WorldWideScience

Sample records for station blackout accident

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  2. Comparative analysis of station blackout accident progression in typical PWR, BWR, and PHWR

    International Nuclear Information System (INIS)

    Park, Soo Young; Ahn, Kwang Il

    2012-01-01

    Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

  3. Analysis of hot leg natural circulation under station blackout severe accident

    International Nuclear Information System (INIS)

    Deng Jian; Cao Xuewu

    2007-01-01

    Under severe accidents, natural circulation flows are important to influence the accident progression and result in a pressurized water reactor (PWR). In a station blackout accident with no recovery of steam generator (SG) auxiliary feedwater (TMLB' severe accident scenario), the hot leg countercurrent natural circulation flow is analyzed by using a severe-accident code, to better understand its potential impacts on the creep-rupture timing among the surge line, the hot leg; and SG tubes. The results show that the natural circulation may delay the failure time of the hot leg. The recirculation ratio and the hot mixing factor are also calculated and discussed. (authors)

  4. Accident Analysis of Chinese CPR1000 in Response to Station Blackout

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Juyoul [FNC Technology Co., Yongin (Korea, Republic of); Cilliers, Anthonie [North-West University, Potchefstroom (South Africa)

    2016-10-15

    Stress tests required evaluation of the consequences of loss of safety functions from any initiating event (e.g., earthquake or flooding) causing loss of electrical power, including station blackout (SBO). The SBO scenario involves a loss of offsite power, failure of the redundant emergency diesel generators, failure of alternate current (AC) power restoration and the eventual degradation of the reactor coolant pump (RCP) seals resulting in a long term loss of coolant. Using PCTRAN/CPR1000, this study analyses the station blackout on a Chinese CPR1000 which is the most representative type reactor in terms of number of reactors, operating period, power capacity and geological distance from Korean Peninsula. Both the physical effects of the accidents as well as the releases of radioisotopes are calculated and discussed. Station blackout simulation was conducted in this study. The resulting effects seen are consistent with other stress test station blackout tests used utilizing licensed simulation codes. An exact comparison is however not possible as the plants on which the simulations was done vary greatly and the limitations of availability to Chinese FSAR. PCTRAN/CPR1000 is an extremely useful simulation package that provides engineers and scientists very accurate feedback to how a nuclear power plant would react as a whole under various plant conditions. It is able to do this extremely fast as well. As a training tool PCTRAN/CPR1000 provides hands-on experience with many of the primary plant operations and develops an intuitive understanding of the plant.

  5. PSB-VVER experimental and analytical investigation of station blackout accident in VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Lipatov, I.A.; Kapustin, A.V.; Nikonov, S.M.; Rovnov, A.A.; Basov, A.V. [Electrogorsk Research and Engineering Centre (EREC), Moscow Region (Russian Federation); Elkin, I.V. [NSI RRC, Kurchatov Institute, Moscow (Russian Federation)

    2007-07-01

    In November 2003, an experiment simulating station blackout accident was carried out in the PSB-VVER integral test facility at the Electrogorsk Research and Engineering Centre (Russia). The purpose of the experiment was to provide missing data for code validation as well as to investigate the VVER thermohydraulics in the blackout conditions. The experiment covers a wide range of phenomena relating not only to transients but also to small break loss-of-coolant accidents. The data gained in the test has been used to assess the RELAP5/MOD3.3 code. In this paper, a special attention has been paid to the code assessment regarding the mixture level and entrainment in steam generator secondary side. The analysis of the recorded transient has shown that the calculation of the heat transfer on the secondary side of steam generators is very sensitive to the steam generator nodalization. (authors)

  6. Thermohydraulic and safety analysis on China advanced research reactor under station blackout accident

    International Nuclear Information System (INIS)

    Tian Wenxi; Qiu Suizheng; Su Guanghui; Jia Dounan; Liu Xingmin; Zhang Jianwei

    2007-01-01

    A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules

  7. Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States); Anders, David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC) system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.

  8. Analysis of station blackout accidents for the Bellefonte pressurized water reactor

    International Nuclear Information System (INIS)

    Gasser, R.D.; Bieniarz, P.P.; Tills, J.L.

    1986-09-01

    An analysis has been performed for the Bellefonte PWR Unit 1 to determine the containment loading and the radiological releases into the environment from a station blackout accident. A number of issues have been addressed in this analysis which include the effects of direct heating on containment loading, and the effects of fission product heating and natural convection on releases from the primary system. The results indicate that direct heating which involves more than about 50% of the core can fail the Bellefonte containment, but natural convection in the RCS may lead to overheating and failure of the primary system piping before core slump, thus, eliminating or mitigating direct heating. Releases from the primary system are significantly increased before vessel breach due to natural circulation and after vessel breach due to reevolution of retained fission products by fission product heating of RCS structures

  9. Plant habitability assessment for Point Lepreau Generating Station during a severe accident resulting from station blackout conditions

    International Nuclear Information System (INIS)

    Mullin, D.

    2015-01-01

    In response to the CNSC Fukushima Action Plan, the CANDU Owners Group (COG) developed a methodology for assessing nuclear power plant habitability under Joint Project 4426 and to determine if any improvement actions are necessary to provide a high degree of assurance that a severe accident can be managed from a human and organizational performance perspective. NB Power has applied the methodology considering a station black-out scenario (representative case), and assessed the effects of non-radiological hazards and radiological hazards in the context of operator dose relative to emergency dose limits. The paper will discuss the overall methodology, findings and recommendations. (author)

  10. Plant habitability assessment for Point Lepreau Generating Station during a severe accident resulting from station blackout conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mullin, D., E-mail: dmullin@nbpower.com [New Brunswick Power Corporation, Point Lepreau Generating Station, Lepreau, NB (Canada)

    2015-07-01

    In response to the CNSC Fukushima Action Plan, the CANDU Owners Group (COG) developed a methodology for assessing nuclear power plant habitability under Joint Project 4426 and to determine if any improvement actions are necessary to provide a high degree of assurance that a severe accident can be managed from a human and organizational performance perspective. NB Power has applied the methodology considering a station black-out scenario (representative case), and assessed the effects of non-radiological hazards and radiological hazards in the context of operator dose relative to emergency dose limits. The paper will discuss the overall methodology, findings and recommendations. (author)

  11. Test study on safety features of station blackout accident for nuclear main pump

    International Nuclear Information System (INIS)

    Liu Xiajie; Wang Dezhong; Zhang Jige; Liu Junsheng; Yang Zhe

    2009-01-01

    The theoretical and experimental studies of reactor coolant pump accidents encountered nation-wide and world-wide were described. To investigate the transient hydrodynamic performance of reactor coolant pump (RCP) during the period of rotational inertia in the station blackout accident, some theoretical and experimental studies were carried out, and the analysis of the test results was presented. The experiment parameters, conditions and test methods were introduced. The flow-rate, rotate speed and vibrations were analyzed emphatically. The quadruplicate polynomial curve equation was used to simulate the flow-rate,rotate speed along with time. The test results indicate that the flow-rate and rotator speed decrease rapidly at the very beginning of cut power and the test results accord with the regulation of safety standard. The vibrant displacement of bearing seat is intensified at the moment of lose power, but after a certain period rotor shaft libration changes. The test and analysis results help to understand the hydrodynamic performance of nuclear primary pump under lost of power accident, and provide the basic reference for safety evaluation. (authors)

  12. Uncertainties in source term estimates for a station blackout accident in a BWR with Mark I containment

    International Nuclear Information System (INIS)

    Lee, M.; Cazzoli, E.; Liu, Y.; Davis, R.; Nourbakhsh, H.; Schmidt, E.; Unwin, S.; Khatib-Rahbar, M.

    1988-01-01

    In this paper, attention is limited to a single accident progression sequence, namly a station blackout accident in a BWR with a Mark I containment building. Identified as an important accident in the draft version of NUREG-1150 a station blackout involves loss of both off-site power and dc power resulting in failure of the diesels to start and in the unavailability of the high pressure injection and core isolation cooling systems. This paper illustrates the calculated uncertainties (Probability Density Functions) associated with the radiological releases into the environment for the nine fission product groups at 10 hours following the initiation of core-concrete interactions. Also shown are the results ofthe STCP base case simulation. 5 refs., 1 fig., 1 tab

  13. Investigation of accident management procedures related to loss of feedwater and station blackout in PSB-VVER integral test facility

    Energy Technology Data Exchange (ETDEWEB)

    Bucalossi, A. [EC JRC, (JRC F.5) PO Box 2, 1755 ZG Petten (Netherlands); Del Nevo, A., E-mail: alessandro.delnevo@enea.it [ENEA, C.R. Brasimone, 40032 Camugnano (Italy); Moretti, F.; D' Auria, F. [GRNSPG, Universita di Pisa, via Diotisalvi 2, 56100 Pisa (Italy); Elkin, I.V.; Melikhov, O.I. [Electrogorsk Research and Engineering Centre, Electrogorsk, Moscow Region (Russian Federation)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Four integral test facility experiments related to VVER-1000 reactor. Black-Right-Pointing-Pointer TH response of the VVER-1000 primary system following total loss of feedwater and station blackout scenarios. Black-Right-Pointing-Pointer Accident management procedures in case of total loss of feedwater and station blackout. Black-Right-Pointing-Pointer Experimental data represent an improvement of existing database for TH code validation. - Abstract: VVER 1000 reactors have some unique and specific features (e.g. large primary and secondary side fluid inventory, horizontal steam generators, core design) that require dedicated experimental and analytical analyses in order to assess the performance of safety systems and the effectiveness of possible accident management strategies. The European Commission funded project 'TACIS 2.03/97', Part A, provided valuable experimental data from the large-scale (1:300) PSB-VVER test facility, investigating accident management procedures in VVER-1000 reactor. A test matrix was developed at University of Pisa (responsible of the project) with the objective of obtaining the experimental data not covered by the OECD VVER validation matrix and with main focus on accident management procedures. Scenarios related to total loss of feed water and station blackout are investigated by means of four experiments accounting for different countermeasures, based on secondary cooling strategies and primary feed and bleed procedures. The transients are analyzed thoroughly focusing on the identification of phenomena that will challenge the code models during the simulations.

  14. Station blackout transient at the Browns Ferry Unit 1 Plant: a severe accident sequence analysis (SASA) program study

    International Nuclear Information System (INIS)

    Schultz, R.R.

    1982-01-01

    Operating plant transients are of great interest for many reasons, not the least of which is the potential for a mild transient to degenerate to a severe transient yielding core damage. Using the Browns Ferry (BF) Unit-1 plant as a basis of study, the station blackout sequence was investigated by the Severe Accident Sequence Analysis (SASA) Program in support of the Nuclear Regulatory Commission's Unresolved Safety Issue A-44: Station Blackout. A station blackout transient occurs when the plant's AC power from a comemrcial power grid is lost and cannot be restored by the diesel generators. Under normal operating conditions, f a loss of offsite power (LOSP) occurs [i.e., a complete severance of the BF plants from the Tennessee Valley Authority (TVA) power grid], the eight diesel generators at the three BF units would quickly start and power the emergency AC buses. Of the eight diesel generators, only six are needed to safely shut down all three units. Examination of BF-specific data show that LOSP frequency is low at Unit 1. The station blackout frequency is even lower (5.7 x 10 - 4 events per year) and hinges on whether the diesel generators start. The frequency of diesel generator failure is dictated in large measure by the emergency equipment cooling water (EECW) system that cools the diesel generators

  15. Development and qualification of a thermal-hydraulic nodalization for modeling station blackout accident in PSB-VVER test facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-15

    Highlights: • A thermal-hydraulic nodalization for PSB-VVER test facility has been developed. • Station blackout accident is modeled with the developed nodalization in MELCOR code. • The developed nodalization is qualified at both steady state and transient levels. • MELCOR predictions are qualitatively and quantitatively in acceptable range. • Fast Fourier Transform Base Method is used to quantify accuracy of code predictions. - Abstract: This paper deals with the development of a qualified thermal-hydraulic nodalization for modeling Station Black-Out (SBO) accident in PSB-VVER Integral Test Facility (ITF). This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr nuclear power plant. In this regard, a nodalization has been developed for thermal-hydraulic modeling of the PSB-VVER ITF by MELCOR integrated code. The nodalization is qualitatively and quantitatively qualified at both steady-state and transient levels. The accuracy of the MELCOR predictions is quantified in the transient level using the Fast Fourier Transform Base Method (FFTBM). FFTBM provides an integral representation for quantification of the code accuracy in the frequency domain. It was observed that MELCOR predictions are qualitatively and quantitatively in the acceptable range. In addition, the influence of different nodalizations on MELCOR predictions was evaluated and quantified using FFTBM by developing 8 sensitivity cases with different numbers of control volumes and heat structures in the core region and steam generator U-tubes. The most appropriate case, which provided results with minimum deviations from the experimental data, was then considered as the qualified nodalization for analysis of SBO accident in the PSB-VVER ITF. This qualified nodalization can be used for modeling of VVER-1000 nuclear power plants when performing SBO accident analysis by MELCOR code.

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

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2013-01-01

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

  17. The safety assessment of OPR-1000 nuclear power plant for station blackout accident applying the combined deterministic and probabilistic procedure

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Gu, E-mail: littlewing@kins.re.kr [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2014-08-15

    Highlights: • The combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. • The safety assessment of OPR-1000 nuclear power plant for SBO accident is performed by applying the CDPP. • By estimating the offsite power restoration time appropriately, the SBO risk is reevaluated. • It is concluded that the CDPP is applicable to safety assessment of BDBAs without significant erosion of the safety margin. - Abstract: Station blackout (SBO) is a typical beyond design basis accident (BDBA) and significant contributor to overall plant risk. The risk analysis of SBO could be important basis of rulemaking, accident mitigation strategy, etc. Recently, studies on the integrated approach of deterministic and probabilistic method for nuclear safety in nuclear power plants have been done, and among them, the combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. In the CDPP, the conditional exceedance probability obtained by the best estimate plus uncertainty method acts as go-between deterministic and probabilistic safety assessments, resulting in more reliable values of core damage frequency and conditional core damage probability. In this study, the safety assessment of OPR-1000 nuclear power plant for SBO accident was performed by applying the CDPP. It was confirmed that the SBO risk should be reevaluated by eliminating excessive conservatism in existing probabilistic safety assessment to meet the targeted core damage frequency and conditional core damage probability. By estimating the offsite power restoration time appropriately, the SBO risk was reevaluated, and it was finally confirmed that current OPR-1000 system lies in the acceptable risk against the SBO. In addition, it is concluded that the CDPP is applicable to safety assessment of BDBAs in nuclear power plants without significant erosion of the safety margin.

  18. Station blackout calculations for Peach Bottom

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1985-01-01

    A calculational procedure for the Station Blackout Severe Accident Sequence at Browns Ferry Unit One has been repeated with plant-specific application to one of the Peach Bottom Units. The only changes required in code input are with regard to the primary continment concrete, the existence of sprays in the secondary containment, and the size of the refueling bay. Combustible gas mole fractions in the secondary containment of each plant during the accident sequence are determined. It is demonstrated why the current state-of-the-art corium/concrete interaction code is inadequate for application to the study of Severe Accident Sequences in plants with the BWR MK I or MK II containment design

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

    International Nuclear Information System (INIS)

    Valle Cepero, Reinaldo

    2004-01-01

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

  20. Pressurized-water-reactor station blackout

    International Nuclear Information System (INIS)

    Dobbe, C.A.

    1983-01-01

    The purpose of the Severe Accident Sequence Analysis (SASA) Program was to investigate accident scenarios beyond the design basis. The primary objective of SASA was to analyze nuclear plant transients that could lead to partial or total core melt and evaluate potential mitigating actions. The following summarizes the pressurized water reactor (PWR) SASA effort at the Idaho National Engineering Laboratory (INEL). The INEL is presently evaluating Unresolved Safety Issue A-44 - Station Blackout from initiation of the transient to core uncovery. The balance of the analysis from core uncovery until fission product release is being performed at Sandia National Laboratory (SNL). The current analyses involve the Bellefonte Nuclear Steam Supply System (NSSS), a Babcock and Wilcox (B and W) 205 Fuel Assembly (205-FA) raised loop design to be operated by the Tennessee Valley Authority

  1. Utilities respond to nuclear station blackout rule

    International Nuclear Information System (INIS)

    Rubin, A.M.; Beasley, B.; Tenera, L.P.

    1990-01-01

    The authors discuss how nuclear plants in the United States have taken actions to respond to the NRC Station Blackout Rule, 10CFR50.63. The rule requires that each light water cooled nuclear power plant licensed to operate must be able to withstand for a specified duration and recover from a station blackout. Station blackout is defined as the complete loss of a-c power to the essential and non-essential switch-gear buses in a nuclear power plant. A station blackout results from the loss of all off-site power as well as the on-site emergency a-c power system. There are two basic approaches to meeting the station blackout rule. One is to cope with a station blackout independent of a-c power. Coping, as it is called, means the ability of a plant to achieve and maintain a safe shutdown condition. The second approach is to provide an alternate a-c power source (AAC)

  2. Station blackout at nuclear power plants: Radiological implications for nuclear war

    International Nuclear Information System (INIS)

    Shapiro, C.S.

    1986-12-01

    Recent work on station blackout is reviewed its radiological implications for a nuclear war scenario is explored. The major conclusion is that the effects of radiation from many nuclear weapon detonations in a nuclear war would swamp those from possible reactor accidents that result from station blackout

  3. Design Provisions for Withstanding Station Blackout at Nuclear Power Plants

    International Nuclear Information System (INIS)

    2015-08-01

    International operating experience has shown that the loss of off-site power supply concurrent with a turbine trip and unavailability of the standby alternating current power system is a credible event. Lessons learned from the past and recent station blackout events, as well as the analysis of the safety margins performed as part of the ‘stress tests’ conducted on European nuclear power plants in response to the Fukushima Daiichi accident, have identified the station blackout event as a limiting case for most nuclear power plants. The magnitude 9.0 earthquake and consequential tsunami which occurred in Fukushima, Japan, in March 2011, led to a common cause failure of on-site alternating current electrical power supply systems at the Fukushima Daiichi nuclear power plant as well as the off-site power grid. In addition, the resultant flooding caused the loss of direct current power supply, which further exacerbated an already critical situation at the plant. The loss of electrical power resulted in the meltdown of the core in three reactors on the site and severely restricted heat removal from the spent fuel pools for an extended period of time. The plant was left without essential instrumentation and controls, and this made accident management very challenging for the plant operators. The operators attempted to bring and maintain the reactors in a safe state without information on the vital plant parameters until the power supply was eventually restored after several days. Although the Fukushima Daiichi accident progressed well beyond the expected consequences of a station blackout, which is the complete loss of all alternating current power supplies, many of the lessons learned from the accident are valid. A failure of the plant power supply system such as the one that occurred at Fukushima Daiichi represents a design extension condition that requires management with predesigned contingency planning and operator training. The extended loss of all power at a

  4. Evaluation of Station Blackout accidents at nuclear power plants. Technical findings related to Unresolved Safety Issue A-44. Draft report for comment

    International Nuclear Information System (INIS)

    Baranowsky, P.W.

    1985-05-01

    ''Station Blackout,'' which is the complete loss of alternating current (ac) electrical power in a nuclear power plant, has been designated as Unresolved Safety Issue A-44. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on ac power, the consequences of a station blackout could be severe. This report documents the findings of technical studies performed as part of the program to resolve this issue. The important factors analyzed include: the frequency of loss of offsite power; the probability that emergency or onsite ac power supplies would be unavailable; the capability and reliability of decay heat removal systems independent of ac power; and the likelihood that offsite power would be restored before systems that cannot operate for extended periods without ac power fail, thus resulting in core damage. This report also addresses effects of different designs, locations, and operational features on the estimated frequency of core damage resulting from station blackout events

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-03-01

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

  6. Station blackout and public confidence: a cautionary tale

    International Nuclear Information System (INIS)

    Cave, L.

