01

As an early response to the Fukushima Daiichi NPP accident, the Information System on Occupational Exposure (ISOE) Bureau decided to focus on the following issues as an initial response of the joint program after having direct communications with the Japanese official participants in April 2011: - Management of high radiation area worker doses: It has been decided to make available the experience and information from the Chernobyl accident in terms of how emergency worker / responder doses were legally and practically managed, - Personal protective equipment for highly-contaminated areas: It was agreed to collect information about the types of personnel protective equipment and other equipment (e.g. air bottles, respirators, air-hoods or plastic suits, etc.), as well as high-radiation area worker dosimetry use (e.g. type, number and placement of dosimetry) for different types of emergency and high-radiation work situations. Detailed information was collected on dose criteria which are used for emergency workers /responders and their basis, dose management criteria for high dose/dose rate areas, protective equipment which is recommended for emergency workers / responders, recommended individual monitoring procedures, and any special requirement for assessment from the ISOE participating nuclear utilities and regulatory authorities and made available for Japanese utilities. With this positive response of the ISOE official participants and interest in the situation in Fukushima, the Expert Group on Occupational Radiation Protection in Severe Accident Management (EG-SAM) was established by the ISOE Management Board in May 2011. The overall objective of the EG-SAM is to contribute to occupational exposure management (providing a view on management of high radiation area worker doses) within the Fukushima plant boundary with the ISOE participants and to develop a state-of-the-art ISOE report on best radiation protection management practices for proper radiation

Containment severe accident thermohydraulic phenomena

International Nuclear Information System (INIS)

Frid, W.

1991-08-01

This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)

A modular structure to accident identification using neural networks

International Nuclear Information System (INIS)

Duque Estrada, Cassius Rodrigo

2005-01-01

This work uses the accident identification method based on Artificial Neural Networks (ANN) as basic blocks of a modular structure, allowing the inclusion of new accidents to be identified without modifying the ANN already trained. This structure comprises several modules for accident identification and one module for analysis. Each identification module follows the structure of the basic block. The identification modules are responsible for the recognition of an accident belonging to the specific set of events for which it were trained. The analysis module processes the output from the identification module to determine the system response. In order to test this structure it was proposed a transient identification problem comprising fifty accidents distributed in five identification modules. The results have demonstrated that the accident identification method used as basic block of a modular structure allows the inclusion of new sets of accidents, or variations of a same accident, without modifying the ANN already trained. For this, it is enough to include into the system an specific module for this new set of accidents. (author)

System Response Analysis of Rod Ejection Accident for APR1400 Using KNAP Hot Spot Model

International Nuclear Information System (INIS)

Kim, Yo-Han; Ha, Sang-Jun; Jun, Hwang-Yong

2006-01-01

Korea Electric Power Research Institute (KEPRI) has been developed the non-loss-of-coolant accident (non- LOCA) analysis methodology, called as the Korea Non- LOCA Analysis Package (KNAP), for the typical Optimized Power Reactor 1000 (OPR1000) plants. Considering current licensing methodology conducted by ABB-CE, however, the KNAP could be applied to Advanced Power Reactor 1400 (APR1400) also. In spite of some difference in design concepts of two plant types, there is a close resemblance between their nuclear steam supply systems (NSSS). So, in this study, the rod ejection accident (REA) event was analyzed using KNAP hot spot model (HSM) for APR1400 to estimate the feasibility of the application and the results were compared with those given in APR1400 Standard Safety Analysis Report (SSAR), which were calculated using the CESEC-III and STRIKIN-II code of ABB-CE. Through the study, it was concluded that the KNAP could be applicable to APR1400 on the view point of REA

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

Latin American radio pathology net, training program for the medical response in cases of accidents

International Nuclear Information System (INIS)

Perez, M. R.; Valverde, N.; Sanhueza, S.; Di Tramo, J. L.; Gisone, P.; Cardenas, J.

2003-01-01

In Latin America exists a wide application of the nuclear techniques in different fields. These tendencies, to the increasing use of nuclear techniques, should carry out the invigoration of the infrastructures and the development of the authorities in radiation protection charge of the regulate and control the sources and practical, associated to ionizing radiation, guided to guarantee their safe use and in consequence to minimize the derived risks of the same ones. In spite of the efforts before mentioned a potential possibility of occurrence of accidental radiological events linked to human errors and violation's of the principles of the radiation protection exists. Reason that they advise to have response capacities to confront and to mitigate the consequences in situations of radiological accidents, including in the same ones the medical assistance of the accident victims. However, the radiological accidents happened in the international environment in the last decades, they have demonstrated inability paradoxically to confront with effectiveness these fortuitous events. Being characterized additionally by the insufficient training of the medical professionals to interpret and to act in consequence before the prejudicial effects to the health of the ionizing radiation. In our geographical context this situation is even more complicated, if we consider the happened radiological accidents of span that put in risk the life and the health of people involved in the same ones, for examples: Argentina (1968, 1963), Brazil (1985, 1987, 1995), Costa Rica (1996). El Salvador (1989), Mexico (1962, 1983, 1984) and Peru (1984, 1999). These reasons justify the development of an action program with the purpose of the consolidate and to integrate the capacity of response of our countries as regards radiological emergencies. Regrettably in Latin American subsist inequalities the development radiation protection programs, that propitiate the possibility of accidental situations

Report on a radiotherapy underdose accident

Energy Technology Data Exchange (ETDEWEB)

Christodoulides, G; Christofides, S [Medical Physics Department, Nicosia General Hospital, 1450 Nicosia (Cyprus)

1999-12-31

Reporting information on accidents and incidents involving radiation sources provides a body of knowledge which can help to prevent accidents of a similar nature. Accident information has to be made available to users, manufacturers and regulators; An international effort to pool and analyse incident and accident information will provide more complete and reliable indicators of root causes and trends and recommendations for future accident avoidance. An accident due to human error involving a superficial x-ray therapy machine and patients treated for postoperative breast cancer is reported here. 43 women receiving radiotherapy treatment have received significantly less radiation dose than the prescribed dose. The worst dose percentage within the radiation field was 20% of the prescribed dose. The worst dose percentage on the operation scar of the breast was 52% of the prescribed radiation dose. The response to accidents/incidents in radiotherapy is discussed. (authors) 4 refs., 5 figs., 1 tabs.

Accidents - Chernobyl accident

International Nuclear Information System (INIS)

2004-01-01

This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

Description of the information and calculation system for combatment of accidents with hazardous materials

International Nuclear Information System (INIS)

Scheur, M.J. van de; Stolk, D.J.