    1990-01-01

    The recent ''station blackout'' (ie loss of on-site and off-site AC power) incidents at the Vogtle PWR in the US and Hinkley Point B AGR in the Uk have led to further public concern about the safety of nuclear power, even though in each case the actual increase in the chance of an accident leading to a release of radioactivity to the environment was negligible. The industry may be wise to invest precautionary measures to reduce the frequency of such incidents and to increase public confidence. (author)

  7. Estimation of station blackout frequency in FBTR

    International Nuclear Information System (INIS)

    Senthil Kumar, C.; John Arul, A.; Anandapadmanaban, B.; Marimuthu, S.; Singh, Om Pal

    2002-01-01

    Full text: In this paper, station blackout (SBO) frequency is computed as a function of blackout duration based on available data in FBTR. The frequency of loss of offsite power (LOSP) at FBTR is found to be 5.2/year. The data on the LOSP and failure data on feeders and transformers at FBTR are used to arrive at the frequency of LOSP as a function of down time for single feeder and double feeder cases. The non-recovery probability of offsite power failure with time is well represented by exponential distribution for times less than 2 h and Weibull distribution beyond 2 h. The unavailability of onsite emergency power supply was evaluated using Markov method and fault tree method and is 1.0E-2 and 3.34E-3 respectively. Using the above data and the non-recovery probability of DG, frequency of SBO at FBTR with single feeder and double feeder cases, for different time durations were evaluated. It is found that the frequency of station blackout at FBTR with double feeder computed using Markov method is ∼10 -4 /yr for 11 h and 10 -5 /yr for 19 h duration. With one feeder out of service the SBO frequency is more by a factor 5. Sensitivity study done with respect to DG repair time and common cause failure indicate that the results in the magnitude of SBO could be uncertain by a factor of ten. The uncertainty is less for shorter SBO duration and more for longer SBO duration

  8. Extended Station Blackout Analyses of an APR1400 with MARS-KS

    International Nuclear Information System (INIS)

    Kim, WoongBae; Jang, HyungWook; Oh, Seungjong; Lee, Sangyong

    2016-01-01

    The Fukushima Dai-ichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electricity required for essential systems during a station blackout is provided from the emergency backup batteries installed at the nuclear power plant. In South Korea, in the event of an extended station blackout, the life of these emergency backup batteries has recently been extended from 8 hours to 24 hours at Shin-Kori 5, 6 and APR1400 for design certification. For a battery life of 24 hours, available safety means system, equipment and procedures are studied and analyzed in their ability to cope with an extended station blackout. A sensitivity study of reactor coolant pump seal leakage is performed to verify how different seal leakages could affect the system. For simulating of extended station blackout scenarios, the best estimate MARS-KS was used. In this paper, an APR1400 RELAP5 input deck was developed for station blackout scenario to analyze operation strategy by manually depressurizing the reactor coolant system through the steam generator's secondary side. Additionally, a sensitivity study was performed on reactor coolant pump seal leakage

  9. Extended station blackout analyses of an APR1400 with MARS-KS

    Directory of Open Access Journals (Sweden)

    Kim Woongbae

    2016-01-01

    Full Text Available The Fukushima Daiichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electric energy required for essential systems during a station blackout is provided from emergency backup batteries installed at the nuclear power plant. In South Korea, in the event of an extended station blackout, the life of these emergency backup batteries has recently been extended from 8 hours to 24 hours at Shin-Kori 5, 6, and APR1400 for design certification. For a battery life of 24 hours, available safety means system, equipment and procedures are studied and analyzed in their ability to cope with an extended station blackout. A sensitivity study of reactor coolant pump seal leakage is performed to verify how different seal leakages could affect the system. For simulating extended station blackout scenarios, the best estimate MARS-KS computer code was used. In this paper, an APR1400 RELAP5 input deck was developed for station blackout scenario to analyze operation strategy by manually depressurizing the reactor coolant system through the steam generator's secondary side. Additionally, a sensitivity study on reactor coolant pump seal leakage was carried out.

  10. Extended Station Blackout Analyses of an APR1400 with MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, WoongBae; Jang, HyungWook; Oh, Seungjong; Lee, Sangyong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-10-15

    The Fukushima Dai-ichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electricity required for essential systems during a station blackout is provided from the emergency backup batteries installed at the nuclear power plant. In South Korea, in the event of an extended station blackout, the life of these emergency backup batteries has recently been extended from 8 hours to 24 hours at Shin-Kori 5, 6 and APR1400 for design certification. For a battery life of 24 hours, available safety means system, equipment and procedures are studied and analyzed in their ability to cope with an extended station blackout. A sensitivity study of reactor coolant pump seal leakage is performed to verify how different seal leakages could affect the system. For simulating of extended station blackout scenarios, the best estimate MARS-KS was used. In this paper, an APR1400 RELAP5 input deck was developed for station blackout scenario to analyze operation strategy by manually depressurizing the reactor coolant system through the steam generator's secondary side. Additionally, a sensitivity study was performed on reactor coolant pump seal leakage.

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

    International Nuclear Information System (INIS)

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

    2011-11-01

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

  12. Analysis of a station blackout transient at the Kori units 3/4

    International Nuclear Information System (INIS)

    Bang, Young Seok; Kim, Hho Jung

    1992-01-01

    A transient analysis of station blackout accident is performed to evaluate the plant specific capability to cope with the accident at the Kori Units 3/4. The RELAP5/MOD3/5m5 code and full three loop modelling scheme are used in the calculation. The leak flow from reactor coolant system due to a failure of reactor coolant pump seal following the accident is assumed to be 25 gpm and the turbine driven aux feedwater unavailable. As a result, it is found that no core uncovery occurs in the plant until 7100 sec following a station blackout, the steam generator has a decay heat removal capability until 3100 sec, and the natural circulation over the reactor coolant loop until the complete loop seal voiding are observed. And the Nuclear Plant Analyzer is used and found to be effective in improving the phenomenological understanding on the accident

  13. Design Provisions for Station Blackout at Nuclear Power Plants

    International Nuclear Information System (INIS)

    Duchac, Alexander

    2015-01-01

    A station blackout (SBO) is generally known as 'a plant condition with complete loss of all alternating current (AC) power from off-site sources, from the main generator and from standby AC power sources important to safety to the essential and nonessential switchgear buses. Direct current (DC) power supplies and un-interruptible AC power supplies may be available as long as batteries can supply the loads. Alternate AC power supplies are available'. A draft Safety Guide DS 430 'Design of Electrical Power Systems for Nuclear Power Plants' provides recommendations regarding the implementation of Specific Safety Requirements: Design: Requirement 68 for emergency power systems. The Safety Guide outlines several design measures which are possible as a means of increasing the capability of the electrical power systems to cope with a station blackout, without providing detailed implementation guidance. A committee of international experts and advisors from numerous countries is currently working on an IAEA Technical Document (TECDOC) whose objective is to provide a common international technical basis from which the various criteria for SBO events need to be established, to support operation under design basis and design extension conditions (DEC) at nuclear power plants, to document in a comprehensive manner, all relevant aspects of SBO events at NPPs, and to outline critical issues which reflect the lessons learned from the Fukushima Dai-ichi accident. This paper discusses the commonly encountered difficulties associated with establishing the SBO criteria, shares the best practices, and current strategies used in the design and implementation of SBO provisions and outline the structure of the IAEA's SBO TECDOC under development. (author)

  14. Main considerations for modelling a station blackout scenario with trace

    International Nuclear Information System (INIS)

    Querol, Andrea; Turégano, Jara; Lorduy, María; Gallardo, Sergio; Verdú, Gumersindo

    2017-01-01

    In the nuclear safety field, the thermal hydraulic phenomena that take place during an accident in a nuclear power plant is of special importance. One of the most studied accidents is the Station BlackOut (SBO). The aim of the present work is the analysis of the PKL integral test facility nodalization using the thermal-hydraulic code TRACE5 to reproduce a SBO accidental scenario. The PKL facility reproduces the main components of the primary and secondary systems of its reference nuclear power plant (Philippsburg II). The results obtained with different nodalization have been compared: 3D vessel vs 1D vessel, Steam Generator (SG) modelling using PIPE or TEE components and pressurizer modelling with PIPE or PRIZER components. Both vessel nodalization (1D vessel and 3D vessel) reproduce the physical phenomena of the experiment. However, there are significant discrepancies between them. The appropriate modelling of the SG is also relevant in the results. Regarding the other nodalization (PIPE or TEE components for SG and PIPE or PRIZER components for pressurizer), do not produce relevant differences in the results. (author)

  15. Main considerations for modelling a station blackout scenario with trace

    Energy Technology Data Exchange (ETDEWEB)

    Querol, Andrea; Turégano, Jara; Lorduy, María; Gallardo, Sergio; Verdú, Gumersindo, E-mail: anquevi@upv.es, E-mail: jaturna@upv.es, E-mail: maloral@upv.es, E-mail: sergalbe@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Instituto Universitario de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), Universitat Politècnica de València (Spain)

    2017-07-01

    In the nuclear safety field, the thermal hydraulic phenomena that take place during an accident in a nuclear power plant is of special importance. One of the most studied accidents is the Station BlackOut (SBO). The aim of the present work is the analysis of the PKL integral test facility nodalization using the thermal-hydraulic code TRACE5 to reproduce a SBO accidental scenario. The PKL facility reproduces the main components of the primary and secondary systems of its reference nuclear power plant (Philippsburg II). The results obtained with different nodalization have been compared: 3D vessel vs 1D vessel, Steam Generator (SG) modelling using PIPE or TEE components and pressurizer modelling with PIPE or PRIZER components. Both vessel nodalization (1D vessel and 3D vessel) reproduce the physical phenomena of the experiment. However, there are significant discrepancies between them. The appropriate modelling of the SG is also relevant in the results. Regarding the other nodalization (PIPE or TEE components for SG and PIPE or PRIZER components for pressurizer), do not produce relevant differences in the results. (author)

  16. Reactor coolant pump shaft seal stability during station blackout

    International Nuclear Information System (INIS)

    Rhodes, D.B.; Hill, R.C.; Wensel, R.G.

    1987-05-01

    Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries

  17. Reactor coolant pump shaft seal stability during station blackout

    Energy Technology Data Exchange (ETDEWEB)

    Rhodes, D B; Hill, R C; Wensel, R G

    1987-05-01

    Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries.

  18. Investigation of a Station Blackout Scenario with the ATLAS Test

    International Nuclear Information System (INIS)

    Kim, Yeon Sik; Yu, Xin Guo; Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok; Min, Kyeong Ho; Choi, Nam Hyeon; Kim, Bok Deuk; Park, Jong Gook; Choi, Ki Yong

    2012-01-01

    KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), for accident simulations pertaining to the OPR1000 (Optimized Power Reactor, 1000MWe) and the APR1400 (Advanced Power Reactor, 1400MWe) which are in operation and under construction in Korea, respectively. After the Fukushima accidents due to the combination of an earthquake followed by a tsunami in east Japan on March 11, 2011, the concept of boundary between the design basis and beyond-design basis accidents became obscure. One scenario is the station blackout (SBO), which is defined as 'the loss of all alternating current (AC) power in a nuclear power plant' by the USNRC 10CFR50 Section 50.63, which has adopted a new safety regulation for the SBO in June of 1988. In any case the SBO that occurred in Fukushima seemed to go beyond the definition of the current SBO scenario. In the mean time, numerous researches has been conducted on the safety concern of the SBO for existing and advanced nuclear power plants worldwide. From the internal review of an SBO scenario, it was concluded that the understanding of the thermo-hydraulic phenomena occurred within the reactor coolant system is a prerequisite although seemed to be quite a simple sequence of events. This was the motivation of an SBO test using the ATLAS facility. For the understanding of the physical phenomena within the primary system, an SBO was assumed with simple intial and boundary conditions, e.g. start of an SBO at time zero, no diesel and AC powers, no auxiliary feedwater pumps (motor-driven and turbine driven) etc. In this paper, overview of the SBO test results was described including a result of analytical calculations simulating the SBO test using the MARS code

  19. Extension of station blackout coping capability and implications on nuclear safety

    Energy Technology Data Exchange (ETDEWEB)

    Volkanovski, Andrija, E-mail: andrija.volkanovski@ijs.si [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Prošek, Andrej [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2013-02-15

    Highlights: ► Modifications enhancing station blackout coping capability are analyzed. ► Analysis is done with deterministic and probabilistic safety analysis methods. ► The core heat up is delayed for at least the extension time interval. ► Auxiliary feedwater system delays core heat up even in presence of pumps seal leakage. ► Extension of station blackout coping capability decreases core damage frequency. -- Abstract: The safety of the nuclear power plant depends on the availability of the continuous and reliable sources of electrical energy during all modes of operation of the plant. The station blackout corresponds to a total loss of all alternate current (AC) power as a result of complete failure of both offsite and on-site AC power sources. The electricity for the essential systems during station blackout is provided from the batteries installed in the nuclear power plant. The results of the probabilistic safety assessment show that station blackout is one of the main and frequently the dominant contributor to the core damage frequency. The accident in Fukushima Daiichi nuclear power plants demonstrates the vulnerability of the currently operating nuclear power plants during the extended station blackout events. The objective of this paper is, considering the identified importance of the station blackout initiating event, to assess the implications of the strengthening of the SBO mitigation capability on safety of the NPP. The assessment is done with state-of-art deterministic and probabilistic methods and tolls with application on reference models of nuclear power plants. The U.S. NRC Station Blackout Rule describes procedure for the assessment of the size and capacity of the batteries in the nuclear power plant. The description of the procedure with the application on the reference plant and identified deficiencies in the procedure is presented. The safety analysis is done on reference model of the nuclear power plant. Obtained results show large

  20. Extension of station blackout coping capability and implications on nuclear safety

    International Nuclear Information System (INIS)

    Volkanovski, Andrija; Prošek, Andrej

    2013-01-01

    Highlights: ► Modifications enhancing station blackout coping capability are analyzed. ► Analysis is done with deterministic and probabilistic safety analysis methods. ► The core heat up is delayed for at least the extension time interval. ► Auxiliary feedwater system delays core heat up even in presence of pumps seal leakage. ► Extension of station blackout coping capability decreases core damage frequency. -- Abstract: The safety of the nuclear power plant depends on the availability of the continuous and reliable sources of electrical energy during all modes of operation of the plant. The station blackout corresponds to a total loss of all alternate current (AC) power as a result of complete failure of both offsite and on-site AC power sources. The electricity for the essential systems during station blackout is provided from the batteries installed in the nuclear power plant. The results of the probabilistic safety assessment show that station blackout is one of the main and frequently the dominant contributor to the core damage frequency. The accident in Fukushima Daiichi nuclear power plants demonstrates the vulnerability of the currently operating nuclear power plants during the extended station blackout events. The objective of this paper is, considering the identified importance of the station blackout initiating event, to assess the implications of the strengthening of the SBO mitigation capability on safety of the NPP. The assessment is done with state-of-art deterministic and probabilistic methods and tolls with application on reference models of nuclear power plants. The U.S. NRC Station Blackout Rule describes procedure for the assessment of the size and capacity of the batteries in the nuclear power plant. The description of the procedure with the application on the reference plant and identified deficiencies in the procedure is presented. The safety analysis is done on reference model of the nuclear power plant. Obtained results show large

  1. Application of a steam injector for passive emergency core cooling during a station blackout

    International Nuclear Information System (INIS)

    Heinze, D.; Behnke, L.; Schulenberg, T.

    2012-01-01

    One of the basic protection targets of reactor safety is the safe heat removal during normal operation but also following shut-down. Since the reactor accident in Fukushima an optimization of the plant robustness in case of beyond-design accident is performed. Special attention is given to the increase of time available for starting appropriate measures for emergency core cooling in case of a station blackout. The state-of the art in engineering and research is presented. Investigations on the applicability of a steam injector for passive emergency core cooling during a station blackout in BWR-type reactors have progressed, experiments on dynamic behavior of the injector are described. A precise design with respect to the thermal hydraulic boundary conditions has been performed.

  2. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    Science.gov (United States)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  3. Reactor coolant pump shaft seal behavior during station blackout

    International Nuclear Information System (INIS)

    Kittmer, C.A.; Wensel, R.G.; Rhodes, D.B.; Metcalfe, R.; Cotnam, B.M.; Gentili, H.; Mings, W.J.

    1985-04-01

    A testing program designed to provide fundamental information pertaining to the behavior of reactor coolant pump (RCP) shaft seals during a postulated nuclear power plant station blackout has been completed. One seal assembly, utilizing both hydrodynamic and hydrostatic types of seals, was modeled and tested. Extrusion tests were conducted to determine if seal materials could withstand predicted temperatures and pressures. A taper-face seal model was tested for seal stability under conditions when leaking water flashes to steam across the seal face. Test information was then used as the basis for a station blackout analysis. Test results indicate a potential problem with an elastomer material used for O-rings by a pump vendor; that vendor is considering a change in material specification. Test results also indicate a need for further research on the generic issue of RCP seal integrity and its possible consideration for designation as an unresolved safety issue

  4. Station blackout core damage frequency in an advanced nuclear reactor

    International Nuclear Information System (INIS)

    Carvalho, Luiz Sergio de

    2004-01-01

    Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

  5. Case study on the use of PSA methods: Station blackout risk at Millstone Unit 3

    International Nuclear Information System (INIS)

    1991-04-01

    In Westinghouse pressurized water reactors, severe accidents sequences resulting from station blackout have been recognized to be significant contributors to risk of core damage and public consequences. To properly quantify the risk of station blackout it is necessary to consider all possible types of core damage scenarios. Having obtained an accurate representation of the types of core damage scenarios involved specific areas of vulnerability can be pinpointed for further improvement. In performing this analysis it was decided to use time dependent probabilistic safety assessment method to provide a more realistic treatment of time dependent failure and recovery. Overview of the analysis, calculation procedures and methods, interpretation of the results are discussed. Peer review process is described. 13 refs, 19 figs

  6. The AP1000R nuclear power plant innovative features for extended station blackout mitigation

    International Nuclear Information System (INIS)

    Vereb, F.; Winters, J.; Schulz, T.; Cummins, E.; Oriani, L.