1987-04-01

On request of the Netherlands government by TNO a decision support system is developed for the assessment of the off-site consequences of an accident with toxic or radioactive materials. The interactive system supports the emergency planning in two ways. First, the risk to the residents in the surroundings of the accident is quantified in terms of severity and magnitude. Second, a set of countermeasures is evaluated by which an optimum strategy to reduce the impact of the accident can be determined. At this moment the system is in a development stage. It turned out that even the preliminary system provides information to the decision process that is urgently needed. This specifically refers to the introduction of the time aspects and the quantification of the damage. 7 refs.; 8 figs.; 3 tabs

Introducing PCTRAN as an evaluation tool for nuclear power plant emergency responses

International Nuclear Information System (INIS)

Cheng, Yi-Hsiang; Shih, Chunkuan; Chiang, Show-Chyuan; Weng, Tung-Li

2012-01-01

Highlights: ► PCTRAN is integrated with an atmospheric dispersion algorithm. ► The improved PCTRAN acts as an accident/incident simulator and a data exchange system. ► The software helps the responsible organizations decide the rescue and protective actions. ► The evaluation results show the nuclear power plant accident and its off-site dose consequences. ► The software can be used for nuclear power plant emergency responses. - Abstract: Protecting the public from radiation exposure is important if a nuclear power plant (NPP) accident occurs. Deciding appropriate protective actions in a timely and effective manner can be fulfilled by using an effective accident evaluation tool. In our earlier work, we have integrated PCTRAN (Personal Computer Transient Analyzer) with the off-site dose calculation model. In this study, we introduce PCTRAN as an evaluation tool for nuclear power plant emergency responses. If abnormal conditions in the plant are monitored or observed, the plant staffs can distinguish accident/incident initiation events. Thus, the responsible personnel can immediately operate PCTRAN and set up those accident/incident initiation events to simulate the nuclear power plant transient or accident in conjunction with off-site dose distributions. The evaluation results consequently help the responsible organizations decide the rescue and protective actions. In this study, we explain and demonstrate the capabilities of PCTRAN for nuclear emergency responses, through applying it to simulate the postulated nuclear power plant accident scenarios.

Response Analysis on Electrical Pulses under Severe Nuclear Accident Temperature Conditions Using an Abnormal Signal Simulation Analysis Module

Directory of Open Access Journals (Sweden)

Kil-Mo Koo

2012-01-01

Full Text Available Unlike design basis accidents, some inherent uncertainties of the reliability of instrumentations are expected while subjected to harsh environments (e.g., high temperature and pressure, high humidity, and high radioactivity occurring in severe nuclear accident conditions. Even under such conditions, an electrical signal should be within its expected range so that some mitigating actions can be taken based on the signal in the control room. For example, an industrial process control standard requires that the normal signal level for pressure, flow, and resistance temperature detector sensors be in the range of 4~20 mA for most instruments. Whereas, in the case that an abnormal signal is expected from an instrument, such a signal should be refined through a signal validation process so that the refined signal could be available in the control room. For some abnormal signals expected under severe accident conditions, to date, diagnostics and response analysis have been evaluated with an equivalent circuit model of real instruments, which is regarded as the best method. The main objective of this paper is to introduce a program designed to implement a diagnostic and response analysis for equivalent circuit modeling. The program links signal analysis tool code to abnormal signal simulation engine code not only as a one body order system, but also as a part of functions of a PC-based ASSA (abnormal signal simulation analysis module developed to obtain a varying range of the R-C circuit elements in high temperature conditions. As a result, a special function for abnormal pulse signal patterns can be obtained through the program, which in turn makes it possible to analyze the abnormal output pulse signals through a response characteristic of a 4~20 mA circuit model and a range of the elements changing with temperature under an accident condition.

49 CFR 840.3 - Notification of railroad accidents.

Science.gov (United States)

2010-10-01

... 49 Transportation 7 2010-10-01 2010-10-01 false Notification of railroad accidents. 840.3 Section... SAFETY BOARD RULES PERTAINING TO NOTIFICATION OF RAILROAD ACCIDENTS § 840.3 Notification of railroad accidents. The operator of a railroad shall notify the Board by telephoning the National Response Center at...

EPRI nuclear fuel-cycle accident risk assessment

International Nuclear Information System (INIS)

Anon.

1981-01-01

The present results of the nuclear fuel-cycle accident risk assessment conducted by the Electric Power Research Institute show that the total risk contribution of the nuclear fuel cycle is only approx. 1% of the accident risk of the power plant; hence, with little error, the accident risk of nuclear electric power is essentially that of the power plant itself. The power-plant risk, assuming a very large usage of nuclear power by the year 2005 is only approx. 0.5% of the radiological risk of natural background. The smallness of the fuel-cycle risk relative to the power-plant risk may be attributed to the lack of internal energy to drive an accident and the small amount of dispersible material. This work aims at a realistic assessment of the process hazards, the effectiveness of confinement and mitigation systems and procedures, and the associated likelihood of errors and the estimated size of errors. The primary probabilistic estimation tool is fault-tree analysis, with the release source terms calculated using physicochemical processes. Doses and health effects are calculated with CRAC (Consequences of Reactor Accident Code). No evacuation or mitigation is considered; source terms may be conservative through the assumption of high fuel burnup (40,000 MWd/t) and short cooling period (90 to 150 d); high-efficiency particulate air filter efficiencies are derived from experiments

[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.

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

Differences in frequency rates and classes of occupational accidents in copper ore and black coal mines

Energy Technology Data Exchange (ETDEWEB)

Stecko, R.; Zacharzewski, J.

1987-02-01

This paper analyzes occupational accidents in copper mines and black coal mines in Poland from 1968 to 1985. Evaluates accident rate calculated per 1000 employees and per 100,000 t coal or copper ore output and distribution of accident classes characterized on the basis of miner absenteeism caused by an accident. Accident rate in coal mines calculated per 1,000 miners was: 58.5 in 1981, 51.0 in 1982, 39.5 in 1983, 36.6 in 1984 and 32.1 in 1985. Accident rate calculated per 100,000 t coal output was: 5.8 in 1980, 13.7 in 1981, 10.7 in 1982, 8.2 in 1983, 7.8 in 1984 and 6.7 in 1985. Curves showing accident rates and indices characterizing accident classes in 1968-1985 are plotted. Comparative evaluations of accident rate fluctuations in copper and coal mines suggest that these fluctuations are associated with causes other than mining and geologic conditions. Certain similiarities in accident rate fluctuations could be determined in spite of differing mining and geologic conditions and mine service life in the Lubin copper basin and Upper Silesia coal basin. 4 refs.

Calculating the Responses of Self-Powered Radiation Detectors.

Science.gov (United States)

Thornton, D. A.

Available from UMI in association with The British Library. The aim of this research is to review and develop the theoretical understanding of the responses of Self -Powered Radiation Detectors (SPDs) in Pressurized Water Reactors (PWRs). Two very different models are considered. A simple analytic model of the responses of SPDs to neutrons and gamma radiation is presented. It is a development of the work of several previous authors and has been incorporated into a computer program (called GENSPD), the predictions of which have been compared with experimental and theoretical results reported in the literature. Generally, the comparisons show reasonable consistency; where there is poor agreement explanations have been sought and presented. Two major limitations of analytic models have been identified; neglect of current generation in insulators and over-simplified electron transport treatments. Both of these are developed in the current work. A second model based on the Explicit Representation of Radiation Sources and Transport (ERRST) is presented and evaluated for several SPDs in a PWR at beginning of life. The model incorporates simulation of the production and subsequent transport of neutrons, gamma rays and electrons, both internal and external to the detector. Neutron fluxes and fuel power ratings have been evaluated with core physics calculations. Neutron interaction rates in assembly and detector materials have been evaluated in lattice calculations employing deterministic transport and diffusion methods. The transport of the reactor gamma radiation has been calculated with Monte Carlo, adjusted diffusion and point-kernel methods. The electron flux associated with the reactor gamma field as well as the internal charge deposition effects of the transport of photons and electrons have been calculated with coupled Monte Carlo calculations of photon and electron transport. The predicted response of a SPD is evaluated as the sum of contributions from individual

CONTEMPT-LT/028: a computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

International Nuclear Information System (INIS)

Hargroves, D.W.; Metcalfe, L.J.; Wheat, L.L.; Niederauer, G.F.; Obenchain, C.F.

1979-03-01

CONTEMPT-LT is a digital computer program, written in FORTRAN IV, developed to describe the long-term behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments. The program is capable of describing the effects of leakage on containment response. Models are provided to describe fan cooler and cooling spray engineered safety systems. An annular fan model is also provided to model pressure control in the annular region of dual containment systems. Up to four compartments can be modeled with CONTEMPT-LT, and any compartment except the reactor system may have both a liquid pool region and an air--vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different

Elements of a national emergency response system for nuclear accidents

International Nuclear Information System (INIS)

Dickerson, M.H.

1987-01-01

The purpose of this paper is to suggest elements for a general emergency response system, employed at a national level, to detect, evaluate and assess the consequences of a radiological atmospheric release occurring within or outside of national boundaries. These elements are focused on the total aspect of emergency response ranging from providing an initial alarm to a total assessment of the environmental and health effects. Elements of the emergency response system are described in such a way that existing resources can be directly applied if appropriate; if not, newly developed or an expansion of existing resources can be employed. The major thrust of this paper is toward a philosophical discussion and general description of resources that would be required to implementation. If the major features of this proposal system are judged desirable for implementation, then the next level of detail can be added. The philosophy underlying this paper is preparedness - preparedness through planning, awareness and the application of technology. More specifically, it is establishment of reasonable guidelines including the definition of reference and protective action levels for public exposure to accidents involving nuclear material; education of the public, government officials and the news media; and the application of models and measurements coupled to computer systems to address a series of questions related to emergency planning, response and assessment. It is the role of a proven national emergency response system to provide reliable, quality-controlled information to decision makers for the management of environmental crises

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Possibility of reducing costs of mining operations - economic aspects of workplace accidents

Science.gov (United States)

Duda, Adam

2017-11-01

The article presents methods of calculating costs of workplace accidents incurred by an employer, and the influence of the number and severity of accidents on changes in the amount of accident insurance contribution paid by an employer within the framework of the social security system.

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

Review of off-site emergency preparedness and response plan of Indian NPPs based on experience of Fukushima nuclear accident

International Nuclear Information System (INIS)

Singh, Hukum; Dash, M.; Shukla, Vikas; Vijayan, P.; Krishnamurthy, P.R.

2012-01-01

Nuclear power plants in India are designed, constructed and operated based on the principle of the highest priority to nuclear safety. To deal with any unlikely situation of radiological emergency, the emergency preparedness and response plans are ensured to be in place at all NPPs prior to their commissioning. These plans are periodically reviewed and tested by conducting emergency exercise with the participation of various agencies such as Nuclear Power Corporation of India Limited, NDMA, district authorities, regulatory body and general public. On March 11, 2011 an earthquake of magnitude 9.0 hit the Fukushima Dai-ichi and Dai-ni followed by tsunami waves of height 15 meters above reference sea level. This resulted in large scale release of radioactive material from Fukushima Dai-ichi NPS. This led to the evacuation of a large number of people from the areas surrounding the affected nuclear power plants. The event was rated as level 7 event in International Nuclear Event Scale (INES). The event also revealed the challenges in handling radiological emergency situation in adverse environmental conditions, The experience of managing radiological emergency situation during Fukushima nuclear accident provides opportunities to review and improve emergency preparedness and response programme. The present paper presents the chronology of the emergency situation, challenges faced and handled in Fukushima. Even though the possibility of a Fukushima type nuclear accident in India is very remote due to the low probability of a high intensity earthquake followed by tsunami at NPP sites, the efforts needs to be initiated from the regulatory point of view for an effective Nuclear and Radiological Emergency Preparedness and Response Plans. The Emergency Preparedness and Response Plans of NPP sites were reviewed in the light of unique challenges of accident at Fukushima. It is realized that multi unit events are the realities that must be addressed as part of Emergency

Review of off-site emergency preparedness and response plan of Indian NPPs based on experience of Fukushima nuclear accident

Energy Technology Data Exchange (ETDEWEB)

Singh, Hukum; Dash, M.; Shukla, Vikas; Vijayan, P.; Krishnamurthy, P.R., E-mail: vshukla@aerb.gov.in [Operating Plants Safety Division, Atomic Energy Regulatory Board, Mumbai (India)

2012-07-01

Nuclear power plants in India are designed, constructed and operated based on the principle of the highest priority to nuclear safety. To deal with any unlikely situation of radiological emergency, the emergency preparedness and response plans are ensured to be in place at all NPPs prior to their commissioning. These plans are periodically reviewed and tested by conducting emergency exercise with the participation of various agencies such as Nuclear Power Corporation of India Limited, NDMA, district authorities, regulatory body and general public. On March 11, 2011 an earthquake of magnitude 9.0 hit the Fukushima Dai-ichi and Dai-ni followed by tsunami waves of height 15 meters above reference sea level. This resulted in large scale release of radioactive material from Fukushima Dai-ichi NPS. This led to the evacuation of a large number of people from the areas surrounding the affected nuclear power plants. The event was rated as level 7 event in International Nuclear Event Scale (INES). The event also revealed the challenges in handling radiological emergency situation in adverse environmental conditions, The experience of managing radiological emergency situation during Fukushima nuclear accident provides opportunities to review and improve emergency preparedness and response programme. The present paper presents the chronology of the emergency situation, challenges faced and handled in Fukushima. Even though the possibility of a Fukushima type nuclear accident in India is very remote due to the low probability of a high intensity earthquake followed by tsunami at NPP sites, the efforts needs to be initiated from the regulatory point of view for an effective Nuclear and Radiological Emergency Preparedness and Response Plans. The Emergency Preparedness and Response Plans of NPP sites were reviewed in the light of unique challenges of accident at Fukushima. It is realized that multi unit events are the realities that must be addressed as part of Emergency

The Human Aspect of the Fukushima Daiichi Accident

International Nuclear Information System (INIS)

Anegawa, T.; Kawano, A.