    2012-01-01

    Station Blackout (SBO) is defined as 'a condition wherein a nuclear power plant sustains a loss of all offsite electric power system concurrent with turbine trip and unavailability of all onsite emergency alternating current (AC) power system. Station blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by alternate AC sources as defined in this section, nor does it assume a concurrent single failure or design basis accident...' in accordance with Reference 1. In this paper, the innovative features of the AP1000 plant design are described with their operation in the scenario of an extended station blackout event. General operation of the passive safety systems are described as well as the unique features which allow the AP1000 plant to cope for at least 7 days during station blackout. Points of emphasis will include: - Passive safety system operation during SBO - 'Fail-safe' nature of key passive safety system valves; automatically places the valve in a conservatively safe alignment even in case of multiple failures in all power supply systems, including normal AC and battery backup - Passive Spent Fuel Pool cooling and makeup water supply during SBO - Robustness of AP1000 plant due to the location of key systems, structures and components required for Safe Shutdown - Diverse means of supplying makeup water to the Passive Containment Cooling System (PCS) and the Spent Fuel Pool (SFP) through use of an engineered, safety-related piping interface and portable equipment, as well as with permanently installed onsite ancillary equipment. (authors)

  7. The AP1000{sup R} nuclear power plant innovative features for extended station blackout mitigation

    Energy Technology Data Exchange (ETDEWEB)

    Vereb, F.; Winters, J.; Schulz, T.; Cummins, E.; Oriani, L. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    Station Blackout (SBO) is defined as 'a condition wherein a nuclear power plant sustains a loss of all offsite electric power system concurrent with turbine trip and unavailability of all onsite emergency alternating current (AC) power system. Station blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by alternate AC sources as defined in this section, nor does it assume a concurrent single failure or design basis accident...' in accordance with Reference 1. In this paper, the innovative features of the AP1000 plant design are described with their operation in the scenario of an extended station blackout event. General operation of the passive safety systems are described as well as the unique features which allow the AP1000 plant to cope for at least 7 days during station blackout. Points of emphasis will include: - Passive safety system operation during SBO - 'Fail-safe' nature of key passive safety system valves; automatically places the valve in a conservatively safe alignment even in case of multiple failures in all power supply systems, including normal AC and battery backup - Passive Spent Fuel Pool cooling and makeup water supply during SBO - Robustness of AP1000 plant due to the location of key systems, structures and components required for Safe Shutdown - Diverse means of supplying makeup water to the Passive Containment Cooling System (PCS) and the Spent Fuel Pool (SFP) through use of an engineered, safety-related piping interface and portable equipment, as well as with permanently installed onsite ancillary equipment. (authors)

  8. Comparison of the MAAP4 code with the station blackout simulation in the IIST facility

    International Nuclear Information System (INIS)

    Robert E Henry; Christopher E Henry; Chan Y Paik; George M Hauser

    2005-01-01

    Full text of publication follows: The Modular Accident Analysis Program (MAAP) is an integral system model to assess challenges to the reactor core, Reactor Coolant System (RCS) and containment for accident conditions. MAAP4 is the current version used by the MAAP Users Group to assess the responses to a spectrum of accident conditions. Benchmarking of the MAAP code with integral system experiments has been a continuing effort by MAAP developers and users. Several of these have been configured into dynamic benchmarks and are included in Volume III (Benchmarking) of the MAAP4 Users Manual (EPRI, 2004). One such integral experiment is the INER integral system test (IIST) constructed at the Institute of Nuclear Energy Research in Taiwan. This experimental facility is a reduced height, reduced pressure representation of a 3-loop PWR and has been used to examine several different types of accident sequences. One of these is a station blackout simulation with loss of auxiliary feedwater at the time that the transient is initiated. This is an important integral experiment to be compared with the MAAP4 code models. A parameter file (those values representing the system design and boundary experimental conditions) has been developed for the IIST facility and an input deck has been configured to represent a station blackout sequence with instantaneous loss of auxiliary feedwater. Of importance in this benchmark is (a) the rate at which the secondary side inventory is depleted, (b) the depletion of water within the reactor pressure vessel and (c) the time at which the top of the reactor core is uncovered. Comparisons have been made with these three different intervals and there is good agreement between the timing of these events for the MAAP4 benchmark. This is important since this reference sequence represents a set of boundary conditions that is continually with subsequent analyses being perturbations on this type of accident sequence. The good agreement between MAAP4 and

  9. Safety aspects of station blackout at nuclear power plants

    International Nuclear Information System (INIS)

    1985-03-01

    The principal focus of this report is on existing light water reactor nuclear power plants. However, many of the considerations discussed herein can be equally applied to new plants, i.e. those not yet in construction. This report is organized to provide a description of design and procedural factors which safety assessments and reviews of operating experience have shown to be important. These are divided into the off-site power system, the on-site AC power systems and alternate (or nearby) sources of power. The latter may be used in the unlikely event that both normal off-site and on-site sources fail. It must be emphasized that first priority should be placed on designing and maintaining high reliability of both the off-site and on-site AC power systems. This basic concept also applies to the capabilities for restoring power sources which failed and making use of all available alternative and nearby power sources during an emergency, to restore AC power in a prompt manner. Discussions on these aspects are provided in chapters 2 and 3 of this report. Because the expected event frequency and associated confidence in such estimations of station blackout are uncertain, preparations should be made to deal with a station blackout. The nature of those preparations, whether they be optimizing emergency procedures to use existing equipment, modifying this equipment to enhance capabilities, or adding new components or systems to cope with station blackout, must be made in light of plant-specific assessments and regulatory safety philosophies/requirements. Discussions on these matters are provided in chapter 4. General and specific conclusions and recommendations are provided in chapter 5. Appendix A provides a description of several case studies on station blackout and loss of off-site power. Abstracts of papers and presentations are provided in Appendix B with authors and affiliations identified to facilitate personal contact. The References and Bibliography contain a

  10. Station blackout with reactor coolant pump seal leakage

    International Nuclear Information System (INIS)

    Evinay, A.

    1993-01-01

    The U.S. Nuclear Regulatory Commission (NRC) amended its regulations in 10CFR50 with the addition of a new section, 50.63, open-quotes Loss of All Alternating Current Power.close quotes The objective of these requirements is to ensure that all nuclear plants have the capability to withstand a station blackout (SBO) and maintain adequate reactor core cooling and containment integrity for a specified period of time. The NRC also issued Regulatory Guide (RG) 1.155, open-quotes Station Blackout,close quotes to provide guidance for meeting the requirements of 10CFR50.63. Concurrent with RG-1.155, the Nuclear Utility Management and Resources Council (NUMARC) has developed NUMARC 87-00 to address SBO-coping duration and capabilities at light water reactors. Licensees are required to submit a topical report based on NUMARC 87-00 guidelines, to demonstrate compliance with the SBO rule. One of the key compliance criteria is the ability of the plant to maintain adequate reactor coolant system (RCS) inventory to ensure core cooling for the required coping duration, assuming a leak rate of 25 gal/min per reactor coolant pump (RCP) seal in addition to technical specification (TS) leak rate

  11. Analysis of Peach Bottom station blackout with MELCOR

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  12. 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

  13. Cooling methods of station blackout scenario for LWR plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The objective of this study is to analyze the cooling method of station blackout scenario for both the BWR and PWR plants by RELAP5 code and to check the validity of the cooling method proposed by the utilities. In the BWR plant cooling scenario, the Reactor Core Isolation Cooling System (RCIC), which is operated with high pressure steam from the reactor, injects cooling water into the reactor to keep the core water level. The steam generated in the core is released into the suppression pool at containment vessel to condense. To restrict the containment vessel pressure rising, the ventilation from the wet-well is operated. The scenario is analyzed by RELAP5 code. In the PWR plant scenario, the primary pressure is decreased by the turbine-driven auxiliary feed water system operated with secondary side steam of the steam generators (SGs). And the core cooling is kept by the natural circulation flow at the primary loop. From the RELAP5 code analysis, it was shown that the primary system cooling was practicable by using the turbine-driven auxiliary feed water system. (author)

  14. Extended Station Blackout Analysis for VVER-1000 MWe Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A. J.; Rao, R. S.; Lakshmanan, S. P.; Gupta, A., E-mail: avinashg@aerb.gov.in [Atomic Energy Regulatory Board, Mumbai (India)

    2014-10-15

    Post Fukushima, the plant behaviour for an extended station black-out (ESBO) scenario with only passive system availability for about 7 days has become imperative. Thermal hydraulic analysis of ESBO with the availability of passive heat removal system (PHRS), passive first stage and second stage hydro accumulators were carried out to demonstrate the design capabilities. Two different cases having primary leak rates of 2.2 tons/hr and 6.6 tons/hr were analyzed to study sustenance of natural circulation. For the study, out of 4 PHRS trains, one PHRS train was assumed to be in failure mode. The objective here is to predict the core cooling capability for a period of 7 days under ESBO conditions with the available water inventories from first and second stage hydroaccumulators only. Over simplified energy balance studies cannot ascertain sustenance of natural circulation in the primary system, steam generators (SGs) and PHRS. The analysis was carried out by using system thermal hydraulic safety code RELAP5/SCDAP/MOD 3.4. It is inferred that the inventory in the first stage accumulators and second stage accumulators compensate the leak and decay heat is removed effectively with the help of passive heat removal systems. It is also observed that even after 7 days of ESBO a large inventory is still available in the second stage accumulators and the primary system remains subcooled. (author)

  15. Cooling methods of station blackout scenario for LWR plants

    International Nuclear Information System (INIS)

    2012-01-01

    The objective of this study is to analyze the cooling method of station blackout scenario for both the BWR and PWR plants by RELAP5 code and to check the validity of the cooling method proposed by the utilities. In the BWR plant cooling scenario, the Reactor Core Isolation Cooling System (RCIC), which is operated with high pressure steam from the reactor, injects cooling water into the reactor to keep the core water level. The steam generated in the core is released into the suppression pool at containment vessel to condense. To restrict the containment vessel pressure rising, the ventilation from the wet-well is operated. The scenario is analyzed by RELAP5 and CONTEMPT-LT code. In the PWR plant scenario, the primary pressure is decreased by the turbine-driven auxiliary feed water system operated with secondary side steam of the steam generators (SGs). And the core cooling is kept by the natural circulation flow at the primary loop. The analytical method of un-uniform flow behavior among the SG U-tubes, which affects the natural circulation flow rate, is developed. (author)

  16. MELCOR simulation of long-term station blackout at Peach Bottom

    International Nuclear Information System (INIS)

    Madni, I.K.

    1990-01-01

    This paper presents the results from MELCOR (Version 1.8BC) calculations of the Long-Term Station Blackout Accident Sequence, with failure to depressurize the reactor vessel, at the Peach Bottom (BWR Mark I) plant, and presents comparisons with Source Term Code Package (STCP) calculations of the same sequence. This sequence assumes that batteries are available for six hours following loss of all power to the plant. Following battery failure, the reactor coolant system (RCS) inventory is boiled off through the relief valves by continued decay heat generation. This leads to core uncovery, heatup, clad oxidation, core degradation, relocation, and, eventually, vessel failure at high pressure. STCP has calculated the transient out to 13.5 hours after core uncovery. The results include the timing of key events, pressure and temperature response in the reactor vessel and containment, hydrogen production, and the release of source terms to the environment. 12 refs., 23 figs., 3 tabs

  17. The Safety Assessment of OPR-1000 for Station Blackout Applying Combined Deterministic and Probabilistic Procedure

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Gu; Ahn, Seung-Hoon; Cho, Dae-Hyung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    This is termed station blackout (SBO). However, it does not generally include the loss of available AC power to safety buses fed by station batteries through inverters or by alternate AC sources. Historically, risk analysis results have indicated that SBO was a significant contributor to overall core damage frequency. In this study, the safety assessment of OPR-1000 nuclear power plant for SBO accident, which is a typical beyond design basis accident and important contributor to overall plant risk, is performed by applying the combined deterministic and probabilistic procedure (CDPP). In addition, discussions are made for reevaluation of SBO risk at OPR-1000 by eliminating excessive conservatism in existing PSA. The safety assessment of OPR-1000 for SBO accident, which is a typical BDBA and significant contributor to overall plant risk, was performed by applying the combined deterministic and probabilistic procedure. However, the reference analysis showed that the CDF and CCDP did not meet the acceptable risk, and it was confirmed that the SBO risk should be reevaluated. By estimating the offsite power restoration time appropriately, the SBO risk was reevaluated, and it was finally confirmed that current OPR-1000 system lies in the acceptable risk against the SBO. In addition, it was demonstrated that the proposed CDPP is applicable to safety assessment of BDBAs in nuclear power plants without significant erosion of the safety margin.

  18. Benchmarking Simulation of Long Term Station Blackout Events

    International Nuclear Information System (INIS)

    Kim, Sung Kyum; Lee, John C.; Fynan, Douglas A.; Lee, John C.

    2013-01-01

    The importance of passive cooling systems has emerged since the SBO events. Turbine-driven auxiliary feedwater (TD-AFW) system is the only passive cooling system for steam generators (SGs) in current PWRs. During SBO events, all alternating current (AC) and direct current (DC) are interrupted and then the water levels of steam generators become high. In this case, turbine blades could be degraded and cannot cool down the SGs anymore. To prevent this kind of degradations, improved TD-AFW system should be installed for current PWRs, especially OPR 1000 plants. A long-term station blackout (LTSBO) scenario based on the improved TD-AFW system has been benchmarked as a reference input file. The following task is a safety analysis in order to find some important parameters causing the peak cladding temperature (PCT) to vary. This task has been initiated with the benchmarked input deck applying to the State-of-the-Art Reactor Consequence Analyses (SOARCA) Report. The point of the improved TD-AFW is to control the water level of the SG by using the auxiliary battery charged by a generator connected with the auxiliary turbine. However, this battery also could be disconnected from the generator. To analyze the uncertainties of the failure of the auxiliary battery, the simulation for the time-dependent failure of the TD-AFW has been performed. In addition to the cases simulated in the paper, some valves (e. g., pressurizer safety valve), available during SBO events in the paper, could be important parameters to assess uncertainties in PCTs estimated. The results for these parameters will be included in a future study in addition to the results for the leakage of the RCP seals. After the simulation of several transient cases, alternating conditional expectation (ACE) algorithm will be used to derive functional relationships between the PCT and several system parameters

  19. Study of a Station Blackout Event in the PWR Plant

    International Nuclear Information System (INIS)

    Ching-Hui Wu; Tsu-Jen Lin; Tsu-Mu Kao

    2002-01-01

    On March 18, 2001, a PWR nuclear power plant located in the Southern Taiwan occurred a Station Blackout (SBO) event. Monsoon seawater mist caused the instability of offsite power grids. High salt-contained mist caused offsite power supply to the nuclear power plant very unstable, and forced the plant to be shutdown. Around 24 hours later, when both units in the plant were shutdown, several inadequate high cycles of bus transfer between 345 kV and 161 kV startup transformers degraded the emergency 4.16 kV switchgears. Then, in the Train-A switchgear room of Unit 1 occurred a fire explosion, when the degraded switchgear was hot shorted at the in-coming 345 kV breaker. Inadequate configuration arrangement of the offsite power supply to the emergency 4.16 kV switchgears led to loss of offsite power (LOOP) events to both units in the plant. Both emergency diesel generators (EDG) of Unit 1 could not be in service in time, but those of Unit 2 were running well. The SBO event of Unit 1 lasted for about two hours till the fifth EDG (DG-5) was lined-up to the Train-B switchgear. This study investigated the scenario of the SBO event and evaluated a risk profile for the SBO period. Guidelines in the SBO event, suggested by probabilistic risk assessment (PRA) procedures were also reviewed. Many related topics such as the re-configuration of offsite power supply, the addition of isolation breakers of the emergency 4.16 kV switchgears, the betterment of DG-5 lineup design, and enhancement of the reliability of offsite power supply to the PWR plant, etc., will be in further studies. (authors)

  20. Station blackout with failure of wired shutdown system for AHWR

    International Nuclear Information System (INIS)

    Srivastava, A.; Contractor, A.D.; Chatterjee, B.; Kumar, Rajesh

    2015-01-01

    Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type boiling light water cooled and heavy water moderated reactor. This reactor has several advance safety features. One of the important passive design features of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power level without primary coolant pumps. Station blackout (SBO) scenario has become very important in aftermath of Fukushima event. The existing reactor has to demonstrate that design features are sufficient to mitigate the scenario whereas the new reactor design are adding specific features to tackle such scenario for prolonged period. The present study demonstrates the design features of AHWR to mitigate the SBO scenario along with failure of wired shutdown system. SBO event leads to feed water pump trip and loss of condenser vacuum which in turn results into loss of feed water and turbine trip on low condenser vacuum signal. Stoppage of steam flow to the turbine and bypass to the condenser lead to bottling up of the system, causing MHT pressure to rise. In the absence of reactor scram, the pressure continues to rise. Isolation Condenser (IC) valve starts opening at a pressure of 7.65 MPa. The pressure continues to rise as IC system is designed for decay heat removal and reactor power is brought down to decay power level through Passive Poison Injection System (PPIS) when the pressure reaches 8.4 MPa. The analysis shows that the event do not lead to undesirable clad surface temperature rise due to reactor trip by PPIS and decay heat removal for prolonged time by IC system. Thermal hydraulic response of different parameters like pressure, temperatures, and flows in MHT system is analyzed for this scenario. Pressure during transient is found to be well below the system pressure criteria of 110% of design pressure. This analysis highlights the design robustness of AHWR. (author)

  1. Relap5 Analysis of Processes in Reactor Cooling Circuit and Reactor Cavity in Case of Station Blackout in RBMK-1500

    International Nuclear Information System (INIS)

    Kaliatka, A.

    2007-01-01

    Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-term core cooling. Such accident could lead to multiple ruptures of fuel channels with severe consequences and should be analyzed in order to estimate the timing of the key events and the possibilities for accident management. This paper presents the results of the analysis of station blackout at Ignalina NPP. Analysis was performed using thermal-hydraulic state-of-the-art RELAP5/MOD3.2 code. The response of reactor cooling system and the processes in the reactor cavity and its venting system in case of a few fuel-channel ruptures due to overheating were demonstrated. The possible measures for prevention of the development of this beyond design basis accident (BDBA) to a severe accident are discussed

  2. Case Study of Multi-Unit Risk: Multi-Unit Station Black-Out

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Kyemin; Jang, Seung-cheol [KAERI, Daejeon (Korea, Republic of); Heo, Gyunyoung [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    After Fukushima Daiichi Accident, importance and public concern for Multi-Unit Risk (MUR) or Probabilistic Safety Assessment (PSA) have been increased. Most of nuclear power plant sites in the world have more than two units. These sites have been facing the problems of MUR or accident such as Fukushima. In case of South Korea, there are generally more than four units on the same site and even more than ten units are also expected. In other words, sites in South Korea also have been facing same problems. Considering number of units on the same site, potential of these problems may be larger than other countries. The purpose of this paper is to perform case study based on another paper submitted in the conference. MUR is depended on various site features such as design, shared systems/structures, layout, environmental condition, and so on. Considering various dependencies, we assessed Multi-Unit Station Black-out (MSBO) accident based on Hanul Unit 3 and 4 model. In this paper, case study for multi-unit risk or PSA had been performed. Our result was incomplete to assess total multi-unit risk because of two challenging issues. First, economic impact had not been evaluated to estimate multi-unit risk. Second, large uncertainties were included in our result because of various assumptions. These issues must be resolved in the future.

  3. Preliminary Evaluations of CSPACE for a Station Blackout Transient in APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Lee, T. B.; Lee, D. K.; Lee, H. S.; Lee, G. W.; Choi, T. S. [KEPCO, Daejeon (Korea, Republic of); Park, R. J.; Kim, D. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This paper discusses the preliminary results of the simulated station blackout (SBO) transients using the CSPACE code and presents the information pertinent to the related safety issues. CSPACE is a merged program of a master processer of Safety and Performance Analysis Code (SPACE) for nuclear power plants and a child processer of Core Meltdown Progression Accident Simulation Software (COMPASS) generated as a dynamic-link library (DLL) codes. It has been developed to predict the best-estimate transient in the pressurized water reactor (PWR) for severe accidents. SPACE and COMPASS codes take charge of the thermal-hydraulic response of PWRs and the analysis of the severe accident progression in a vessel, respectively. The initial phase is estimated starting from time zero when the loss of off-site and on-site powers occurs simultaneously. Shortly after the RCS pressure initially falls and rises slightly due to the effects of the reactor and turbine trips, the RCS pressure declines in response to the cooling provided by heat removed to the SGs. During the period of the primary heat-up and boil-off, the RCS pressure increase is limited by two cycles of the POSRV. The RCS fluid mass is lost through the pressurizer POSRV and then the core uncovers and superheated steam flows out from the RV into the coolant loops starting at 5513.0 seconds.

  4. Comparative study of the hydrogen generation during short term station blackout (STSBO) in a BWR

    International Nuclear Information System (INIS)

    Polo-Labarrios, M.A.; Espinosa-Paredes, G.