2016-01-01

Recognizing itself as the main party involved in the nuclear accident triggered by the Tohoku-Chihou-Taiheiyo-Oki Earthquake on March 11, 2011, Tokyo Electric Power Company (TEPCO) has performed accident investigation from various aspects. Results of the investigation are reported mainly in two reports; (1) Fukushima Nuclear Accident Analysis Report (June 20, 2012), which identified the timeline and the proximate causes of the accident, and (2) Summary of Fukushima Nuclear Accident and Nuclear Safety Reform Plan (March 29, 2013) to set forth the results of the investigation and provide an analysis of the background factors surrounding the accident and countermeasures taken. This presentation will first provide overview of the accident response at Fukushima Daiichi and Daini Nuclear Power Stations. Voices from the first responders at the sites will be introduced in order to share thoughts of individuals involved in the emergency response. Summary of retrospective study of the accident by one of the shift supervisors at the time of the accident will be presented in order to share the facts that happened at main control rooms. The shift supervisor and his crew had to manage the situation for extended period of time that exceeded the scenarios that they had been trained, in a situation with no lightning and high radiation condition. During the accident response, shift supervisors had to decide to dispatch some of his crew members to the field to open valves, check the status of equipment etc., in the situation where the high radiation exposure is foreseen. The presentation will include conflict of shift supervisors and crew focusing on the human aspects. In addition, actions being taken at the Emergency Response Centers (ERC) set up at the seismic-isolated building on-site and the Headquarters in Tokyo will be shared focusing on the human aspects related to the accident progress. This includes difficult decisions to dispatch first responders to the field, in the

Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 2

International Nuclear Information System (INIS)

Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

1993-08-01

The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 1

International Nuclear Information System (INIS)

Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

1993-08-01

The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

Calculation of stricken to mortality and incidence cancers due to beyond design basis accidents of the Esfahan Fuel Production Factory

International Nuclear Information System (INIS)

Heydari Azar, A.; Shahshahani, M.; Roshanzamir, M.; Sabouhi, R.

2008-01-01

In this investigation the amount of absorbed doses by the different pathways of Cloud shine, Ground shine, deposition of radioactive materials on skin and cloths, ingestion, inhalation and the consequences of radioactive material releases due to Beyond Design Basis Accidents such as fire, sintering furnace explosion, criticality and earthquake in Esfahan Fuel Production factory by the residents are evaluated. The calculations related to atomic cloud distribution, estimation of delivered dose and decay chains are performed by PCCOSYMA dose. These computations are based on radioactive source terms, distribution height of radioactive materials. actions for reducing the absorbed dose, human body physiological characteristics, metrological condition and population distribution. Finally, the number of peoples who are stricken to mortality and morbidity cancers and risk values are calculated for 1 year and 50 years

Transient response of small molten salt reactor at duct blockage accident

International Nuclear Information System (INIS)

Yamamoto, Takahisa; Mitachi, Koshi; Ikeuchi, Koji; Suzuki, Takashi

2005-01-01

This paper performed transient core analysis of a small Molten Salt Reactor (MSR) at the time of a duct blockage accident. The numerical model employed in this study consists of continuity and momentum conservation equations for fuel salt flow, two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The analysis shows that (1) the effective multiplication factor and reactor power after the blockage accident hardly change because of the self-control performance of the MSR, (2) fuel salt and graphite moderator temperatures rise at the blockage point and its vicinity, drastically but locally, (3) the highest temperature after the blockage accident is 1 363 K, very lower than the boiling point of fuel salt and melt point of reactor vessel, (4) fast and thermal neutron fluxes distributions after the blockage accident hardly change, and (5) delayed neutron precursors accumulate at the blockage point, especially 1st delayed neutron precursor due to is large decay constant. These results lead that the safety of MSR is assured in the blockage accident. (author)

Radiological Consequence Analyses Following a Hypothetical Severe Accident in Japan

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

International Nuclear Information System (INIS)

Hsii, Y.H.

1976-06-01

The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The report describes the analytical model used for the program. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The user is required to input the description of the discharge of coolant, the boiling of residual water by reactor decay heat, the superheating of steam passing through the core, and metal-water reactions. The reactor building is separated into liquid and vapor regions. Each region is in thermal equilibrium itself, but the two may not be in thermal equilibrium; the liquid and gaseous regions may have different temperatures. The reactor building is represented as consisting of several heat-conducting structures whose thermal behavior can be described by the one-dimensional multi-region heat conduction equation. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc

A new approach to calculating spatial impulse responses

DEFF Research Database (Denmark)

Jensen, Jørgen Arendt

1997-01-01

Using linear acoustics the emitted and scattered ultrasound field can be found by using spatial impulse responses as developed by Tupholme (1969) and Stepanishen (1971). The impulse response is calculated by the Rayleigh integral by summing the spherical waves emitted from all of the aperture...

Psychological and social impacts of post-accident situations: lessons from the Chernobyl accident

International Nuclear Information System (INIS)

Lochard, J.

1996-01-01

This paper presents the main features, from the psychological and social points of view, of the post-accident situation in the contaminated areas around Chernobyl. This is based on a series of surveys performed in the concerned territories of the CIS republics. The high level of stress affecting a large segment of the population is related to the perception of the situation by those living in a durably contaminated environment but also to the side-effects of some of the countermeasures adopted to mitigate the radiological consequences or to compensate the affected population. The distinction between the accident and the post-accident phase is enlarged to take into account the various phases characterizing the dynamics of the social response. Although the size of the catastrophe as well as the economic and political conditions that were prevailing at the time and after the accident have resulted in a maximal intensity of the reactions of the population, many lessons can be drawn for the management of potential post-accident situations. (author)

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

An analysis of the Three Mile Island accident

International Nuclear Information System (INIS)

Brooks, G.L.; Siddal, E.

1980-09-01

Starting with a systematic analysis of the chain of events that took place during the Three Mile Island accident, the authors assess the significance of the four distinct phases of the accident. Inferences that can be drawn with respect to the safety of CANDU reactors are discussed. A rational reaction to the accident is suggested, and several factors are shown not to have played an important part, contrary to public impressions. The authors point out that over-reaction to the accident could detract from public safety. The Canadian response to the accident is discussed. (auth)

An approach for estimating the radiological significance of a hypothetical major nuclear accident over long distance transboundary scales

Energy Technology Data Exchange (ETDEWEB)

Mitrakos, D., E-mail: dimitris.mitrakos@eeae.gr; Potiriadis, C.; Housiadas, C.

2016-04-15

Highlights: • Actions may be warranted after a major nuclear accident even at long distances. • Distance may not be the decisive parameter for longer term radiological impact. • Remote impact may vary orders of magnitude depending on the meteorological conditions. • The potential impact can be assessed using computationally inexpensive calculations. - Abstract: After the Fukushima accident important initiatives were taken in European level to enhance the nuclear safety level of the existing and planned nuclear reactors, such as the so-called nuclear “stress-tests” and the amendment of the Nuclear Safety Directive. A recent work of HERCA and WENRA focused on the need for a more consistent and harmonized response in a transboundary context in case of a hypothetical major nuclear accident in Europe. Such an accident, although very improbable, cannot be totally excluded and so, should be considered in emergency preparedness arrangements among the various European countries. In case of a hypothetical severe Fukushima-like accident in Europe, the role of the neighboring countries may be important, since the authorities should be able to provide information and advice to the government and the public, but also can contribute to the overall assessment of the situation be their own means. In this work we assess the radiological significance of a hypothetical major nuclear accident for distances longer than 300 km that are not typically covered by the internationally accepted emergency planning zones. The approach is simple and computationally inexpensive, since it is based on the calculation of only a few release scenarios at dates selected within a whole year on the basis of bounding the deposition levels at long distances in relation to the occurrence of precipitation. From the calculated results it is evident that distance is not the only decisive parameter in estimating the potential radiological significance of a severe nuclear accident. The hypothetical

SECURE-400 MW: Failure analysis and calculation of release

International Nuclear Information System (INIS)

Bento, J.-P.

1978-11-01

The environmental effects of SECURE-400 MW has been investigated for normal operation and after an accident. When calculating the doses during normal operation it has been assumed that 0.1 % of the fuel elements in the core have fuel cladding damages. The doses after an accident have been calculated only for the cases when there might be an effect on the environment: accidents at change of fuel, fracture of the biggest gas pipe and fracture of the main coolant loop. The result of the investigation is that a core melting accident in SECURE 400 is impossible due to the design of the different systems. Other typs of accidents which might happen will give extremly low doses to the environment. The doses will not differ significantly from doses during normal operation. (K.K.)

LMFBR subassembly response to local pressure loadings: an experimental approach

International Nuclear Information System (INIS)

Marciniak, T.J.; Ash, J.E.; Marchertas, A.H.; Cagliostro, D.J.

1975-01-01

An experimental program to determine the response of LMFBR-type subassemblies to local subassembly accidents caused by pressure loadings is described. Some results are presented and compared with computer calculations

Chernobyl reactor accident

International Nuclear Information System (INIS)

1986-05-01

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

WHO's public health agenda in response to the Fukushima Daiichi nuclear accident

International Nuclear Information System (INIS)

Van Deventer, Emilie; Del Rosario Perez, Maria; Carr, Zhanat; Tritscher, Angelika; Fukushima, Kazuko

2012-01-01

The World Health Organization (WHO) has responded to the 2011 East-Japan earthquake and tsunami through the three levels of its decentralised structure. It has provided public health advice regarding a number of issues relating to protective measures, potassium iodide use, as well as safety of food and drinking water, mental health, travel, tourism, and trade. WHO is currently developing an initial health risk assessment linked to a preliminary evaluation of radiation exposure around the world from the Fukushima Daiichi nuclear accident. Lessons learned from this disaster are likely to help future emergency response to multi-faceted disasters. (note)

Safety regulations regarding to accident monitoring and accident sampling at Russian NPPs with VVER type reactors

International Nuclear Information System (INIS)

Sharafutdinov, Rachet; Lankin, Michail; Kharitonova, Nataliya

2014-01-01

The paper describes a tendency by development of regulatory document requirements related to accident monitoring and accident sampling at Russia's NPPs. Lessons learned from the Fukushima Daiichi accident pointed at the importance and necessary to carry out an additional safety check at Russia's nuclear power plants in the preparedness for management of severe accidents at NPPs. Planned measures for improvement of severe accidents management include development and implementation of the accident instrumentation systems, providing, monitoring, management and storage of information in a severe accident conditions. The draft of Safety Guidelines <accident monitoring system of nuclear power plants with VVER reactors' prepared by Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) established the main criteria for accident monitoring instrumentation that can monitor relevant plant parameters in the reactor and inside containment during and after a severe accident in nuclear power plants. Development of these safety guidelines is in line with the recommendations of IAEA Action Plan on Nuclear Safety in response to the Fukushima Daiichi event and recommendations of the IAEA Nuclear Energy series Report <<Accident Monitoring Systems for Nuclear Power Plants' (Draft V 2.7). The paper presents the principles, which are used as the basis for selection of plant parameters for accident monitoring and for establishing of accident monitoring instrumentation. The recommendations to the accident sampling system capable to obtain the representative reactor coolant and containment air and fluid samples that support accurate analytical results for the parameters of interest are considered. The radiological and chemistry parameters to be monitored for primary coolant and sump and for containment air are specified. (author)

Planning and preparing for emergency response to transport accidents involving radioactive material. Safety guide

International Nuclear Information System (INIS)

2005-01-01

The objective of this Safety Guide is to provide guidance to the public authorities and others (including consignors, carriers and emergency response authorities) who are responsible for developing and establishing emergency arrangements for dealing effectively and safely with transport accidents involving radioactive material. It may assist those concerned with establishing the capability to respond to such transport emergencies. It provides guidance for those Member States whose involvement with radioactive material is just beginning. It also provides guidance for those Member States that have already developed their radioactive material industries and the attendant emergency plans but that may need to review and improve these plans

The development of severe accident analysis technology

Energy Technology Data Exchange (ETDEWEB)

Kim, Heuy Dong; Cho, Sung Won; Kim, Sang Baek; Park, Jong Hwa; Lee, Kyu Jung; Park, Lae Joon; Hu, Hoh; Hong, Sung Wan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-07-01

The objective of the development of severe accident analysis technology is to understand the severe accident phenomena such as core melt progression and to provide a reliable analytical tool to assess severe accidents in a nuclear power plant. Furthermore, establishment of the accident management strategies for the prevention/mitigation of severe accidents is also the purpose of this research. The study may be categorized into three areas. For the first area, two specific issues were reviewed to identify the further research direction, that is the natural circulation in the reactor coolant system and the fuel-coolant interaction as an in-vessel and an ex-vessel phenomenological study. For the second area, the MELCOR and the CONTAIN codes have been upgraded, and a validation calculation of the MELCOR has been performed for the PHEBUS-B9+ experiment. Finally, the experimental program has been established for the in-vessel and the ex-vessel severe accident phenomena with the in-pile test loop in KMRR and the integral containment test facilities, respectively. (Author).

With power comes responsibility: motorcycle engine power and power-to-weight ratio in relation to accident risk.