    2015-01-01

    Highlights: • Comparative study of generation in a simulated STSBO severe accident. • MELCOR and SCDAP/RELAP5 codes were used to understanding the main phenomena. • Both codes present similar thermal-hydraulic behavior for pressure and boil off. • SCDAP/RELAP5 predicts 15.8% lower hydrogen production than MELCOR. - Abstract: The aim of this work is the comparative study of hydrogen generation and the associated parameters in a simulated severe accident of a short-term station blackout (STSBO) in a typical BWR-5 with Mark-II containment. MELCOR (v.1.8.6) and SCDAP/RELAP5 (Mod.3.4) codes were used to understand the main phenomena in the STSBO event through the results comparison obtained from simulations with these codes. Due that the simulation scope of SCDAP/RELAP5 is limited to failure of the vessel pressure boundary, the comparison was focused on in-vessel severe accident phenomena; with a special interest in the vessel pressure, boil of cooling, core temperature, and hydrogen generation. The results show that at the beginning of the scenario, both codes present similar thermal-hydraulic behavior for pressure and boil off of cooling, but during the relocation, the pressure and boil off, present differences in timing and order of magnitude. Both codes predict in similar time the beginning of melting material drop to the lower head. As far as the hydrogen production rate, SCDAP/RELAP5 predicts 15.8% lower production than MELCOR

  5. Case Study of Multi-Unit Risk: Multi-Unit Station Black-Out

    International Nuclear Information System (INIS)

    Oh, Kyemin; Jang, Seung-cheol; Heo, Gyunyoung

    2015-01-01

    After Fukushima Daiichi Accident, importance and public concern for Multi-Unit Risk (MUR) or Probabilistic Safety Assessment (PSA) have been increased. Most of nuclear power plant sites in the world have more than two units. These sites have been facing the problems of MUR or accident such as Fukushima. In case of South Korea, there are generally more than four units on the same site and even more than ten units are also expected. In other words, sites in South Korea also have been facing same problems. Considering number of units on the same site, potential of these problems may be larger than other countries. The purpose of this paper is to perform case study based on another paper submitted in the conference. MUR is depended on various site features such as design, shared systems/structures, layout, environmental condition, and so on. Considering various dependencies, we assessed Multi-Unit Station Black-out (MSBO) accident based on Hanul Unit 3 and 4 model. In this paper, case study for multi-unit risk or PSA had been performed. Our result was incomplete to assess total multi-unit risk because of two challenging issues. First, economic impact had not been evaluated to estimate multi-unit risk. Second, large uncertainties were included in our result because of various assumptions. These issues must be resolved in the future

  6. Toward quantification of the uncertainty in estimating frequency of critical station blackout

    International Nuclear Information System (INIS)

    Rodgers, Shawn; Betancourt, Coral; Kee, Ernie; Nelson, Paul; Rodi, Paul

    2011-01-01

    A formal statement of the critical station blackout problem is provided, and a solution given, up to evaluation of an n-dimensional 'non recovery integral' n =number of trains (parallel backup sources of electrical power). Several approaches that have been developed in the industry to estimate probability of critical station blackout are shown to be interpretable as special cases of such integrals. Computational results, for a simple model problem, suggest there is yet a very substantial over conservatism in current state-of-the-art techniques for estimating probability of critical station blackout. Research issues associated to possibly meeting this need via computational evaluation of the non recovery integral are discussed. (author)

  7. Comparison of static model and dynamic model for the evaluation of station blackout sequences

    International Nuclear Information System (INIS)

    Lee, Kwang-Nam; Kang, Sun-Koo; Hong, Sung-Yull.

    1992-01-01

    Station blackout is one of major contributors to the core damage frequency (CDF) in many PSA studies. Since station blackout sequence exhibits dynamic features, accurate calculation of CDF for the station blackout sequence is not possible with event tree/fault tree (ET/FT) method. Although the integral method can determine accurate CDF, it is time consuming and is difficult to evaluate various alternative AC source configuration and sensitivities. In this study, a comparison is made between static model and dynamic model and a new methodology which combines static model and dynamic model is provided for the accurate quantification of CDF and evaluation of improvement alternatives. Results of several case studies show that accurate calculation of CDF is possible by introducing equivalent mission time. (author)

  8. Assessment of the potential for HPME during a station blackout in the Surry and Zion PWRS

    International Nuclear Information System (INIS)

    Knudson, D.L.; Bayless, P.D.; Dobbe, C.A.; Odar, F.

    1994-01-01

    The integrity of a PWR (pressurized water reactor) containment structure could be challenged by direct heating associated with a HPME (high pressure melt ejection) of core materials following reactor vessel lower head breach during certain severe accidents. Structural failure resulting from direct containment heating is a contributor to the risk of operating a PWR. Intentional RCS (reactor coolant system) depressurization, where operators latch pressurizer relief valves open, has been proposed as an accident management strategy to reduce those risks by mitigating the severity of the HPME. However, decay heat levels, valve capacities, and other plant-specific characteristics determine whether the required operator action will be effective. Without operator action, natural circulation flows could heat ex-vessel RCS pressure boundaries (surge line and hot leg piping, steam generator tubes, etc.) to the point of failure before failure of the lower head providing an unintentional mechanism for depressurization and HPME mitigation. This paper summarizes an assessment of RCS depressurization with respect to the potential for HPME during a station blackout in the Surry and Zion PWRs. The assessment included a detailed transient analysis using the SCDAP/RELAP5/MOD3 computer code and an evaluation of RCS depressurization-related probabilities primarily based on the code results

  9. Implications of Extension of Station Blackout Cooping Capability on Nuclear Power Plant Safety

    International Nuclear Information System (INIS)

    Volkanovski, Andrija

    2015-01-01

    The safety of the nuclear power plant depends on the availability of the continuous and reliable sources of electrical energy during all modes of operation of the plant. The station blackout corresponds to a total loss of all alternate current (AC) power as a result of complete failure of both offsite and on-site AC power sources. The electricity for the essential systems during station blackout is provided from the batteries installed in the nuclear power plant. The results of the probabilistic safety assessment show that station blackout is one of the main and frequently the dominant contributor to the core damage frequency. Results of the analysis of the implications of the strengthening of the SBO mitigation capability on safety of the NPP will be presented. The assessment is done with state-of-art deterministic and probabilistic methods and tolls with application on reference models of nuclear power plants. The safety analysis is done on reference model of the nuclear power plant. Obtained results show large decrease of the core damage frequency with strengthening of the station blackout mitigation capability. The time extension of blackout coping capability results in the delay of the core heat up for at least the extension time interval. Availability and operation of the steam driven auxiliary feedwater system maintains core integrity up to 72 h after the successful shutdown, even in the presence of the reactor coolant pumps seal leakage. The largest weighted decrease of the core damage frequency considering the costs for the modification is obtained for the modification resulting in extension of the station blackout coping capability. The importance of the common cause failures of the emergency diesel generators for the obtained decrease of the core damage frequency and overall safety of the plant is identified in the obtained results. (authors)

  10. Transfer Effect Ratio of Loosely Coupled Coils for Wireless Power through CB Wall under Station Blackout(SBO)

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Kil Mo; Hong, Seong Wan; Song, Jin Ho; Baek, Won Pil [KAERI, Daejeon (Korea, Republic of); Cheon, Sang Hoon [ETRI, Daejeon (Korea, Republic of)

    2016-05-15

    Instrumentations have had the bad situation like a station blackout(SBO) as the severe accident in nuclear power plants. In recent years, there has been an increasing interest in wireless power transfer technology, In particular, significant processing has been charted for inductively coupled systems. In this paper, we introduce some new method as transfer effect ratio of loosely coupled coils for wireless power through the CB(Container Building) wall as an alternative method under a station blackout of severe accident conditions in nuclear power plants. As an equivalent circuit model that can describe wireless energy transfer systems via coupled magnetic resonances for the CB thickness wall. The solution shows that the transmission efficiency can be decreased simply by adjusting the spacing between the power and the sending coils or between the receiving and the load coils. The system design can be calculated the frequency characteristics, and then an equivalent circuit model was developed from the node equation and established in an electric design automation tool.

  11. Transfer Effect Ratio of Loosely Coupled Coils for Wireless Power through CB Wall under Station Blackout(SBO)

    International Nuclear Information System (INIS)

    Koo, Kil Mo; Hong, Seong Wan; Song, Jin Ho; Baek, Won Pil; Cheon, Sang Hoon

    2016-01-01

    Instrumentations have had the bad situation like a station blackout(SBO) as the severe accident in nuclear power plants. In recent years, there has been an increasing interest in wireless power transfer technology, In particular, significant processing has been charted for inductively coupled systems. In this paper, we introduce some new method as transfer effect ratio of loosely coupled coils for wireless power through the CB(Container Building) wall as an alternative method under a station blackout of severe accident conditions in nuclear power plants. As an equivalent circuit model that can describe wireless energy transfer systems via coupled magnetic resonances for the CB thickness wall. The solution shows that the transmission efficiency can be decreased simply by adjusting the spacing between the power and the sending coils or between the receiving and the load coils. The system design can be calculated the frequency characteristics, and then an equivalent circuit model was developed from the node equation and established in an electric design automation tool

  12. Analysis of core uncovery time in Kuosheng station blackout transient with MELCOR

    International Nuclear Information System (INIS)

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

    1996-01-01

    The MELCOR code, developed by the Sandia National Laboratories, is capable of simulating severe accident phenomena of nuclear power plants. Core uncovery time is an important parameter in the probabilistic risk assessment. However, many MELCOR users do not generate the initial conditions in a station blackout (SBO) transient analysis. Thus, achieving reliable core uncovery time is difficult. The core uncovery time for the Kuosheng nuclear power plant during an SBO transient is analyzed. First, full-power steady-state conditions are generated with the application of a developed self-initialization algorithm. Then the response of the SBO transient up to core uncovery is simulated. The effects of key parameters including the initialization process and the reactor feed pump (RFP) coastdown time on the core uncovery time are analyzed. The initialization process is the most important parameter that affects the core uncovery time. Because SBO transient analysis, the correct initial conditions must be generated to achieve a reliable core uncovery time. The core uncovery time is also sensitive to the RFP coastdown time. A correct time constant is required

  13. Analysis of economics and safety to cope with station blackout in PWR

    International Nuclear Information System (INIS)

    Al Shehhi, Ahmed Saeed; Chang, Soon Heung; Kim, Sang Ho; Kang, Hyun Gook

    2013-01-01

    Highlights: • Proposed framework covers all aspects of very complicated decision making. • We addressed the various options against SBO. • Emergency water supply through the steam generator hookup was considered. • Optimal testing interval of EDG was determined in various design options. • Effect of risk aversion factor on decision making was quantitatively illustrated. - Abstract: Design and operation options that can reduce both the initiating event frequency and the accident mitigation probability were addressed in an integrated framework to cope with station blackout. The safety, engineering cost, water delivery cost and testing/maintenance cost of each option were quantitatively evaluated to calculate the cost variation and to find an optimal point in the reference reactor, OPR1000. Design variables that represent additional emergency water supply, diverse emergency diesel generator, and surveillance test period modification were investigated. Based on these design variables, we applied the developed formula to quantify cost items, which were presented as changes of the economics and the safety. A case study was provided to illustrate the change of the total cost. Different risk aversion factors that represent different attitudes of the public were also investigated. The result shows that the costs and benefits of various complicated options can be effectively addressed with the proposed risk-informed decision making framework

  14. Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using Trace

    International Nuclear Information System (INIS)

    El-Sahlamy, N.M.

    2017-01-01

    One of the main concerns in the area of severe accidents in nuclear reactors is that of station blackout (SBO). The loss of offsite electrical power concurrent with the unavailability of the onsite emergency alternating current (AC) power system can result in loss of decay heat removal capability, leading to a potential core damage which may lead to undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. This paper provides a best estimate assessment of the SBO scenario in a 3-loop Westinghouse PWR reactor. The evaluation is performed using TRACE, a best estimate computer code for thermal-hydraulic calculations. Two sets of scenarios for SBO analyses are discussed in the current work. The first scenario is the short term SBO where it is assumed that in addition to the loss of AC power, there is no DC power; i.e., no batteries are available. In the second scenario, a long term SBO is considered. For this scenario, DC batteries are available for four hours. The aim of the current SBO analyses for the 3-loop pressurized water reactor presented in this paper is to focus on the effect of the availability of a DC power source to delay the time to core uncovers and heatup

  15. The integrity of NSSS and containment during extended station blackout for Kuosheng BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Keng-Hsien; Yuann, Yng-Ruey; Lin, Ansheng [Atomic Energy Council, Taoyuan City, Taiwan (China). Inst. of Nuclear Energy Research

    2017-11-15

    The Fukushima Daiichi accident occurring on March 11, 2011, reveals that Station Blackout (SBO) may last longer than 8 h. However, the original design may not have sufficient capacity to cope with a SBO for more than 8 h. In view of this, Taiwan Power Company has initiated several enhancements to mitigate the severity of the extended SBO. Based on the improved plant configuration, a SBO coping analysis is performed in this study to assess whether the Kuosheng BWR plant has sufficient capability to cope with SBO for 24 h with respect to maintaining the integrity of the reactor core and containment. The analyses in the Nuclear Steam Supply System (NSSS) and the containment are based on the RETRAN-3D and GOTHIC models, respectively. The flow conditions calculated by RETRAN-3D during the event are retrieved and input to the GOTHIC containment model to determine the containment pressure and temperature response. These boundary conditions include SRV flow rate, SRV flow enthalpy, and total reactor coolant system leakage flow rate.

  16. The integrity of NSSS and containment during extended station blackout for Kuosheng BWR plant

    International Nuclear Information System (INIS)

    Hsu, Keng-Hsien; Yuann, Yng-Ruey; Lin, Ansheng

    2017-01-01

    The Fukushima Daiichi accident occurring on March 11, 2011, reveals that Station Blackout (SBO) may last longer than 8 h. However, the original design may not have sufficient capacity to cope with a SBO for more than 8 h. In view of this, Taiwan Power Company has initiated several enhancements to mitigate the severity of the extended SBO. Based on the improved plant configuration, a SBO coping analysis is performed in this study to assess whether the Kuosheng BWR plant has sufficient capability to cope with SBO for 24 h with respect to maintaining the integrity of the reactor core and containment. The analyses in the Nuclear Steam Supply System (NSSS) and the containment are based on the RETRAN-3D and GOTHIC models, respectively. The flow conditions calculated by RETRAN-3D during the event are retrieved and input to the GOTHIC containment model to determine the containment pressure and temperature response. These boundary conditions include SRV flow rate, SRV flow enthalpy, and total reactor coolant system leakage flow rate.

  17. Summary of ROSA-4 LSTF first phase test program and station blackout (TMLB) test results

    International Nuclear Information System (INIS)

    Tasaka, K.; Kukita, Y.; Anoda, Y.

    1990-01-01

    This paper summarizes major test results obtained at the ROSA-4 Large Scale Test Facility (LSTF) during the first phase of the test program. The results from a station blackout (TMLB) test conducted at the end of the first-phase program are described in some detail. The LSTF is an integral test facility being operated by the Japan Atomic Energy Research Institute for simulation of pressurized water reactor (PWR) thermal-hydraulic responses during small-break loss-of-coolant accidents (SBLOCAs) and operational/abnormal transients. It is a 1/48 volumetrically scaled, full-height, full-pressure simulator of a Westinghouse-type 4-loop PWR. The facility includes two symmetric primary loops each one containing an active inverted-U tube steam generator and an active reactor coolant pump. The loop horizontal legs are sized to conserve the scaled (1/24) volumes as well as the length to the square root of the diameter ratio in order to simulate the two-phase flow regime transitions. The primary objective of the LSTF first-phase program was to define the fundamental PWR thermal-hydraulic responses during SBLOCAs and transients. Most of the tests were conducted with simulated component/operator failures, including unavailability of the high pressure injection system and auxiliary feedwater system, as well as operator failure to take corrective actions. The forty-two first phase tests included twenty-nine SBLOCA tests conducted mainly for cold leg breaks, three abnormal transient tests and ten natural circulation tests. Attempts were made in several of the SBLOCA tests to simulate the plant recovery procedures as well as candidate accident management measures for prevention of high-pressure core melt situation. The natural circulation tests simulated the single-phase and two-phase natural circulation as well as reflux condensation behavior in the primary loops in steady or quasi-steady states

  18. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout caused by external flooding using the RISMC toolkit

    International Nuclear Information System (INIS)

    2014-01-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.

  19. Probabilistic assessment of the Juragua NPP response under Station Blackout conditions

    International Nuclear Information System (INIS)

    Valhuerdi Debesa, C.; Vilaragut Llanes, J.J.

    1996-01-01

    Assessment of the NPP response under SBO (station Blackout) conditions is a current safety issue of special interest, In the case of Juragua NPP, the safety assessment related to this topic is very important, taking into account the peculiarities of the Cuban Electro energetic System: small and long island, without possibilities of conexion beyond its borders and under the incidence of tropical phenomena In this papers a preliminary evaluation is presented of the potential incidence of Station Blackout conditions for Juragua NPP. the importance of this sort of events for the safety of the plant is evaluated, the factors which condition it are identified and measures for its prevention or recovering the normal situation if such an event takes place are proposed

  20. Regulatory analysis for the resolution of Unresolved Safety Issue A-44, Station Blackout. Draft report

    International Nuclear Information System (INIS)

    Rubin, A.M.

    1986-01-01

    ''Station Blackout'' is the complete loss of alternating current (ac) electric power to the essential and nonessential buses in a nuclear power plant; it results when both offsite power and the onsite emergency ac power systems are unavailable. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on ac power, the consequences of a station blackout could be severe. Because of the concern about the frequency of loss of offsite power, the number of failures of emergency diesel generators, and the potentially severe consequences of a loss of all ac power, ''Station Blackout'' was designated as Unresolved Safety Issue (USI) A-44. This report presents the regulatory analysis for USI A-44. It includes: (1) a summary of the issue, (2) the proposed technical resolution, (3) alternative resolutions considered by the Nuclear Regulatory Commission (NRC) staff, (4) an assessment of the benefits and costs of the recommended resolution, (5) the decision rationale, and (6) the relationship between USI A-44 and other NRC programs and requirements

  1. AP1000 station blackout study with and without depressurization using RELAP5/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

    2016-10-15

    Highlights: • A representative RELAP5/SCDAPSIM model of AP1000 has been developed. • Core is modeled using SCDAP. • A SBO for the AP1000 has been simulated for high pressure (no depressurization) and low pressure (depressurization). • Significant differences in the damage progression have been observed for the two cases. • Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario. - Abstract: Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most of the countries. For such improbable accidents, a SBO for the AP1000 using RELAP5/SCDAPSIM has been simulated. Many improvements have been made in fuel damage progression models of SCDAP after the Fukushima accident which are now being tested for the new reactor designs. AP1000 is a 2-loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies. The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. The

  2. Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

    International Nuclear Information System (INIS)

    Hidaka, Akihide; Asaka, Hideaki; Sugimoto, Jun; Ueno, Shingo; Yoshino, Takehito

    1999-12-01

    In PWR severe accidents such as station blackout, the integrity of steam generator U-tube would be threatened early at the transient among the pipes of primary system. This is due to the hot leg countercurrent natural circulation (CCNC) flow which delivers the decay heat of the core to the structures of primary system if the core temperature increases after the secondary system depressurization. From a view point of accident mitigation, this steam generator tube rupture (SGTR) is not preferable because it results in the direct release of primary coolant including fission products (FP) to the environment. Recent SCDAP/RELAP5 analyses by USNRC showed that the creep failure of pressurizer surge line which results in release of the coolant into containment would occur earlier than SGTR during the secondary system depressurization. However, the analyses did not consider the decay heat from deposited FP on the steam generator U-tube surface. In order to investigate the effect of decay heat on the steam generator U-tube integrity, the hot leg CCNC flow model used in the USNRC's calculation was, at first, validated through the analysis for JAERI's LSTF experiment. The CCNC model reproduced well the thermohydraulics observed in the LSTF experiment and thus the model is mostly reliable. An analytical study was then performed with SCDAP/RELAP5 for TMLB' sequence of Surry plant with and without secondary system depressurization. The decay heat from deposited FP was calculated by JAERI's FP aerosol behavior analysis code, ART. The ART analysis showed that relatively large amount of FPs may deposit on steam generator U-tube inlet mainly by thermophoresis. The SCDAP/RELAP5 analyses considering the FP decay heat predicted small safety margin for steam generator U-tube integrity during secondary system depressurization. Considering associated uncertainties in the analyses, the potential for SGTR cannot be ignored. Accordingly, this should be considered in the evaluation of merits

  3. Electrical Power System Design and Station Blackout (SBO) Management in Indian Fast Breeder Reactors

    International Nuclear Information System (INIS)

    Vijaya, N. M.; Theivarajan, N.; Madhusoodanan, K.