Science.gov (United States)

Mattsson, Markus; Summala, Heikki

2010-02-01

Current European legislation allows the EU member states to restrict the maximum power output of motorcycles to 74 kW even though evidence supporting the limit is scarce and has produced mixed results-perhaps because motorcycle performance has been measured by engine displacement, not engine power, in most of the previous studies. This study investigates the relationship of motorcycle engine power and power-to-weight ratio to risk of fatal and nonfatal crashes in Finland. The fatality rate (number of fatal accidents/number of registered motorbikes) for riders of different ages riding bikes belonging to different power and power-to-weight ratio classes was examined using a comprehensive in-depth database. Data on nonfatal accidents were acquired from a Web questionnaire (N = 2708), which also served as a basis for estimating riders' annual mileage. Mileage data allowed the calculation of accident risk per kilometer ridden for bikes differing in power and power-to-weight ratio. The fatality risk per number of registered motorcycles and per kilometer ridden increases both with power and power-to-weight ratio, independently of rider's age. No relationship between performance and risk of a less severe crash was found. The pre-accident speed of the most powerful bikes was 20 km/h or more over the speed limit in a large proportion of the fatal accidents (odds ratio = 4.8 for > 75 kW motorbikes; odds ratio = 6.2 for > 0.3 kW/kg motorbikes). The risk of being involved in a fatal crash is higher among the riders of powerful motorcycles. However, it is not clear whether the results are related to the riding habits of the riders that choose the most powerful bikes available or whether the high risk is due to the properties of the bikes themselves. Therefore, further research is needed before considering legal limits on motorcycle performance.

Community response against the nuclear accident. Confusion in Sweden after the Chernobyl nuclear accident and its features

International Nuclear Information System (INIS)

Sato, Yoshihiro

2014-01-01

The Chernobyl nuclear accident, which occurred in April 1986, became popular in Sweden after two days, and Sweden was hit by a big mess immediately after that. This paper introduces various actions taken in Sweden at that time. The authors analyzed the situation based on the following materials to tell the situation at that time: (1) materials summarized by researchers upon request of the administrative organs of the country, (2) two diaries that were written by Sven Aner, who was a former reporter of a major daily newspaper published after the accident and an activist of antinuclear groups, and Sven Lofvegerg, who handled the accident as a technical officer at Radiation Protection Agency, and (3) newspaper articles at that time. The situations that was revealed after the accident were summarized from the following viewpoints: (1) governmental remarks toward safety standards and effects on residents, and the anxiety of residents, (2) grazing ban on livestock as an important industry and its lifting, (3) correspondence of antinuclear activists, (4) anxiety against the effects of radiation on humans, and counseling on the safety addressed to the Headquarters for Disaster Control, (5) roles of regional radio stations, (6) defects of bureaucracy, (7) criticism against the actions of the Headquarters for Disaster Control, and (8) influence of extreme experts. (A.O.)

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

Energy Technology Data Exchange (ETDEWEB)

Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

2012-09-30

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

Chernobylsk accident (Causes and Consequences)-Part 1

International Nuclear Information System (INIS)

Esteves, D.

1986-07-01

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

Radioactive material (RAM) accident/incident data analysis program

International Nuclear Information System (INIS)

Emerson, E.L.; McClure, J.D.

1985-03-01

This report describes the development of the Radioactive Material Transportation Accident/Incident Data Base (RAM-AIDB), which contains information on the occurrences of transportation accidents and incidents, for radioactive materials (RAM) that are involved in the process of transportation, loading and unloading operation, or temporary storage. These transportation operations are in support of the nuclear fuel cycle for electrical energy generation. This study analyzes in some detail basic accident/incident statistical data, RAM packaging accident response data, and the health effects associated with RAM transport accidents/incidents. This report presents a summary of US RAM transport accident/incident experience for the period 1971 through December 1981. In addition, a sample annual summary of accident/incident experience is presented for the calendar year 1981

41 CFR 102-74.360 - What are the specific accident and fire prevention responsibilities of occupant agencies?

Science.gov (United States)

2010-07-01

... other hanging materials that are made of non-combustible or flame-resistant fabric; (f) Use only... resistant; (g) Cooperate with GSA to develop and maintain fire prevention programs that provide the maximum... accident and fire prevention responsibilities of occupant agencies? 102-74.360 Section 102-74.360 Public...

Radiation doses in accidents at sea-transportation of spent fuel

International Nuclear Information System (INIS)

Appelgren, A.; Bergstroem, U.; Devell, L.

1978-01-01

In order to investigate the consequences of shipping accidents, a release of activity is assumed. This report presents the calculations of individual and collective doses from the two most severe postulated accidents which are given in a special accident analysis. One of the accidents is a ship collision together with fire on-board, the ship is floating after the collision and a certain quantity volatile fission products gives airborne activity. In the other case, it is a fire on-board, the ship will sink and cause a certain leakage to the sea

Evaluation of containment peak pressure and structural response for a large-break loss-of-coolant accident in a VVER-440/213 NPP

Energy Technology Data Exchange (ETDEWEB)

Spencer, B.W.; Sienicki, J.J.; Kulak, R.F.; Pfeiffer, P.A. [Argonne National Lab., IL (United States); Voeroess, L.; Techy, Z. [VEIKI Inst. for Electric Power Research, Budapest (Hungary); Katona, T. [Paks Nuclear Power Plant (Hungary)

1998-07-01

A collaborative effort between US and Hungarian specialists was undertaken to investigate the response of a VVER-440/213-type NPP to a maximum design-basis accident, defined as a guillotine rupture with double-ended flow from the largest pipe (500 mm) in the reactor coolant system. Analyses were performed to evaluate the magnitude of the peak containment pressure and temperature for this event; additional analyses were performed to evaluate the ultimate strength capability of the containment. Separate cases were evaluated assuming 100% effectiveness of the bubbler-condenser pressure suppression system as well as zero effectiveness. The pipe break energy release conditions were evaluated from three sources: (1) FSAR release rate based on Soviet safety calculations, (2) RETRAN-03 analysis and (3) ATHLET analysis. The findings indicated that for 100% bubbler-condenser effectiveness the peak containment pressures were less than the containment design pressure of 0.25 MPa. For the BDBA case of zero effectiveness of the bubbler-condenser system, the peak pressures were less than the calculated containment failure pressure of 0.40 MPa absolute.

Analysis of SBO accident for a swimming pool reactor

International Nuclear Information System (INIS)

Wang Guimin; Li Weiwei; Li Ning; Guo Wenhui

2015-01-01

The RELAP5/MOD3.3 code was adopted to compute the SBO accident condition of a swimming pool reactor. The coolant flow reversal process was calculated, and the influence of parameters of the flow between the core leakage and components on the flow reversal in the SBO accident condition was analyzed. The calculated results show that in the situation the reactor loses all forced flow, the residual heat of the reactor can be removed by the natural circulation flow, and the fuel subassembly will not be damaged. (authors)

Construction safety: Can management prevent all accidents or are workers responsible for their own actions?

International Nuclear Information System (INIS)

Cotten, G.B.; Jenkins, S.L.

1997-01-01

The construction industry has struggled for many years with the answer to the question posed in the title: Can Management Prevent All Accidents or Are Workers Responsible for Their Own Actions? In the litigious society that we live, it has become more important to find someone open-quotes at faultclose quotes for an accident than it is to find out how we can prevent it from ever happening again. Most successful companies subscribe to the theme that open-quotes all accidents can be prevented.close quotes They institute training and qualification programs, safe performance incentives, and culture-change-driven directorates such as the Voluntary Protection Program (VPP); yet we still see construction accidents that result in lost time, and occasionally death, which is extremely costly in the shortsighted measure of money and, in real terms, impact to the worker''s family. Workers need to be properly trained in safety and health protection before they are assigned to a job that may expose them to safety and health hazards. A management committed to improving worker safety and health will bring about significant results in terms of financial savings, improved employee morale, enhanced communities, and increased production. But how can this happen, you say? Reduction in injury and lost workdays are the rewards. A decline in reduction of injuries and lost workdays results in lower workers'' compensation premiums and insurance rates. In 1991, United States workplace injuries and illnesses cost public and private sector employers an estimated $62 billion in workers'' compensation expenditures

Emergency response and nuclear risk governance. Nuclear safety at nuclear power plant accidents; Notfallschutz und Risk Governance. Zur nuklearen Sicherheit bei Kernkraftwerksunfaellen

Energy Technology Data Exchange (ETDEWEB)

Kuhlen, Johannes

2014-07-01

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

Approximate calculation method for integral of mean square value of nonstationary response

International Nuclear Information System (INIS)

Aoki, Shigeru; Fukano, Azusa

2010-01-01

The response of the structure subjected to nonstationary random vibration such as earthquake excitation is nonstationary random vibration. Calculating method for statistical characteristics of such a response is complicated. Mean square value of the response is usually used to evaluate random response. Integral of mean square value of the response corresponds to total energy of the response. In this paper, a simplified calculation method to obtain integral of mean square value of the response is proposed. As input excitation, nonstationary white noise and nonstationary filtered white noise are used. Integrals of mean square value of the response are calculated for various values of parameters. It is found that the proposed method gives exact value of integral of mean square value of the response.

Years of potential life lost due to motorcycle accidents

Directory of Open Access Journals (Sweden)

Flávia Emília Cavalcante Valença Fernandes

2017-05-01

Full Text Available Introduction: Traffic accidents represent a serious public health problem, because they kill approximately 1.24 million persons annually, and leave another 20 to 50 million with non-fatal lesions and traumatisms worldwide. In Brazil, in the year 2011, motorcyclists alone were responsible for one third of these deaths. Therefore, the aim of this study was to estimate the years of potential life lost due to motorcycle accidents, according to sex and age group, and analyze the trend of the indicator for the state of Pernambuco in the period from 2005 to 2014. Methods and Results an ecological study based on data from the System of Information about Mortality was used. The indicator and rate were calculated by using the age limit of 70 years. The linear regression model and Kruskal-Wallis and Mann-Whitney tests were used, at the level of significance of 5% and confidence of 95%. The most affected sex and age-range were men between 20-29 years of age. The rates followed a trend of growth in both sexes, in the young population with the exception of those from 10 to 19 years of age. Conclusions: This context points out the magnitude and precociousness of motorcycle accidents in both sexes and the young population.

Case for integral core-disruptive accident analysis

International Nuclear Information System (INIS)

Luck, L.B.; Bell, C.R.

1985-01-01

Integral analysis is an approach used at the Los Alamos National Laboratory to cope with the broad multiplicity of accident paths and complex phenomena that characterize the transition phase of core-disruptive accident progression in a liquid-metal-cooled fast breeder reactor. The approach is based on the combination of a reference calculation, which is intended to represent a band of similar accident paths, and associated system- and separate-effect studies, which are designed to determine the effect of uncertainties. Results are interpreted in the context of a probabilistic framework. The approach was applied successfully in two studies; illustrations from the Clinch River Breeder Reactor licensing assessment are included

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.

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)

Graphic console for analysis of severe accidents visualization of OEs, NAEs and calculation of the source term for the NPP-LV (CoGrAAS); Consola grafica para analisis de accidentes severos visualizacion de OEs, NAEs y calculo del termino fuente para la CLV (CoGrAAS)

Energy Technology Data Exchange (ETDEWEB)

Sandoval V, S.; Mendoza E, P. R.; Gonzalez C, J. M.; Cecenas F, M. [Instituto Nacional de Electricidad y Energias Limpias, Reforma 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico); Tijerina S, F., E-mail: francisco.tijerina@cfe.gob.mx [CFE, Central Nucleoelectrica Laguna Verde, Carretera Federal Cardel-Nautla Km 42.5, 91476 Municipio Alto Lucero, Veracruz (Mexico)

2016-09-15

In response to the Fukushima Daiichi nuclear power plant accident, the NRC conducted an analysis and issued recommendations to improve the safety of the nuclear reactors. These include strengthening and integrating emergency response capabilities and emphasizing periodic staff training, the performance of simulation exercises. As a tool to observe these recommendations, the Graphic Console was developed for Analysis of Severe Accidents, Visualization of OEs, NAEs and calculation of the source term for nuclear power plant of Laguna Verde (NPP-LV ; CoGrAAS). The CoGrAAS is a computer system that displays in an integrated, graphic and dynamic way the information of a catalog of previously simulated accident scenarios. Has core mimics, vessel, primary containment and safety systems, trend graph of thermodynamic and radiological variables and the emergency procedures (OEs), chronological list of events, windows with detailed information for the dry-well, among others. The use of CoGrAAS allows that staff to understand and become familiar with the thermo-hydraulic progression of actual scenarios that exceed the design basis including those with core damage as severe accidents. The system enables personnel to develop an integral vision of the scenarios during the exercises and drills by observing and analyzing the evolution of the main reactor, core and primary containment variables, the response of emergency systems and the influence of that progression on OEs and the emergency action levels (NAEs). The CoGrAAS allows o observe the radiological variables and obtain the source term, to make the projection of doses, at any time within the scenario evolution. Thus, not only can the phenomenology of severe accidents be analyzed and understood, it is also possible to exercise, verify and evaluate the performance of critical tasks in the application of procedures, guidelines and emergency management plans. (Author)

EAC european accident code. A modular system of computer programs to simulate LMFBR hypothetical accidents

International Nuclear Information System (INIS)

Wider, H.; Cametti, J.; Clusaz, A.; Devos, J.; VanGoethem, G.; Nguyen, H.; Sola, A.

1985-01-01

One aspect of fast reactor safety analysis consists of calculating the strongly coupled system of physical phenomena which contribute to the reactivity balance in hypothetical whole-core accidents: these phenomena are neutronics, fuel behaviour and heat transfer together with coolant thermohydraulics in single- and two-phase flow. Temperature variations in fuel, coolant and neighbouring structures induce, in fact, thermal reactivity feedbacks which are added up and put in the neutronics calculation to predict the neutron flux and the subsequent heat generation in the reactor. At this point a whole-core analysis code is necessary to examine for any hypothetical transient whether the various feedbacks result effectively in a negative balance, which is the basis condition to ensure stability and safety. The European Accident Code (EAC), developed at the Joint Research Centre of the CEC at Ispra (Italy), fulfills this objective. It is a modular informatics structure (quasi 2-D multichannel approach) aimed at collecting stand-alone computer codes of neutronics, fuel pin mechanics and hydrodynamics, developed both in national laboratories and in the JRC itself. EAC makes these modules interact with each other and produces results for these hypothetical accidents in terms of core damage and total energy release. 10 refs

SEVERE ACCIDENT MANAGEMENT STATUS AT Loviisa

International Nuclear Information System (INIS)

Kymalainen, O.; Tuomisto, H.