    2015-01-01

    In the nuclear new builds and projects in design stage SBO management measures have significant role. Depending on the onsite and offsite power supply configurations, deterministic SBO duration is established. Design of systems with adequately sized battery capacities for SBO duration, special SBO Diesel Generator Sets, structured load shedding strategy to conserve battery availability to cope with SBO and to monitor the plant safety beyond SBO duration are considered as part of electrical system design now. In the design of PFBR, SBO is given due importance right from conceptual design stage. Both deterministic SBO duration and probabilistic SBO duration versus frequency were established by detailed analysis. Dedicated DC power supply systems and additional SBO DG back-up systems are in place to cope with normal and extended SBO. After the Fukushima event, there is greater requirement to demonstrate plant safety during SBO for a long duration extended over several days. In light of this accident, thermal hydraulic synthesis of PFBR has been carried out to ascertain the capability of the plant to manage a prolonged station blackout event. This has brought out the robustness of the design. Safety design features of PFBR ensure comfortable management of extended SBO. In the design of future FBR projects, current trends in the new nuclear builds and recommendations of international bodies considering Fukushima are duly considered. SBO measures by means of alternate AC power sources, redundant emergency power supply sources with less dependence on other auxiliary systems and dedicated DC power systems are considered to cope with normal and extended SBO beyond design basis. Right from the conceptual design, the system robustness to manage normal and extended SBO will be taken care with the related thermal hydraulic and associated analysis. The paper highlights these SBO management strategies in PFBR and future FBRs. (author)

  4. Modeling and analysis of selected organization for economic cooperation and development PKL-3 station blackout experiments using TRACE

    Directory of Open Access Journals (Sweden)

    Roman Mukin

    2018-04-01

    Full Text Available A series of tests dedicated to station blackout (SBO accident scenarios have been recently performed at the Primärkreislauf-Versuchsanlage (primary coolant loop test facility; PKL facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal–hydraulic code TRACE (v5.0 Patch4 of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for

  5. A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Haihua [Idaho National Laboratory; Zhang, Hongbin [Idaho National Laboratory; Zou, Ling [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory

    2015-03-01

    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoid overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety

  6. Development of the APR1400 model for countercurrent natural circulation in hot leg and steam generator under station blackout

    International Nuclear Information System (INIS)

    Park, Sang Gil; Kim, Han Chul

    2012-01-01

    In order to analyze severe accident phenomena, Korea Institute of Nuclear Safety (KINS) made a MELCOR model for APR1400 to examine natural circulation and creep rupture failure in the Reactor Coolant System (RCS) under station blackout (SBO). In this study, we are trying to advance the former model to describe natural circulation more accurately. After Fukushima accident, the concerns of severe accident management, assuring the heat removal capability, has risen for the case when the SBO is happened and there are no more electric powers to cool down decay heat. Under SBO there are three kinds of natural circulation which can delay the core heatup. One is in vessel natural circulation in the upper plenum of reactor vessel and the second is countercurrent natural circulation in hot leg through steam generator tubes and the last is full loop natural circulation when the reactor coolant pump loop seal is cleared and reactor coolant pump sealing is damaged by high temperature and high pressure. Among them this study focuses on the countercurrent natural circulation model using MELCOR1.8.6

  7. Probabilistic assessment of Juragua Nuclear Power Plant response under station blackout conditions

    International Nuclear Information System (INIS)

    Valhuerdi, C.; Vilaragut, J.J.; Perdomo, M.; Torres, A.

    1995-01-01

    The preliminary results concerning the response of station blackout are shown in this paper. These results have been obtained in the framework of initiator lass of external electrical supply as a aport of the revision o of the current probabilistic safety analysis. The work is also based on the results reported in the thermohydraulic calculations of VVER 440 plants responses under these conditions and the experience of this type of notified incidents. Finally, a comparative analysis with the results obtained for other reactor technologies is presented

  8. Thermohydraulic status and component behavior in the PWR during the selected meltdown scenario station blackout (SBO); Thermohydraulisches Verhalten und Komponentenverhalten eines DWR bei ausgewaehltem Kernschmelzszenarium infolge Station Blackout (SBO). Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Band, Sebastian; Blaesius, Christoph; Scheuerer, Martina; Steinroetter, Thomas

    2017-09-15

    The report on the thermohydraulic status and component behavior in the PWR during the selected meltdown scenario station blackout (SBO) includes the following issues: status of science and technology on this topic, analysis of a high-pressure meltdown scenario using ATHLET-CD for a German PWR starting from the initiating event station blackout, three-dimensional computational fluid dynamic (CFD) analyses of the pressurizer coolant loop in a generic German PWR, evaluation of the thermohydraulic steam generator behavior and its effect on the involved primary circuit components.

  9. Investigation of station blackout scenario in VVER440/v230 with RELAP5 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Gencheva, Rositsa Veselinova, E-mail: roseh@mail.bg; Stefanova, Antoaneta Emilova, E-mail: antoanet@inrne.bas.bg; Groudev, Pavlin Petkov, E-mail: pavlinpg@inrne.bas.bg

    2015-12-15

    Highlights: • We have modeled SBO in VVER440. • RELAP5/MOD3 computer code has been used. • Base case calculation has been done. • Fail case calculation has been done. • Operator and alternative operator actions have been investigated. - Abstract: During the development of symptom-based emergency operating procedures (SB-EOPs) for VVER440/v230 units at Kozloduy Nuclear Power Plant (NPP) a number of analyses have been performed using the RELAP5/MOD3 (Carlson et al., 1990). Some of them investigate the response of VVER440/v230 during the station blackout (SBO). The main purpose of the analyses presented in this paper is to identify the behavior of important VVER440 parameters in case of total station blackout. The RELAP5/MOD3 has been used to simulate the SBO in VVER440 NPP model (Fletcher and Schultz, 1995). This model was developed at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events and design based scenarios. The model provides a significant analytical capability for specialists working in the field of NPP safety.

  10. PWR station blackout transient simulation in the INER integral system test facility

    International Nuclear Information System (INIS)

    Liu, T.J.; Lee, C.H.; Hong, W.T.; Chang, Y.H.

    2004-01-01

    Station blackout transient (or TMLB' scenario) in a pressurized water reactor (PWR) was simulated using the INER Integral System Test Facility (IIST) which is a 1/400 volumetrically-scaled reduce-height and reduce-pressure (RHRP) simulator of a Westinghouse three-loop PWR. Long-term thermal-hydraulic responses including the secondary boil-off and the subsequent primary saturation, pressurization and core uncovery were simulated based on the assumptions of no offsite and onsite power, feedwater and operator actions. The results indicate that two-phase discharge is the major depletion mode since it covers 81.3% of the total amount of the coolant inventory loss. The primary coolant inventory has experienced significant re-distribution during a station blackout transient. The decided parameter to avoid the core overheating is not the total amount of the coolant inventory remained in the primary core cooling system but only the part of coolant left in the pressure vessel. The sequence of significant events during transient for the IIST were also compared with those of the ROSA-IV large-scale test facility (LSTF), which is a 1/48 volumetrically-scaled full-height and full-pressure (FHFP) simulator of a PWR. The comparison indicates that the sequence and timing of these events during TMLB' transient studied in the RHRP IIST facility are generally consistent with those of the FHFP LSTF. (author)

  11. Assessment of the potential for high-pressure melt ejection resulting from a Surry station blackout transient

    International Nuclear Information System (INIS)

    Knudson, D.L.; Dobbe, C.A.

    1993-11-01

    Containment integrity could be challenged by direct heating associated with a high pressure melt ejection (HPME) of core materials following reactor vessel breach during certain severe accidents. Intentional reactor coolant system (RCS) depressurization, where operators latch pressurizer relief valves open, has been proposed as an accident management strategy to reduce risks by mitigating the severity of HPME. However, decay heat levels, valve capacities, and other plant-specific characteristics determine whether the required operator action will be effective. Without operator action, natural circulation flows could heat ex-vessel RCS pressure boundaries (surge line and hot leg piping, steam generator tubes, etc.) to the point of failure before vessel breach, providing an alternate mechanism for RCS depressurization and HPME mitigation. This report contains an assessment of the potential for HPME during a Surry station blackout transient without operator action and without recovery. The assessment included a detailed transient analysis using the SCDAP/RELAP5/MOD3 computer code to calculate the plant response with and without hot leg countercurrent natural circulation, with and without reactor coolant pump seal leakage, and with variations on selected core damage progression parameters. RCS depressurization-related probabilities were also evaluated, primarily based on the code results

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

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1995-06-01

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

  13. Branch-and-Bound algorithm applied to uncertainty quantification of a Boiling Water Reactor Station Blackout

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, Joseph, E-mail: joseph.nielsen@inl.gov [Idaho National Laboratory, 1955 N. Fremont Avenue, P.O. Box 1625, Idaho Falls, ID 83402 (United States); University of Idaho, Department of Mechanical Engineering and Nuclear Engineering Program, 1776 Science Center Drive, Idaho Falls, ID 83402-1575 (United States); Tokuhiro, Akira [University of Idaho, Department of Mechanical Engineering and Nuclear Engineering Program, 1776 Science Center Drive, Idaho Falls, ID 83402-1575 (United States); Hiromoto, Robert [University of Idaho, Department of Computer Science, 1776 Science Center Drive, Idaho Falls, ID 83402-1575 (United States); Tu, Lei [University of Idaho, Department of Mechanical Engineering and Nuclear Engineering Program, 1776 Science Center Drive, Idaho Falls, ID 83402-1575 (United States)

    2015-12-15

    state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. Unfortunately DPRA methods introduce issues associated with combinatorial explosion of states. This paper presents a methodology to address combinatorial explosion using a Branch-and-Bound algorithm applied to Dynamic Event Trees (DET), which utilize LENDIT (L – Length, E – Energy, N – Number, D – Distribution, I – Information, and T – Time) as well as a set theory to describe system, state, resource, and response (S2R2) sets to create bounding functions for the DET. The optimization of the DET in identifying high probability failure branches is extended to create a Phenomenological Identification and Ranking Table (PIRT) methodology to evaluate modeling parameters important to safety of those failure branches that have a high probability of failure. The PIRT can then be used as a tool to identify and evaluate the need for experimental validation of models that have the potential to reduce risk. In order to demonstrate this methodology, a Boiling Water Reactor (BWR) Station Blackout (SBO) case study is presented.

  14. Branch-and-Bound algorithm applied to uncertainty quantification of a Boiling Water Reactor Station Blackout

    International Nuclear Information System (INIS)

    Nielsen, Joseph; Tokuhiro, Akira; Hiromoto, Robert; Tu, Lei

    2015-01-01

    state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. Unfortunately DPRA methods introduce issues associated with combinatorial explosion of states. This paper presents a methodology to address combinatorial explosion using a Branch-and-Bound algorithm applied to Dynamic Event Trees (DET), which utilize LENDIT (L – Length, E – Energy, N – Number, D – Distribution, I – Information, and T – Time) as well as a set theory to describe system, state, resource, and response (S2R2) sets to create bounding functions for the DET. The optimization of the DET in identifying high probability failure branches is extended to create a Phenomenological Identification and Ranking Table (PIRT) methodology to evaluate modeling parameters important to safety of those failure branches that have a high probability of failure. The PIRT can then be used as a tool to identify and evaluate the need for experimental validation of models that have the potential to reduce risk. In order to demonstrate this methodology, a Boiling Water Reactor (BWR) Station Blackout (SBO) case study is presented.

  15. Comparison of oxide- and metal-core behavior during CRBRP [Clinch River Breeder Reactor Plant] station blackout

    International Nuclear Information System (INIS)

    Polkinghorne, S.T.; Atkinson, S.A.

    1986-01-01

    A resurrected concept that could significantly improve the inherently safe response of Liquid-Metal cooled Reactors (LMRs) during severe undercooling transients is the use of metallic fuel. Analytical studies have been reported on for the transient behavior of metal-fuel cores in innovative, inherently safe LMR designs. This paper reports on an analysis done, instead, for the Clinch River Breeder Reactor Plant (CRBRP) design with the only innovative change being the incorporation of a metal-fuel core. The SSC-L code was used to simulate a protected station blackout accident in the CRBRP with a 943 MWt Integral Fast Reactor (IFR) metal-fuel core. The results, compared with those for the oxide-fueled CRBRP, show that the margin to boiling is greater for the IFR core. However, the cooldown transient is more severe due to the faster thermal response time of metallic fuel. Some additional calculations to assess possible LMR design improvements (reduced primary system pressure losses, extended flow coastdown) are also discussed. 8 refs., 13 figs., 2 tabs

  16. RELAP5/MOD3.3 Analyses of Core Heatup Prevention Strategy During Extended Station Blackout in PWR

    International Nuclear Information System (INIS)

    Prosek, A.

    2016-01-01

    The accident at the Fukushima Dai-ichi nuclear power plant demonstrated the vulnerability of the plants on the loss of electrical power for several days, so called extended station blackout (SBO). A set of measures have been proposed and implemented in response of the accident at the Fukushima Dai-ichi nuclear power plant. The purpose of the study was to investigate the application of the deterministic safety analysis for core heatup prevention strategy of the extended SBO in pressurized water reactor, lasting 72 h. The prevention strategy selected was water injection into steam generators using turbine driven auxiliary feedwater pump (TD-AFW) or portable water injection pump. Method for assessment of the necessary pump injection flowrate is developed and presented. The necessary injection flowrate to the steam generators is determined from the calculated cumulative water mass injected by the turbine driven auxiliary feedwater pump in the analysed scenarios, when desired normal level is maintained automatically. The developed method allows assessment of the necessary injection flowrates of pump, TD-AFW or portable, for different plant configurations and number of flowrate changes. The RELAP5/MOD3.3 Patch04 computer code and input model of a two-loop pressurized water reactor is used for analyses, assuming different injection start times, flowrates and reactor coolant system losses. Three different reactor coolant system (RCS) coolant loss pathways, with corresponding leakage rate, can be expected in the pressurized water reactor (PWR) during the extended SBO: normal system leakage, reactor coolant pump seal leakage, and RCS coolant loss through letdown relief valve unless automatically isolated or until isolation is procedurally directed. Depressurization of RCS was also considered. In total, six types of RCS coolant loss scenarios were considered. Two cases were defined regarding the operation of the emergency diesel generators. Different delays of the pump

  17. Investigation of SAM measures during selected MBLOCA sequences along with Station Blackout in a generic Konvoi PWR using ASTECV2.0

    International Nuclear Information System (INIS)

    Gómez-García-Toraño, Ignacio; Sánchez Espinoza, Víctor Hugo; Stieglitz, Robert

    2017-01-01

    Highlights: • Reflooding is investigated for selected MBLOCA sequences in a Konvoi PWR using ASTEC. • After SBO, there is a grace time of 40 min up to the detection of a CET = 650 °C. • Major core damage prevented if reflood is launched at CET = 650 °C with 25-40 kg/s. • Values depend on the time when the plant is struck by Station Blackout. • Vessel failure cannot be prevented if supplied mass flow rates are lower than 10 kg/s. - Abstract: The Fukushima accidents have shown that further improvement of Severe Accident Management Guidelines (SAMGs) is necessary for the current fleet of Light Water Reactors. The elaboration of SAMGs requires a broad database of deterministic analyses performed with state-of-the art simulation tools. Within this work, the ASTECV2.0 integral severe accident code is used to study the efficiency of core reflooding (as a SAM measure) during postulated Medium Break LOCA (MBLOCA) scenarios in a German Konvoi PWR. In a first step, the progression of selected MBLOCA sequences without SAM measures has been analysed. The sequences postulate a break in the cold leg of the pressurizer loop and the total loss of AC power at a given stage of the accident. Results show the existence of a 40 min grace time up to the detection of a Core Exit Temperature (CET) of 650 °C providing that the AC power has been maintained at least 1 h after SCRAM. In a second step, an extensive analysis on core reflooding has been carried out. The sequences assume that the plant remains in Station Blackout (SBO) and that reflooding occurs at different times with different mobile pumps. The simulations yield the following results: • Reflooding mass flow rates above 60 kg/s have to be supplied as soon as the CET exceeds 650 °C in order to prevent core melting. • Reflooding mass flow rates ranging from 25–40 kg/s at CET = 650 °C mitigate the accident without major core damage depending on when the plant enters in SBO. • Reflooding mass flow rates lower

  18. Analyses of natural circulation during a Surry station blackout using SCDAP/RELAP5

    International Nuclear Information System (INIS)

    Bayless, P.D.

    1988-10-01

    The effects of reactor coolant system natural circulation on the response of the Surry nuclear power plant during a station blackout transient were investigated. A TMLB' sequence (loss of all ac power, immediate loss of auxillary feedwater) was simulated from transient initiation until after fuel rod relocation had begun. Integral analyses of the system thermal-hydraulics and the core damage behavior were performed using the SCDAP/RELAP5 computer code and several different models of the plant. Three scoping calculations were performed in which the complexity of the plant model was progressively increased to determine the overall effects of in-vessel and hot leg natural circulation flows on the plant response. The natural circulation flows extended the transient, slowing the core heatup and delaying core damage by transferring energy from the core to structures in the upper plenum and coolant loops. Increased temperatures in the ex-core structures indicated that they may fail, however. Nine sensitivity calculations were then performed to investigate the effects of modeling uncertainties on the multidimensional natural circulation flows and the system response. Creep rupture failure of the pressurizer surge line was predicted to occur in eight of the calculations, with the hot leg failing in the ninth. The failure time was fairly insensitive to the parameters varied. The failures occurred near the time that fuel rod relocation began, well before failure of the reactor vessel would be expected. A calculation was also performed in which creep rupture failure of the surge line was modeled. The subsequent blowdown led to rapid accumulator injection and quenching of the entire core. 18 refs., 105 figs., 17 tabs

  19. Mitigating fuel handling situations during station blackout in TAPP-3 and

    International Nuclear Information System (INIS)

    Chugh, V.K.; Roy, Shibaji; Gupta, H.; Inder Jit

    2002-01-01

    Full text: On power refueling is one of the important features of PHWRs. fuelling machine (FM) Head becomes part of the reactor pressure boundary during refueling operations. Hot irradiated (spent) fuel bundles are received in the FM Head from the Reactor and transferred to spent fuel storage bay (SFSB). These bundles pass through various fuel handling (FH) Equipment under submerged condition except during the dry transfer operation. Situations of station blackout (SBO) i.e. postulated simultaneous failure of Class III and Class IV electric power, could persist for a long period, during on-reactor or off-reactor FH operations, with the spent fuel bundles being any where in the system between the reactor and SFSB. The cooling provisions for the spent fuel bundles vary depending upon the stage of operation. During SBO, it becomes difficult to maintain cooling to these fuel bundles due to the limited availability of Class II power and instrument air. However, cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like stay-put, gravity- fill, D 2 O-steaming etc. for cooling the bundles. Various scenarios have been identified for cooling provisions of the bundles in the system. The paper also describes consequences like loss of D 2 O inventory, rise in ambient temperature and pressure and tritium build-up in Reactor Building, emanating from these cooling schemes

  20. Source term analysis in severe accident induced by large break loss of coolant accident coincident with ship blackout for ship reactor

    International Nuclear Information System (INIS)

    Zhang Yanzhao; Zhang Fan; Zhao Xinwen; Zheng Yingfeng

    2013-01-01

    Using MELCOR code, the accident analysis model was established for a ship reactor. The behaviors of radioactive fission products were analyzed in the case of severe accident induced by large break loss of coolant accident coincident with ship blackout. The research mainly focused on the behaviors of release, transport, retention and the final distribution of inert gas and CsI. The results show that 83.12% of inert gas releases from the core, and the most of inert gas exists in the containment. About 83.08% of CsI release from the core, 72.66% of which is detained in the debris and the primary system, and 27.34% releases into the containment. The results can give a reference for the evaluation of cabin dose and nuclear emergency management. (authors)

  1. Uncertainty and sensitivity analysis for the simulation of a station blackout scenario in the Jules Horowitz Reactor

    International Nuclear Information System (INIS)

    Ghione, Alberto; Noel, Brigitte; Vinai, Paolo; Demazière, Christophe

    2017-01-01

    Highlights: • A station blackout scenario in the Jules Horowitz Reactor is analyzed using CATHARE. • Input and model uncertainties relevant to the transient, are considered. • A statistical methodology for the propagation of the uncertainties is applied. • No safety criteria are exceeded and sufficiently large safety margins are estimated. • The most influential uncertainties are determined with a sensitivity analysis. - Abstract: An uncertainty and sensitivity analysis for the simulation of a station blackout scenario in the Jules Horowitz Reactor (JHR) is presented. The JHR is a new material testing reactor under construction at CEA on the Cadarache site, France. The thermal-hydraulic system code CATHARE is applied to investigate the response of the reactor system to the scenario. The uncertainty and sensitivity study was based on a statistical methodology for code uncertainty propagation, and the ‘Uncertainty and Sensitivity’ platform URANIE was used. Accordingly, the input uncertainties relevant to the transient, were identified, quantified, and propagated to the code output. The results show that the safety criteria are not exceeded and sufficiently large safety margins exist. In addition, the most influential input uncertainties on the safety parameters were found by making use of a sensitivity analysis.