1997-01-01

Some of the specific design features of IVO's Loviisa Plant, most notably the ice-condenser containment, strongly affect the plant response in a hypothetical core melt accident. They have together with the relatively stringent Finnish regulatory requirements forced IVO to develop a tailor made severe accident management strategy for Loviisa. The low design pressure of the ice-condenser containment complicates the design of the hydrogen management system. On the other hand, the ice-condensers and the water available from them are facilitating factors regarding in-vessel retention of corium by external cooling of reactor pressure vessel. This paper summarizes the Finnish severe accident requirements, IVO's approach to severe accidents, and its application to the Loviisa Plant

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

Immediate medical consequences of nuclear accidents: lessons from Chernobyl

International Nuclear Information System (INIS)

Gale, R.P.

1987-01-01

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

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)

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)

A generic approach for containment success criteria under severe accident loads

International Nuclear Information System (INIS)

Sammataro, R.F.; Solonick, W.R.; Edwards, N.W.

1992-01-01

The U.S. Department of Energy (DOE), Office of New Production Reactors (NP), has identified safety as the foremost design criterion for the Heavy Water New Production Reactor (NPR-HWR). The DOE-NP has issued the Deterministic Severe Accident Criteria (DSACs) to guide the design of the NPR-HWR containment for resistance to severe accidents. The DSAC concept provides for a generic approach for success criteria to predict the threshold of containment failure under severe accident loads. This concept consists of two parts: (1) Problem Statements that are qualitative and quantitative bases for calculating associated loadings and containment response to those loadings, and (2) Success Criteria that specify acceptable containment response measures and limits for each problem statement. This paper is limited to a discussion of a generic approach for containment success criteria. The main elements of these success criteria are expressed in terms of elastic stresses and inelastic strains. Containment performance is based on the best estimate of failure as predicted by either stress or strain, buckling, displacements, or ability to withstand missile perforation. Since these limits are best estimates of failure, no conservatism exists in these success criteria. Rather, conservatism is to be provided in the problem statements, i.e., the quantified severe accident loads. These success criteria are presented on a multi-tiered basis for static pressure and temperature loadings, dynamic loadings, and missiles. Within the static pressure and temperature loadings and the dynamic loadings, the criteria are separated into elastic analysis success criteria and inelastic analysis success criteria. Each of these areas, in turn, defines limits on either the stress or strain measures as well as on measures for buckling and displacements

Radionuclide release calculations for selected severe accident scenarios. Volume 3. PWR, subatmospheric containment design

International Nuclear Information System (INIS)

Denning, R.S.; Gieseke, J.A.; Cybulskis, P.; Lee, K.W.; Jordan, H.; Curtis, L.A.; Kelly, R.F.; Kogan, V.; Schumacher, P.M.

1986-07-01

This report presents results of analyses of the enviromental releases of fission products (source terms) for severe accident scenarios in a pressurized water reactor with a subatmospheric containment design. The analyses were performed to support the Severe Accident Risk Reduction/Risk Rebaselining Program (SARRP) which is being undertaken for the US Nuclear Regulatory Commission by Sandia National Laboratories. In the SARRP program, risk estimates are being generated for a number of reference plant designs. the Surry plant has been used in this study as the reference plant for a subatmospheric design

Database on aircraft accidents

International Nuclear Information System (INIS)

Nishio, Masahide; Koriyama, Tamio

2012-09-01

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

Accident management for severe accidents

International Nuclear Information System (INIS)

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

1988-01-01

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

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

International Nuclear Information System (INIS)

Strip, D.R.

1982-09-01

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

JAERI's activities in JCO accident

International Nuclear Information System (INIS)

2000-09-01

The Japan Atomic Energy Research Institute (JAERI) was actively involved in a variety of technical supports and cooperative activities, such as advice on terminating the criticality condition, contamination checks of the residents and consultation services for the residents, as emergency response actions to the criticality accident at the uranium processing facility operated by the JCO Co. Ltd., which occurred on September 30, 1999. These activities were carried out in collaborative ways by the JAERI staff from the Tokai Research Establishment, Naka Fusion Research Establishment, Oarai Research Establishment, and Headquarter Office in Tokyo. As well, the JAERI was engaged in the post-accident activities such as identification of accident causes, analyses of the criticality accident, and dose assessment of exposed residents, to support the Headquarter for Accident Countermeasures of the Science and Technology Agency (STA), the Accident Investigation Committee and the Health Control Committee of the Nuclear Safety Commission of Japan (NSC). This report compiles the activities, that the JAERI has conducted to date, including the discussions on measures for terminating the criticality condition, evaluation of the fission number, radiation monitoring in the environment, dose assessment, analyses of criticality dynamics. (author)

Effectiveness of electronic stability control on single-vehicle accidents.

Science.gov (United States)

Lyckegaard, Allan; Hels, Tove; Bernhoft, Inger Marie

2015-01-01

This study aims at evaluating the effectiveness of electronic stability control (ESC) on single-vehicle injury accidents while controlling for a number of confounders influencing the accident risk. Using police-registered injury accidents from 2004 to 2011 in Denmark with cars manufactured in the period 1998 to 2011 and the principle of induced exposure, 2 measures of the effectiveness of ESC were calculated: The crude odds ratio and the adjusted odds ratio, the latter by means of logistic regression. The logistic regression controlled for a number of confounding factors, of which the following were significant. For the driver: Age, gender, driving experience, valid driving license, and seat belt use. For the vehicle: Year of registration, weight, and ESC. For the accident surroundings: Visibility, light, and location. Finally, for the road: Speed limit, surface, and section characteristics. The present study calculated the crude odds ratio for ESC-equipped cars of getting in a single-vehicle injury accident as 0.40 (95% confidence interval [CI], 0.34-0.47) and the adjusted odds ratio as 0.69 (95% CI, 0.54-0.88). No difference was found in the effectiveness of ESC across the injury severity categories (slight, severe, and fatal). In line with previous results, this study concludes that ESC reduces the risk for single-vehicle injury accidents by 31% when controlling for various confounding factors related to the driver, the car, and the accident surroundings. Furthermore, it is concluded that it is important to control for human factors (at a minimum age and gender) in analyses where evaluations of this type are performed.

Severe accident tests and development of domestic severe accident system codes

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

Severe accident tests and development of domestic severe accident system codes