  2. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout Caused by External Flooding Using the RISMC Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools. This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.

  3. Additional examination on station blackout caused by tsunami in Fukushima Daiichi NPS

    International Nuclear Information System (INIS)

    Yamauchi, Daisuke; Date, Kenji; Mizokami, Masato; Honda, Takeshi; Nozaki, Kenichiro; Mizokami, Shinya; Endo, Ryohei

    2017-01-01

    This study is additional examination to verify in a more reliable way the assessment that the emergency AC power supply was lost due to tsunami at Fukushima Daiichi Nuclear Power Station. It confirmed the relationship between the path length of tsunami intrusion reaching each power supply facility and the function loss time. As a result of examination, it was confirmed that as the path length of the tsunami intrusion reaching each power supply facility was longer, the function loss time tended to be later. So, conventional assessment that the function of each power supply facility was lost due to the run-up and flooding of tsunami has become more probable. For facilities, where the overall trend and the loss time of function were divergent, it was found that there were scenarios that could reasonably be explained. Based on the fact that the Fukushima Daiichi Nuclear Power Station lost power due to the tsunami, Kashiwazaki-Kariwa Nuclear Power Station carries out various safety measures. First of all, as the measures to prevent accidents caused by tsunami, the following have been applied: (1) prevention of the inflow of tsunamis into premises, (2) water prevention of the areas installed with important equipment, (3) securement of seawater at the time of backwashing, (4) storage of portable equipment at high ground, (5) installation of tsunami surveillance cameras. To prepare for the loss of power supply, this station implemented power supply facilities such as generator cars and distribution boards, as well as the placement of power supply cars at high ground. (A.O.)

  4. THE APPLICATION OF MAMMOTH FOR A DETAILED TIGHTLY COUPLED FUEL PIN SIMULATION WITH A STATION BLACKOUT

    Energy Technology Data Exchange (ETDEWEB)

    Gleicher, Frederick; Ortensi, Javier; DeHart, Mark; Wang, Yaqi; Schunert, Sebastian; Novascone, Stephen; Hales, Jason; Williamson, Rich; Slaughter, Andrew; Permann, Cody; Andrs, David; Martineau, Richard

    2016-09-01

    Accurate calculation of desired quantities to predict fuel behavior requires the solution of interlinked equations representing different physics. Traditional fuels performance codes often rely on internal empirical models for the pin power density and a simplified boundary condition on the cladding edge. These simplifications are performed because of the difficulty of coupling applications or codes on differing domains and mapping the required data. To demonstrate an approach closer to first principles, the neutronics application Rattlesnake and the thermal hydraulics application RELAP-7 were coupled to the fuels performance application BISON under the master application MAMMOTH. A single fuel pin was modeled based on the dimensions of a Westinghouse 17x17 fuel rod. The simulation consisted of a depletion period of 1343 days, roughly equal to three full operating cycles, followed by a station blackout (SBO) event. The fuel rod was depleted for 1343 days for a near constant total power loading of 65.81 kW. After 1343 days the fission power was reduced to zero (simulating a reactor shut-down). Decay heat calculations provided the time-varying energy source after this time. For this problem, Rattlesnake, BISON, and RELAP-7 are coupled under MAMMOTH in a split operator approach. Each system solves its physics on a separate mesh and, for RELAP-7 and BISON, on only a subset of the full problem domain. Rattlesnake solves the neutronics over the whole domain that includes the fuel, cladding, gaps, water, and top and bottom rod holders. Here BISON is applied to the fuel and cladding with a 2D axi-symmetric domain, and RELAP-7 is applied to the flow of the circular outer water channel with a set of 1D flow equations. The mesh on the Rattlesnake side can either be 3D (for low order transport) or 2D (for diffusion). BISON has a matching ring structure mesh for the fuel so both the power density and local burn up are copied accurately from Rattlesnake. At each depletion time

  5. Economic simplified boiling water reactor (ESBWR) response to an extended station blackout/ loss of all AC power

    International Nuclear Information System (INIS)

    Barrett, A.J.; Marquino, W.

    2013-01-01

    U.S. federal regulations require light water cooled nuclear power plants to cope with Station Blackout for a predetermined amount of time based on design factors for the plant. U.S. regulations define Station Blackout (SBO) as a loss of the offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency AC power system. According to U.S. regulations, typically the coping period for an SBO is 4 hours and can be as long as 16 hours for currently operating BWR plants. Being able to cope with an SBO and loss of all AC power is required by international regulators as well. The U.S. licensing basis for the ESBWR is a coping period of 72 hours for an SBO based on U.S. NRC requirements for passive safety plants. In the event of an extended SBO (viz., greater than 72 hours), the ESBWR response shows that the design is able to cope with the event for at least 7 days without AC electrical power or operator action. ESBWR is a Generation III+ reactor design with an array of passive safety systems. The ESBWR primary success path for mitigation of an SBO event is the Isolation Condenser System (ICS). The ICS is a passive, closed loop, safety system that initiates automatically on a loss of power. Upon Station Blackout or loss of all AC power, the ICS begins removing decay heat from the Reactor Pressure Vessel (RPV) by (i) condensing the steam into water in heat exchangers located in pools of water above the containment, and (ii) transferring the decay heat to the atmosphere. The condensed water is then returned by gravity to cool the reactor again. The ICS alone is capable of maintaining the ESBWR in a safe shutdown condition after an SBO for an extended period. The fuel remains covered throughout the SBO event. The ICS is able to remove decay heat from the RPV for at least 7 days and maintains the reactor in a safe shutdown condition. The water level in the RPV remains well above the top of active fuel for the duration of the SBO event

  6. Severe Accident Test Station Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A [ORNL; Terrani, Kurt A [ORNL

    2015-06-01

    Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.

  7. Analysis of fuel handling system for fuel bundle safety during station blackout in 500 MWe PHWR unit of India

    Energy Technology Data Exchange (ETDEWEB)

    Madhuresh, R; Nagarajan, R; Jit, I; Sanatkumar, A [Nuclear Power Corporation of India Ltd., Mumbai (India)

    1997-12-31

    Situations of Station Blackout (SBO) i.e. postulated concurrent unavailability of Class Ill and Class IV power, could arise for a long period, while on-power refuelling or other fuel handling operations are in progress with the hot irradiated fuel bundles being anywhere in the system from the Reactor Building to the Spent Fuel Storage Bay. The cooling provisions for these fuel bundles are diverse and specific to the various stages of fuel handling operations and are either on Class Ill or on Class II power with particular requirements of instrument air. Therefore, during SBO, due to the limited availability of Class II power and instrument air, it becomes difficult to maintain cooling to these fuel bundles. However, some minimal cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like `stay-put`, `gravity- fill`, `D{sub 2}0- steaming` etc. for cooling the bundles. The paper also describes various consequences emanating from these cooling schemes. (author). 6 refs., 2 tabs., 8 figs.

  8. A portable backup power supply to assure extended decay heat removal during natural phenomena-induced station blackout

    International Nuclear Information System (INIS)

    Proctor, L.D.; Merryman, L.D.; Sallee, W.E.

    1989-01-01

    The High Flux Isotope Reactor (HFIR) is a light water cooled and moderated flux-trap type research reactor located at Oak Ridge National Laboratory (ORNL). Coolant circulation following reactor shutdown is provided by the primary coolant pumps. DC-powered pony motors drive these pumps at a reduced flow rate following shutdown of the normal ac-powered motors. Forced circulation decay heat removal is required for several hours to preclude core damage following shutdown. Recent analyses identified a potential vulnerability due to a natural phenomena-induced station blackout. Neither the offsire power supply nor the onsite emergency diesel generators are designed to withstand the effects of seismic events or tornadoes. It could not be assured that the capacity of the dedicated batteries provided as a backup power supply for the primary coolant pump pony motors is adequate to provide forced circulation cooling for the required time following such events. A portable backup power supply added to the plant to address this potential vulnerability is described

  9. Analysis of fuel handling system for fuel bundle safety during station blackout in 500 MWe PHWR unit of India

    International Nuclear Information System (INIS)

    Madhuresh, R.; Nagarajan, R.; Jit, I.; Sanatkumar, A.

    1996-01-01

    Situations of Station Blackout (SBO) i.e. postulated concurrent unavailability of Class Ill and Class IV power, could arise for a long period, while on-power refuelling or other fuel handling operations are in progress with the hot irradiated fuel bundles being anywhere in the system from the Reactor Building to the Spent Fuel Storage Bay. The cooling provisions for these fuel bundles are diverse and specific to the various stages of fuel handling operations and are either on Class Ill or on Class II power with particular requirements of instrument air. Therefore, during SBO, due to the limited availability of Class II power and instrument air, it becomes difficult to maintain cooling to these fuel bundles. However, some minimal cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like 'stay-put', 'gravity- fill', 'D 2 0- steaming' etc. for cooling the bundles. The paper also describes various consequences emanating from these cooling schemes. (author). 6 refs., 2 tabs., 8 figs

  10. Development of Abnormal Operating Strategies for Station Blackout in Shutdown Operating Mode in Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Duk-Joo; Lee, Seung-Chan; Sung, Je-Joong; Ha, Sang-Jun [KHNP CRI, Daejeon (Korea, Republic of); Hwang, Su-Hyun [FNC Tech. Co., Yongin (Korea, Republic of)

    2016-10-15

    Loss of all AC power is classified as one of multiple failure accident by regulatory guide of Korean accident management program. Therefore we need develop strategies for the abnormal operating procedure both of power operating and shutdown mode. This paper developed abnormal operating guideline for loss of all AC power by analysis of accident scenario in pressurized water reactor. This paper analyzed the loss of ultimate heat sink (LOUHS) in shutdown operating mode and developed the operating strategy of the abnormal procedure. Also we performed the analysis of limiting scenarios that operator actions are not taken in shutdown LOUHS. Therefore, we verified the plant behavior and decided operator action to taken in time in order to protect the fuel of core with safety. From the analysis results of LOUHS, the fuel of core maintained without core uncovery for 73 minutes respectively for opened RCS states after the SBO occurred. Therefore, operator action for the emergency are required to take in 73 minutes for opened RCS state. Strategy is to cooldown by using spent fuel pool cooling system. This method required to change the plant design in some plant. In RCS boundary closed state, first abnormal operating strategy in shutdown LOUHS is first abnormal operating strategy in shutdown LOUHS is to remove the residual heat of core by steam dump flow and auxiliary feedwater of SG.

  11. Severe Accident Test Station Design Document

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Keiser, James R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

  12. Severe Accident Test Station Design Document

    International Nuclear Information System (INIS)

    Snead, Mary A.; Yan, Yong; Howell, Michael; Keiser, James R.; Terrani, Kurt A.

    2015-01-01

    The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

  13. Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego; Prescott, Steven R; Smith, Curtis L; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua J; Kinoshita, Robert A

    2011-07-01

    In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.

  14. Thermal-hydraulic analysis of NSSS and containment response during extended station blackout for Maanshan PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Hsu, Keng-Hsien, E-mail: hardlycampus@iner.gov.tw; Lin, Chin-Tsu, E-mail: jtling@iner.gov.tw

    2015-07-15

    Highlights: • Calculate NSSS and containment transient response during extended SBO of 24 h. • RELAP5-3D and GOTHIC models are developed for Maanshan PWR plant. • Reactor coolant pump seal leakage is specifically modeled for each loop. • Analyses are performed with and without secondary-side depressurization, respectively. • Considering different total available time for turbine driven auxiliary feedwater system. - Abstract: A thermal-hydraulic analysis has been performed with respect to the response of the nuclear steam supply system (NSSS) and the containment during an extended station blackout (SBO) duration of 24 h in Maanshan PWR plant. Maanshan plant is a Westinghouse three-loop PWR design with rated core thermal power of 2822 MWt. The analyses in the NSSS and the containment are based on the RELAP5-3D and GOTHIC models, respectively. Important design features of the plant in response to SBO are considered in the respective models, e.g., the steam generator PORVs, turbine driven auxiliary feedwater system (TDAFWS), accumulators, reactor coolant pump (RCP) seal design, various heat structures in the containment, etc. In the analysis it is assumed that the shaft seal in each RCP failed due to loss of seal cooling and the RCS fluid flows to the containment directly. Some parameters calculated from the RELPA5-3D model are input to the containment GOTHIC model, including the RCS average temperature and the RCP seal leakage flow and enthalpy. The RCS average temperature is used to drive the sensible heat transfer to the containment. It is found that the severity of the event depends mainly on whether the secondary side is depressurized or not. If the secondary side is depressurized in time (within 1 h after SBO) and the TDAFWS is available greater than 19 h, then the reactor core will be covered with water throughout the SBO duration, which ensures the integrity of the reactor core. On the contrary, if the secondary side is not depressurized, then the RCS

  15. Report on the accident at the Chernobyl Nuclear Power Station

    International Nuclear Information System (INIS)

    1987-01-01

    This report presents the compilation of information obtained by various organizations regarding the accident (and the consequences of the accident) that occurred at Unit 4 of the nuclear power station at Chernobyl in the USSR on April 26, 1986. The various authors are identified in a footnote to each chapter. An overview of the report is provided. Very briefly the other chapters cover: the design of the Chernobyl nuclear station Unit 4; safety analyses for Unit 4; the accident scenario; the role of the operator; an assessment of the radioactive release, dispersion, and transport; the activities associated with emergency actions; and information on the health and environmental consequences from the accident. These subjects cover the major aspects of the accident that have the potential to present new information and lessons for the nuclear industry in general

  16. Alternative cooling water flow path for RHR heat exchanger and its effect on containment response during extended station blackout for Chinshan BWR-4 plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

    2016-04-15

    Highlights: • Motivating alternative RHR heat exchanger tube-side flow path and determining required capacity. • Calculate NSSS and containment response during 24-h SBO for Chinshan BWR-4 plant. • RETRAN and GOTHIC models are developed for NSSS and containment, respectively. • Safety relief valve blowdown flow and energy to drywell are generated by RETRAN. • Analyses are performed with and without reactor depressurization, respectively. - Abstract: The extended Station Blackout (SBO) of 24 h has been analyzed with respect to the containment response, in particular the suppression pool temperature response, for the Chinshan BWR-4 plant of MARK-I containment. The Chinshan plant, owned by Taiwan Power Company, has twin units with rated core thermal power of 1840 MW each. The analysis is aimed at determining the required alternative cooling water flow capacity for the residual heat removal (RHR) heat exchanger when its tube-side sea water cooling flow path is blocked, due to some reason such as earthquake or tsunami, and is switched to the alternative raw water source. Energy will be dissipated to the suppression pool through safety relief valves (SRVs) of the main steam lines during SBO. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The shaft seals of the two recirculation pumps are conservatively assumed to fail due to loss of seal cooling and a total leakage flow rate of 36 gpm to the drywell is included in the GOTHIC model. Based on the given SRV blowdown conditions, the GOTHIC containment calculation is performed several times, through the adjustment of the heat transfer rate of the RHR heat exchanger, until the criterion that the maximum suppression pool temperature

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

  19. Seabrook Station Level 2 PRA Update to Include Accident Management

    International Nuclear Information System (INIS)

    Lutz, Robert; Lucci, Melissa; Kiper, Kenneth; Henry, Robert

    2006-01-01

    A ground-breaking study was recently completed as part of the Seabrook Level 2 PRA update. This study updates the post-core damage phenomena to be consistent with the most recent information and includes accident management activities that should be modeled in the Level 2 PRA. Overall, the result is a Level 2 PRA that fully meets the requirements of the ASME PRA Standard with respect to modeling accident management in the LERF assessment and NRC requirements in Regulatory Guide 1.174 for considering late containment failures. This technical paper deals only with the incorporation of operator actions into the Level 2 PRA based on a comprehensive study of the Seabrook Station accident response procedures and guidance. The paper describes the process used to identify the key operator actions that can influence the Level 2 PRA results and the development of success criteria for these key operator actions. This addresses a key requirement of the ASME PRA Standard for considering SAMG. An important benefit of this assessment was the identification of Seabrook specific accident management insights that can be fed back into the Seabrook Station accident management procedures and guidance or the training provided to plant personnel for these procedures and guidance. (authors)

  20. Report on the accident at the Chernobyl Nuclear Power Station

    International Nuclear Information System (INIS)

    1987-12-01

    This report presents the compilation of information obtained by various organizations regarding the accident (and the consequences of the accident) that occurred at Unit 4 of the nuclear power station at Chernobyl in the USSR on April 26, 1986. Each organization has independently accepted responsibility for one or more chapters. The specific responsibility of each organization is indicated. The various authors are identified in a footnote to each chapter. Very briefly the other chapters cover: the design of the Chernobyl nuclear station Unit 4; safety analyses for Unit 4; the accident scenario; the role of the operator; an assessment of the radioactive release, dispersion, and transport; the activities associated with emergency actions; and information on the health and environmental consequences from the accident. These subjects cover the major aspects of the accident that have the potential to present new information and lessons for the nuclear industry in general. The task of evaluating the information obtained in these various areas and the assessment of the potential implications has been left to each organization to pursue according to the relevance of the subject to their organization. Those findings will be issued separately by the cognizant organizations. The basic purpose of this report is to provide the information upon which such assessments can be made

  1. A study for the establishment of regulatory requirement and evaluation guide for station blackout in nuclear power plants

    International Nuclear Information System (INIS)

    Lim, J. H.; Koo, C. S.; Joo, W. P.; Oh, S. H.; Shin, W. K.

    1999-01-01

    The consequence of SBO event could be a severe accident unless AC power was restored within a proper time, because many safety systems depend upon AC power. Based on the severity, the SBO has been extensively studied since it was identified as Unresolved Safety Issue at USNRC. The resolution of those studies is a rule-making such as 10 CFR 50.63 and Regulatory Guide 1.155. But there is no regulatory requirements of SBO for an operating domestic nuclear power plant up to the present time. This tudy has established SBO rule(regulatory requirements and evaluation guides) for an operating PWR type of the operating nuclear power plants in Korea

  2. Steam Oxidation Testing in the Severe Accident Test Station

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    After the March 2011 accident at Fukushima Daiichi, Oak Ridge National Laboratory (ORNL) began conducting high temperature steam oxidation testing of candidate materials for accident tolerant fuel (ATF) cladding in August 2011 [1-11]. The ATF concept is to enhance safety margins in light water reactors (LWR) during severe accident scenarios by identifying materials with 100× slower steam oxidation rates compared to current Zr-based alloys. In 2012, the ORNL laboratory equipment was expanded and made available to the entire ATF community as the Severe Accident Test Station (SATS) [4,12]. Compared to the current UO2/Zr-based alloy fuel system, an ATF alternative would significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [13-14]. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models [15-17]. However, initial modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. Also, because many accident scenarios include steadily increasing temperatures, the required data are not traditional isothermal exposures but exposures with varying “ramp” rates. In some cases, the steam oxidation behavior has been surprising and difficult to interpret. Thus, more fundamental information continues to be collected. In addition, more work continues to focus on commercially-manufactured tube material. This report summarizes recent work to characterize the behavior of candidate alloys exposed to high temperature steam, evaluate steam oxidation behavior in various ramp scenarios and continue to collect integral data on FeCrAl compared to conventional Zr-based cladding.

  3. A station blackout simulation for the Advanced Neutron Source Reactor using the integrated primary and secondary system model

    International Nuclear Information System (INIS)

    Schneider, E.A.

    1994-01-01

    The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at Oak Ridge National Laboratory. This paper deals with thermal-hydraulic analysis of ANSR's cooling systems during nominal and transient conditions, with the major effort focusing upon the construction and testing of computer models of the reactor's primary, secondary and reflector vessel cooling systems. The code RELAP5 was used to simulate transients, such as loss of coolant accidents and loss of off-site power, as well as to model the behavior of the reactor in steady state. Three stages are involved in constructing and using a RELAP5 model: (1) construction and encoding of the desired model, (2) testing and adjustment of the model until a satisfactory steady state is achieved, and (3) running actual transients using the steady-state results obtained earlier as initial conditions. By use of the ANSR design specifications, a model of the reactor's primary and secondary cooling systems has been constructed to run a transient simulating a loss of off-site power. This incident assumes a pump coastdown in both the primary and secondary loops. The results determine whether the reactor can survive the transition from forced convection to natural circulation

  4. Lessons drawn from serious accidents in nuclear power stations

    International Nuclear Information System (INIS)

    Kosciusko-Morizet, F.; Tanguy, P.

    1981-01-01

    Taking a number of serious accidents considered to be particularly representative (Windscale, Enrico Fermi, Lucens, Browns Ferry, Three Mile Island and Saint-Laurent-des-Eaux), the paper analyses the conclusions reached in subsequent enquiries and the lessons drawn from them by the responsible authorities. While design problems sometimes come to light, it is much more generally operational safety - problems related to instructions, the training of operators, the man/machine relationship - which appears to be inadequate. The organization of relations between the different partners - builders, operators and safety bodies - likewise gives rise to some observations. Certain measures should be pursued on a broader scale in order to improve our ability to prevent serious accidents: (i) incidents important from the standpoint of safety must be identified; (ii) these incidents must be brought to the knowledge of all partners concerned, in all interested countries; (iii) the lessons drawn from them must be exchanged and compared; and (iv) the lessons must be made generally available in a directly usable form (i.e. as design modifications, changes in instructions and so on). Particular attention must be given to the problems of countries which are embarking on nuclear programmes and which, with a small number of installations, need direct and permanent access to all the lessons drawn from the operation of a large power station park, and must be able to call upon the assistance of teams from outside in the event of an accident. (author)

  5. Summary of the Current Status of Lessons Learned From Fukushima Accident

    International Nuclear Information System (INIS)

    Pasamehmetoglu, Kemal

    2013-01-01

    This presentation introduced the current status of the lessons learned from the Fukushima accident, and in particular, the recommendations released by a NRC Near-term Task Force to enhance reactor safety in the 21. century. The near-term recommendations are focused on emergency power and emergency cooling availability during station blackout accidents

  6. Accident assessment under emergency situation in Daya Bay nuclear power station

    International Nuclear Information System (INIS)

    Yang Ling; Chen Degan; Lin Shumou; Fu Guohui

    2004-01-01

    The accident assessment under emergency situation includes the accident status evaluation and its consequence estimation. This paper introduces evaluation methods for accident status and its assistant computer system (SESAME-GNP) utilized during the emergency situation in Guangdong Daya Bay Nuclear Power Station (GNPS) in detail. At the same time, an improved accident consequence estimation system in GNPS (RACAS-GNP) is briefly described. With the improvement of the accident assessment systems, the capability of emergency response in GNPS is strengthened

  7. 77 FR 16175 - Station Blackout

    Science.gov (United States)

    2012-03-20

    ... of the near- term actions based on lessons-learned stemming from the March 2011 Fukushima Dai-ichi... http://www.regulations.gov and search for documents filed under Docket ID NRC- 2011-0299. Address.... Fukushima Dai-ichi Event and the NRC Regulatory Response III. Background A. General Design Criteria 2 and 17...

  8. Lessons learned from our accident at Fukushima nuclear power stations

    International Nuclear Information System (INIS)

    Kawano, A.

    2012-01-01

    This paper is given in order to share the detailed information on the Fukushima Accident which occurred on March 11, 2011, and the lessons learned from it which worldwide nuclear experts might currently have more interest in. The paper first reflects how the facilities were damaged by a very strong earthquake and a series of beyond design-basis tsunamis. The earthquake caused loss of all off-site electric power at Fukushima Daiichi Nuclear Power Station (1F), and the following series of tsunami made all emergency diesel generators except one for Unit 6 and most of DC batteries inoperable and severely damaged most of the facilities located on the ocean side. Thus all the units at 1F resulted in the loss of cooling function and ultimate heat sink for a long time period. TEPCO focused on restoration of the instruments and lights in the Main Control Room (MCR), preparation of alternative water injection and venting of Primary Containment Vessel (PCV) in the recovery process. However, the workers faced a lot of difficulties such as total darkness, repeated aftershocks, high radiation dose, a lot of debris on the ground, loss of communication means, etc. Massive damages by the tsunami and lack of necessary equipments and resources hampered a quick recovery. It eventually resulted in the severe core damage of Unit 1, 2, and 3 and also the hydrogen explosions in the reactor buildings of Unit 1, 3, and 4. This paper finally extracts the lessons learned from the accident and proposes the countermeasures, such as flood protection for essential facilities, preparation of practical and effective tools, securing communication means and so on. These would help the people involved in the nuclear industries all over the world properly understand the accident and develop their own countermeasures appropriately. (authors)

  9. The accident at TEPCO's Fukushima Dai-ichi Nuclear Power Station - occurrence of the accident, current situation and Future

    International Nuclear Information System (INIS)

    Hirose, K.

    2013-01-01

    In this presentation author analyse course of accident on Fukushima Dai-chi NPPs as well as consequences of this disaster. The following parts are presented: (1) Occurrence of the accident; (2) Evacuation of the residential people; (3) Deterioration and protraction of the accident; (4) Impact on society; (5) Situation of decontamination; (6) Long-term steps towards decommissioning; (7) Situation of other nuclear power stations; (8) Conclusions and lessons learned.

  10. Extended blackout mitigation strategy for PWR

    International Nuclear Information System (INIS)

    Prošek, Andrej; Volkanovski, Andrija

    2015-01-01

    Highlights: • Equipment for mitigation of the extended blackout is investigated. • Analysis is done with deterministic safety analysis methods. • Strategy to prevent core heatup and not overfill steam generator is proposed. • Six types of reactor coolant system loss scenarios are investigated. • Pump flowrates and available start time to feed steam generators is determined. - Abstract: The accident at the Fukushima Daiichi nuclear power plant demonstrated the vulnerability of the plants on the loss of electrical power and loss of the ultimate heat sink events. A set of measures are proposed and currently implemented in response of the accident at the Fukushima Daiichi nuclear power plant. Those measures include diverse and flexible mitigation strategies that increase the defence-in-depth for beyond-design-basis scenarios. Mitigation strategies are based on the utilization of the portable equipment to provide power and water to the nuclear power plants in order to maintain or restore key safety functions. The verification of the proposed measures with the plant specific safety analyses is endorsed in the mitigation strategies. This paper investigates utilization of the turbine driven auxiliary feedwater pump (TD-AFW) or portable water injection pump for the mitigation of the event of loss of all alternate current sources and batteries (extended station blackout). Methodology for assessment of the required pump injection flow rate with the application of the standard deterministic safety analysis code is developed and presented. The required injection rate to the steam generators is calculated from the cumulative water mass injected by the turbine driven auxiliary feedwater pump in the analysed scenarios, when desired normal level is maintained automatically. The developed methodology allows assessment of the required injections rates of pump, TD-AFW or portable, for different plant configurations and number of flow rate changes. The methodology is applied

  11. Extended blackout mitigation strategy for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Prošek, Andrej, E-mail: andrej.prosek@ijs.si; Volkanovski, Andrija, E-mail: andrija.volkanovski@ijs.si

    2015-12-15

    Highlights: • Equipment for mitigation of the extended blackout is investigated. • Analysis is done with deterministic safety analysis methods. • Strategy to prevent core heatup and not overfill steam generator is proposed. • Six types of reactor coolant system loss scenarios are investigated. • Pump flowrates and available start time to feed steam generators is determined. - Abstract: The accident at the Fukushima Daiichi nuclear power plant demonstrated the vulnerability of the plants on the loss of electrical power and loss of the ultimate heat sink events. A set of measures are proposed and currently implemented in response of the accident at the Fukushima Daiichi nuclear power plant. Those measures include diverse and flexible mitigation strategies that increase the defence-in-depth for beyond-design-basis scenarios. Mitigation strategies are based on the utilization of the portable equipment to provide power and water to the nuclear power plants in order to maintain or restore key safety functions. The verification of the proposed measures with the plant specific safety analyses is endorsed in the mitigation strategies. This paper investigates utilization of the turbine driven auxiliary feedwater pump (TD-AFW) or portable water injection pump for the mitigation of the event of loss of all alternate current sources and batteries (extended station blackout). Methodology for assessment of the required pump injection flow rate with the application of the standard deterministic safety analysis code is developed and presented. The required injection rate to the steam generators is calculated from the cumulative water mass injected by the turbine driven auxiliary feedwater pump in the analysed scenarios, when desired normal level is maintained automatically. The developed methodology allows assessment of the required injections rates of pump, TD-AFW or portable, for different plant configurations and number of flow rate changes. The methodology is applied

  12. Analysis of station blackout accident at the nuclear power plant; Analiza gubitka svih izmjenichnih izvora napajanja u NE

    Energy Technology Data Exchange (ETDEWEB)

    Bencik, V [Elektrotehnichki Inst. Rade Koncar, Zagreb (Yugoslavia); Kozaric, M [Zagreb Univ. (Yugoslavia). Elektrotehnicki Fakultet

    1990-07-01

    In 1986 the US NRC published a proposed rule making to tighten the requirements on nuclear plants with regard to their ability to deal safely with a total loss of all a-c electric power, both from external and internal sources of supply. The proposed rule would require all licensees and applicants to: 1. Assess the capability of their plants to cope with a total loss all a-c power(that is, determine the amount of time the plant could maintain core cooling and containment integrity with a-c power unavailable); 2. Have procedures and training to cope with such an event; 3. Make modification, if necessary, to cope with an acceptable minimum duration loss of all a-c power. A total loss of all a-c electric power has been identified as an 'unresolved safety issue' and subjected to considerably study. The article presents an idea to resolve this issue. (author)

  13. Alcohol-Induced Blackout

    Directory of Open Access Journals (Sweden)

    Dai Jin Kim

    2009-11-01

    Full Text Available For a long time, alcohol was thought to exert a general depressant effect on the central nervous system (CNS. However, currently the consensus is that specific regions of the brain are selectively vulnerable to the acute effects of alcohol. An alcohol-induced blackout is the classic example; the subject is temporarily unable to form new long-term memories while relatively maintaining other skills such as talking or even driving. A recent study showed that alcohol can cause retrograde memory impairment, that is, blackouts due to retrieval impairments as well as those due to deficits in encoding. Alcoholic blackouts may be complete (en bloc or partial (fragmentary depending on severity of memory impairment. In fragmentary blackouts, cueing often aids recall. Memory impairment during acute intoxication involves dysfunction of episodic memory, a type of memory encoded with spatial and social context. Recent studies have shown that there are multiple memory systems supported by discrete brain regions, and the acute effects of alcohol on learning and memory may result from alteration of the hippocampus and related structures on a cellular level. A rapid increase in blood alcohol concentration (BAC is most consistently associated with the likelihood of a blackout. However, not all subjects experience blackouts, implying that genetic factors play a role in determining CNS vulnerability to the effects of alcohol. This factor may predispose an individual to alcoholism, as altered memory function during intoxication may affect an individual‟s alcohol expectancy; one may perceive positive aspects of intoxication while unintentionally ignoring the negative aspects. Extensive research on memory and learning as well as findings related to the acute effects of alcohol on the brain may elucidate the mechanisms and impact associated with the alcohol- induced blackout.

  14. Main results of assessing integrity of RNPP-3 steam generator heat exchange tubes in accident management

    International Nuclear Information System (INIS)

    Shugajlo, Al-j P.; Mustafin, M.A.; Shugajlo, Al-r P.; Ryzhov, D.I.; Zhabin, O.I.

    2017-01-01

    Tubes integrity evaluation under accident conditions considering drain of SG and current technical state of steam exchange tubes is an important question regarding SG long-term operation and improvement of accident management strategy.The main investigation results prepared for heat exchange surface of RNPP-3 steam generator are presented in this research aimed at assessing integrity of heat exchange tubes under accident conditions, which lead to full or partial drain of heat exchange surface, in particular during station blackout.

  15. Risk Perception and Occupational Accidents: A Study of Gas Station Workers in Southern Brazil

    OpenAIRE

    Cezar-Vaz, Marta Regina; Rocha, Laurelize Pereira; Bonow, Clarice Alves; da Silva, Mara Regina Santos; Vaz, Joana Cezar; Cardoso, Letícia Silveira

    2012-01-01

    The present study aimed to identify the perceptions of gas station workers about physical, chemical, biological and physiological risk factors to which they are exposed in their work environment; identify types of occupational accidents involving gas station workers and; report the development of a socioenvironmental intervention as a tool for risk communication to gas station workers. A quantitative study was performed with 221 gas station workers in southern Brazil between October and Decem...

  16. MELCOR DB Construction for the Severe Accident Analysis DB

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  17. 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

  18. Risk perception and occupational accidents: a study of gas station workers in southern Brazil.

    Science.gov (United States)

    Cezar-Vaz, Marta Regina; Rocha, Laurelize Pereira; Bonow, Clarice Alves; Silva, Mara Regina Santos da; Vaz, Joana Cezar; Cardoso, Letícia Silveira

    2012-07-01

    The present study aimed to identify the perceptions of gas station workers about physical, chemical, biological and physiological risk factors to which they are exposed in their work environment; identify types of occupational accidents involving gas station workers and; report the development of a socioenvironmental intervention as a tool for risk communication to gas station workers. A quantitative study was performed with 221 gas station workers in southern Brazil between October and December 2010. Data collection was performed between October to December 2010 via structured interviews. The data were analyzed using SPSS 19.0. The participants identified the following risk types: chemical (93.7%), physical (88.2%), physiological (64.3%) and biological (62.4%). In this sample, 94.1% of gas station workers reported occupational accidents, and 74.2% reported fuel contact with the eyes (p accidents as an indicator of the dangerous nature of their work environment.

  19. On preparation for accident management in LWR power stations

    International Nuclear Information System (INIS)

    1996-01-01

    Nuclear Safety Commission received the report from Reactor Safety General Examination Committee which investigated the policy of executing the preparation for accident management. The basic policy on the preparation for accident management was decided by Nuclear Safety Commission in May, 1992. This Examination Committee investigated the policy of executing the preparation for accident management, which had been reported from the administrative office, and as the result, it judged the policy as adequate, therefore, the report is made. The course to the foundation of subcommittee is reported. The basic policy of the examination on accident management by the subcommittee conforming to the decision by Nuclear Safety Commission, the measures of accident management which were extracted for BWR and PWR facilities, the examination of the technical adequacy of selecting accident sequences in BWR and PWR facilities and the countermeasures to them, the adequacy of the evaluation of the possibility of executing accident management measures and their effectiveness and the adequacy of the evaluation of effect to existing safety functions, the preparation of operation procedure manual, and education and training plan are reported. (K.I.)

  20. The accident prevention regulation 'Thermal Power Stations' and its effects in practice

    International Nuclear Information System (INIS)

    Albert, O.

    1983-01-01

    The origin of the accident prevention regulation - ''Thermal Power Stations'' is attributable mainly to two tragic accidents. It has made organizational changes and interventions in the operational process necessary in thermal power stations. Emphasis is laid upon the consistent issue of written permits-to-work on plant components carrying a heating medium and operating under pressure and on written operating licences for the operation of boilers. The paper describes additional ways in which regulation influences the daily practices of the power station operator. Brief references is made to the draft of the revised regulation. (orig./HP) [de

  1. Accident analysis of Fukushima Daiichi Nuclear Power Station unit 1

    International Nuclear Information System (INIS)

    Kobayashi, Masahide; Narabayashi, Tadashi; Tsuji, Masashi; Chiba, Go; Nagata, Yasunori; Shimoe, Tomohiro

    2015-01-01

    As a result of the Great East Japan Earthquake that occurred on 11 March 2011, all AC and DC power at the Fukushima Daiichi NPP units 1 to 3 were lost soon after the tsunami. The core cooling function was lost, and the cores of units 1 to 3 were damaged. The purpose of this work is to clarify the progress of the accident in unit 1, which was damaged the earliest among the 3 units. Therefore, an original severe accident analysis code was developed, and the progress of the accident was evaluated from the analysis results and the actual data. As a result, the leakage path from a pressure vessel was clarified, and some lessons and knowledge were gained. (author)

  2. Calculation of spent fuel pool severe accident with MELCOR

    International Nuclear Information System (INIS)

    Deng Jian; Xiang Qing'an; Zhou Kefeng

    2014-01-01

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

  3. In-depth investigation of escalator riding accidents in heavy capacity MRT stations.

    Science.gov (United States)

    Chi, Chia-Fen; Chang, Tin-Chang; Tsou, Chi-Lin

    2006-07-01

    In 2000, the accident rate for escalator riding was about 0.815 accidents per million passenger trips through Taipei Metro Rapid Transit (MRT) heavy capacity stations. In order to reduce the probability and severity of escalator riding accidents and enhance the safety of passengers, the Drury and Brill model [Drury, C.G., Brill, M., 1983. Human factors in consumer product accident investigation. Hum. Factors 25 (3), 329-342] for in-depth investigation was adopted to analyze the 194 escalator riding accidents in terms of victim, task, product and environment. Prevention measures have been developed based on the major causes of accidents and other related contributing factors. The results from the analysis indicated that the majority of the escalator riding accidents was caused by passengers' carrying out other tasks (38 cases, including carrying luggage 24 cases, looking after accompany persons 9 cases, and 5 others), loss of balance (26 cases, 13.4%), not holding the handrail (20 cases, 10.3%), unhealthy passengers (18 cases, 9.3%), followed by people struck by other passenger (16 cases, 8.2%). For female passengers aged 15-64 years, their rushing for trains accidents could have been prevented by wearing safer footwear or by appropriate signing being provided indicating the location and traveling direction of escalators. Female passengers aged 65 years and above whose accidents were caused by loss of balance, should be encouraged to take the elevator instead. To prevent entrapment injuries, following a stricter design code can be most effective. Further in-depth accident investigation is suggested to cover the activity of the victim prior to the accident, any involved product, the location of the accident on the escalator, any medical treatment, what went wrong, opinion of the respondent on the causes of the accident, and personal characteristics of the passengers. Also, management must trade off productivity and safety appropriately to prevent "Organizational

  4. Accident at the Fukushima Dai-ichi nuclear power stations of TEPCO. Outline and lessons learned

    International Nuclear Information System (INIS)

    Tanaka, Shun-ichi

    2012-01-01

    The severe accident that broke out at Fukushima Dai-ichi nuclear power stations on March 11, 2011, caused seemingly infinite damage to the daily life of residents. Serious and wide-spread contamination of the environment occurred due to radioactive materials discharged from nuclear power stations (NPSs). At the same time, many issues were highlighted concerning countermeasures to severe nuclear accidents. The accident is outlined, and lessons learned are extracted with respect to the safety of NPSs, as well as radiation protection of residents under the emergency involving the accident. The materials of the current paper are those released by governmental agencies, academic societies, interim reports of committees under the government, and others. (author)

  5. Loss of cooling accident simulation of nuclear power station spent-fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.; Liang, K-S., E-mail: mlee@ess.nthu.edu.tw, E-mail: ksliang_1@hotmail.com [National Tsing Hua Univ., Hsinchu, Taiwan (China); Lin, K-Y., E-mail: syrup760914@gmail.com [Taiwan Power Company, Taiwan (China)

    2014-07-01

    The core melt down accident of Fukushima Nuclear Power Station on March 11th, 2011 alerted nuclear industry that the long term loss of cooling of spent fuel pool may need some attention. The target plant analyzed is the Chinshan Nuclear Power Station of Taiwan Power Company. The 3-Dimensional RELAP5 input deck of the spent fuel pool of the station is built. The results indicate that spent fuel of Chinshan Nuclear Power Station is uncovered at 6.75 days after an accident of loss cooling takes place and cladding temperature rises above 2,200{sup o}F around 8 days. The time is about 13 hours earlier than the results predicted using simple energy balance method. The results also show that the impact of Counter Current Flow Limitation (CCFL) and radiation heat transfer model is marginal. (author)

  6. Preliminary Assessment of the Possible BWR Core/Vessel Damage States for Fukushima Daiichi Station Blackout Scenarios Using RELAP/SCDAPSIM

    Directory of Open Access Journals (Sweden)

    C. M. Allison

    2012-01-01

    Full Text Available Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1–3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.

  7. Safety of Ikata Nuclear Power Station from the accident of Three Mile Island

    International Nuclear Information System (INIS)

    Nonaka, Hiroshi

    1979-01-01

    The leak of radioactive substances occurred on March 28, 1979, in the No. 2 plant of Three Mile Island Nuclear Power Station, and this accident must be put to use to prevent similar accidents and to secure safety hereafter in the nuclear power stations being operated in Japan. In the TMI accident, too many problems concerning the operation management seemed to exist in a series of events. In this paper, a few matters related to the TMI accident among the aspects of the operation management in Ikata Nuclear Power Station are reported. As the problems of operation management, it is considered that the operation of the TMI plant was continued as the exit valve of auxiliary feed line was closed, that it took long time to close the root valve for a pressurizer relief valve manually, and that the ECCS was stopped manually. In TMI, the abnormal phenomenon of losing main feed water has occurred 6 times since the attainment of criticality in March, 1978, and the opening and sticking of pressurizer relief valves occurred at least twice in about 150 times of their actuation in the nuclear reactors designed by Babcock and Wilcox Co. In Ikata Nuclear Power Station, these problems are detected early and the suitable measures are taken immediately, therefore it never happens to continue the operation as the problems are left as they are. It is not conceivable that similar troubles occur many times. (Kako, I.)

  8. Hot Cell Installation and Demonstration of the Severe Accident Test Station

    Energy Technology Data Exchange (ETDEWEB)

    Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burns, Zachary M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    A Severe Accident Test Station (SATS) capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident (DBA) and beyond design basis accident (BDBA) scenarios has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory (IFEL), a hot cell facility at Oak Ridge National Laboratory. The two test station modules provide various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident (LOCA) testing, defueled oxidation quench testing and high temperature BDBA testing. The installation of the SATS system restores the domestic capability to examine postulated and extended LOCA conditions on spent fuel and cladding and provides a platform for evaluation of advanced fuel and accident tolerant fuel (ATF) cladding concepts. This document reports on the successful in-cell demonstration testing of unirradiated Zircaloy-4. It also contains descriptions of the integral test facility capabilities, installation activities, and out-of-cell benchmark testing to calibrate and optimize the system.

  9. Accident at the Fukushima Dai-ichi Nuclear Power Stations of TEPCO —Outline & lessons learned—

    OpenAIRE

    TANAKA, Shun-ichi

    2012-01-01

    The severe accident that broke out at Fukushima Dai-ichi nuclear power stations on March 11, 2011, caused seemingly infinite damage to the daily life of residents. Serious and wide-spread contamination of the environment occurred due to radioactive materials discharged from nuclear power stations (NPSs). At the same time, many issues were highlighted concerning countermeasures to severe nuclear accidents. The accident is outlined, and lessons learned are extracted with respect to the safety o...

  10. Risk Perception and Occupational Accidents: A Study of Gas Station Workers in Southern Brazil

    Directory of Open Access Journals (Sweden)

    Letícia Silveira Cardoso

    2012-07-01

    Full Text Available The present study aimed to identify the perceptions of gas station workers about physical, chemical, biological and physiological risk factors to which they are exposed in their work environment; identify types of occupational accidents involving gas station workers and; report the development of a socioenvironmental intervention as a tool for risk communication to gas station workers. A quantitative study was performed with 221 gas station workers in southern Brazil between October and December 2010. Data collection was performed between October to December 2010 via structured interviews. The data were analyzed using SPSS 19.0. The participants identified the following risk types: chemical (93.7%, physical (88.2%, physiological (64.3% and biological (62.4%. In this sample, 94.1% of gas station workers reported occupational accidents, and 74.2% reported fuel contact with the eyes (p < 0.05. It is concluded that workers perceive risks, and that they tend to relate risks with the occurrence of occupational accidents as an indicator of the dangerous nature of their work environment.

  11. Innovative safety features of VVER for ensuring high degree of autonomy during beyond design basis accidents

    International Nuclear Information System (INIS)

    Kumar, Abhay; Mohan, Joe; Kumar, Devesh; Chaudhry, S.M.; Rao, Srinivasa; Gupta, S.K.

    2010-01-01

    The effectiveness of Passive Heat Removal System (PHRS) in during a station black-out (SBO) accident was assessed by using SCDAP/Relap5. The analysis gave evidence that (i) the Passive Heat Removal System (PHRS) is capable of limiting the consequences of station black out (SBO) and acts as an effective engineered safety system, and (ii) the PHRS intervention prevents core degradation and excessive core heat-up. (P.A.)

  12. Reactor coolant pump shaft seal behavior during blackout conditions

    International Nuclear Information System (INIS)

    Mings, W.J.

    1985-01-01

    The United States Nuclear Regulatory Commission has classified the problem of reactor coolant pump seal failures as an unresolved safety issue. This decision was made in large part due to experimental results obtained from a research program developed to study shaft seal performance during station blackout and reported in this paper. Testing and analysis indicated a potential for pump seal failure under postulated blackout conditions leading to a loss of primary coolant with a concomitant danger of core uncovery. The work to date has not answered all the concerns regarding shaft seal failure but it has helped scope the problem and focus future research needed to completely resolve this issue

  13. Proceedings of the international symposium on environmental monitoring and dose estimation of residents after accident of TEPCO's Fukushima Daiichi Nuclear Power Stations

    International Nuclear Information System (INIS)

    Takahashi, Sentaro; Yamana, Hajimu; Takahashi, Tomoyuki; Takamiya, Koichi; Fukutani, Satoshi; Sato, Nobuhiro; Nakatani, Maki

    2013-02-01

    In March 2011, a massive earthquake and the resulting tsunami struck the Tohoku area in Japan, causing serious damages to TEPCO's Fukushima Daiichi nuclear plant and the release of a significant quantity of radionuclides into the surrounding environment. This accident underlined the necessity of establishing new and comprehensive scientific research for promoting safety in nuclear technology. With this aim, the Kyoto University Research Reactor Institute (KURRI) developed a new research program called the “KUR Research Program for Scientific Basis of Nuclear Safety” from this year. In this program, we are planning to hold an annual series of international symposiums along with many other research activities. The first in this series of symposiums, entitled “The International Symposium on Environmental Monitoring and Dose Estimation of Residents after Accident of TEPCO's Fukushima Daiichi Nuclear Power Station,” deals with the radiological effect of the March 2011 accident in Fukushima Daiichi NPP on the public. The purpose of this symposium is to collate data on environmental radioactivity anadiation dose in residents, discuss and verify these data, and clarify the actual situation of environmental contamination anesultant radiation exposed to the residents. We believe that an accurate estimation of the radiation dose is quite essential for planning for the healthy life and mental contentment of the residents, and we hope that many researchers who are studying the radiological effects of the accident will join us for these purposes. The environmental monitoring data are important for the dose assessment for residents. However, the monitoring data in the early stage are not sufficient for dose assessment, particularly near the NPP site, because of the confusion and blackout caused by the earthquake. However, many researchers and organizations in Japan and other countries have independently carried out radiation monitoring. We believe that the publication and

  14. The Fukushima Dai Ichi accident. The narrative of the station manager. Volume 1. The destruction

    International Nuclear Information System (INIS)

    Guarnieri, Franck; Travadel, Sebastien; Martin, Christophe; Portelli, Aurelien; Afrouss, Aissame; Takesada, Tomoko

    2015-01-01

    While outlining that the Fukushima accident could have been more severe without the courage and action of men who stayed at the controls of the plant under the management of Masao Yoshida, this book proposes a translation of the manager's narrative made for the official inquiry commission. He tells the story of a team of workers facing a disaster foretold. Besides this narrative, the authors propose a discussion on emergency engineering, present the Kan inquiry commission, present the power station and recall the circumstances of the accident and its consequences. Several hearings are reported

  15. Review of Ontario Hydro Pickering 'A' and Bruce 'A' nuclear generating stations' accident analyses

    International Nuclear Information System (INIS)

    Serdula, K.J.

    1988-01-01

    Deterministic safety analysis for the Pickering 'A' and Bruce 'A' nuclear generating stations were reviewed. The methodology used in the evaluation and assessment was based on the concept of 'N' critical parameters defining an N-dimensional safety parameter space. The reviewed accident analyses were evaluated and assessed based on their demonstrated safety coverage for credible values and trajectories of the critical parameters within this N-dimensional safety parameter space. The reported assessment did not consider probability of occurrence of event. The reviewed analyses were extensive for potential occurrence of accidents under normal steady-state operating conditions. These analyses demonstrated an adequate assurance of safety for the analyzed conditions. However, even for these reactor conditions, items have been identified for consideration of review and/or further study, which would provide a greater assurance of safety in the event of an accident. Accident analyses based on a plant in a normal transient operating state or in an off-normal condition but within the allowable operating envelope are not as extensive. Improvements in demonstrations and/or justifications of safety upon potential occurrence of accidents would provide further assurance of adequacy of safety under these conditions. Some events under these conditions have not been analyzed because of their judged low probability; however, accident analyses in this area should be considered. Recommendations are presented relating to these items; it is also recommended that further study is needed of the Pickering 'A' special safety systems

  16. Application of improved topsis method to accident emergency decision-making at nuclear power station

    International Nuclear Information System (INIS)

    Zhang Jin; Cai Qi; Zhang Fan; Chang Ling

    2009-01-01

    Given the complexity in multi-attribute decision-making on nuclear accident emergency, and by integrating subjective weight and impersonal weight of each evaluating index, a decision-making model for emergency plan at nuclear power stations is established with the application of improved TOPSIS model. The testing results indicated that the improved TOPSIS-based multi-attribute decision-making has a better assessment results. (authors)

  17. Complementary safety evaluation of the Phenix power station (INB n 71) in the light of the Fukushima power station accident

    International Nuclear Information System (INIS)

    2011-01-01

    This report proposes a complementary safety evaluation of the Phenix power station, one of the French basic nuclear installations (BNI, in French INB) in the light of the Fukushima accident. This evaluation takes the following risks into account: risks of flooding, earthquake, loss of power supply and loss of cooling, in addition to operational management of accident situations. It presents some characteristics of the Phenix installation (location, operator, industrial environment, installation characteristics), identifies the risks of cliff effect and the main structures and equipment, evaluates the seismic risk (installation sizing, installation conformity, margin evaluation), evaluates the flooding risk (installation sizing, installation conformity, margin evaluation), briefly examines other extreme natural phenomena (extreme meteorological conditions related to flooding, earthquake or flooding with a higher level than that for which the installation is designed). It analyzes the risk of a loss of power supply and of cooling (loss of external and internal electric sources, loss of the ultimate cooling system). It analyzes the management of severe accidents: crisis management organization, available intervention means, robustness of available means. It discusses the conditions of the use of subcontractors

  18. Review of Cytogenetic analysis of restoration workers for Fukushima Daiichi nuclear power station accident

    International Nuclear Information System (INIS)

    Suto, Yumiko

    2016-01-01

    Japan faced with the nuclear accident of the Fukushima Daiichi Nuclear Power Station (NPS) caused by the combined disaster of the Great East Japan Earthquake and the subsequent tsunamis on 11 March 2011. National Institute of Radiological Sciences received all nuclear workers who were engaged in emergency response tasks at the NPS and suspected of being overexposed to acute radiation. Biological dosimetry by dicentric chromosome assay was helpful for medical triage and management of the workers. When an unplanned radiation exposure occurs, biological dosimetry based on cytogenetic assays has been used to estimate the absorbed dose in the exposed individual to get useful information for the medical management of radiological casualties with suspected acute radiation syndrome (ARS). Nowadays, more cytogenetic assays to measure chromosomal aberrations, such as micronuclei in bi-nucleated cells, prematurely condensed chromosomes (PCCs) and inter-chromosomal exchanges detected by fluorescence in situ hybridization (FISH) techniques, are available. However, the dicentric chromosome assay (DCA) using peripheral blood lymphocytes is still considered to be the 'gold standard' of biological dosimetry for the radiation emergency medicine. Experimental protocols of DCA has been standardized and shared among laboratories all over the world. In fact, DCA was useful in previous radiation accidents, e.g. the Chernobyl accident in 1986, the Goiania accident in 1987, the JCO criticality accident in 1999 and the Tokyo electric power company (TEPCO) Fukushima Daiichi Nuclear Power Station (NPS) accident in 2011. The recent development of microscopic image analysis system with automatic metaphase finding and capturing functions was helpful for rapid detection of dicentric chromosomes to perform DCA for the Fukushima NPS restoration workers. (author)

  19. Determinants of injuries and Road Traffic Accidents amongst service personnel in a large Defence station.

    Science.gov (United States)

    Neelakantan, Anand; Kotwal, Brig Atul; Ilankumaran, Mookkiah

    2017-07-01

    Injuries are assuming epidemic proportions globally; and in India. Also, previous decade witnessed carnage on Indian roads, with nearly 12 lakh people killed and 55 lakhs disabled in road crashes. The trend in Armed Forces is reflective of the aforesaid patterns. Behaviour and socio-demographic background of the victims are significant determinants of injuries and road accidents. Community-based epidemiological information on these aspects is envisaged to contribute in their preventive strategy. Towards this direction, the present study was conducted with aim to generate socio-behavioural profile of injuries and Road Traffic Accidents (RTAs) amongst service personnel in a large defence station; and to evaluate their determinants. A cross sectional descriptive study was carried out among 796 Naval personnel onboard warships in large Naval station. Data on socio-behavioural aspects and determinants of injuries and road accidents was collected using a pre-validated questionnaire; and by scrutiny of relevant records. Data was analysed using MSExcel, Epi-info and SPSS 17. Young and middle-aged persons were predominantly involved in injuries and road accidents. Two-wheeler users sustained maximum road accidents. Human factor was a significant determinant in RTAs and injuries. A majority of victims admitted that human factors were the predominant cause of road accidents; and opined that the events were preventable. Age-specific Behavioural Change Communication strategies aimed at refining user outlook are imperative; tailored to sociodemographic milieu of user/victim. Incorporation of a dynamic feedback/reporting mechanism, creation of 'armed forces-specific road safety and injury prevention policy' and safety audits on injuries and road crashes are measures in this direction.

  20. Interim report on the accident at Fukushima Nuclear Power Stations of Tokyo Electric Power Company

    International Nuclear Information System (INIS)

    2011-12-01

    The Investigation Committee on the Accident at the Fukushima Nuclear Power Stations (the Investigation Committee) of Tokyo Electric Power Company (TEPCO) was established by the Cabinet decision on May 24, 2011. Its objectives are: to conduct investigation for finding out the causes of accidents at the Fukushima Dai-ichi Nuclear Power Station (Fukushima Dai-ichi NPS) and Fukushima Dai-ni Nuclear Power Station (Fukushima Dai-ni NPS) of TEPCO as well as the causes of accident damage; and to make policy recommendations for limiting the expansion of damage and preventing reoccurrence of similar accidents. The Investigation Committee has conducted its investigation and evaluation since its first meeting on June 7, 2011. Its activities included: site visits to the Fukushima Dai-ichi and Dai-ni NPSs, as well as to other facilities; hearing of heads of local governments around the Fukushima Dai-ichi NPS; and hearing of people concerned through interviews mainly arranged by the Secretariat. As of December 16, 2011, the number of interviewees reached 456. The investigation and evaluation by the Investigation Committee are still ongoing and the Interim Report does not cover every item that the Committee aims at investigating and evaluating. Fact-finding of even some of those items discussed in the Interim Report are not yet completed. The Investigation Committee continues to conduct its investigation and evaluation and will issue its Final Report in the summer of 2012. This brief executive summary covers mainly considerations and evaluation of the issues in Chapter VII of the Interim Report, with brief reference to Chapters I to VI. The Investigation Committee recommendations are printed in bold. (author)

  1. Computer chaos and the blackout

    CERN Multimedia

    Malik, Rex

    1971-01-01

    A recent electricity dispute resulted in power black-outs with unfortunate consequences for organizations relying on computers. Article discusses the implications of similar events in Britain in the future when computers are even more widely in use (1 1/2 pages).

  2. Analysis of SBO accident and natural circulation of 49-2 swimming pool reactor

    International Nuclear Information System (INIS)

    Wu Yuanyuan; Liu Tiancai; Sun Wei

    2012-01-01

    The transient thermal hydraulic characteristics of 49-2 Swimming Pool Reactor (SPR) were analyzed by RELAP5/MOD3.3 code to verify the capability of natural circulation and minus reactivity feedback for accident mitigation under the condition of station blackout (SBO). Then, the effects on accident consequence and sequence for core channels and primary pumps were briefly discussed. The calculation results show that the reactor can be shutdown by the effect of minus reactivity feedback, and the residual heat can be removed through the stable natural circulation. Therefore, it demonstrates that the 49-2 SPR is safe during the accident of SBO. (authors)

  3. 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

  4. Blackout cloth for dormancy induction

    Science.gov (United States)

    Tom Jopson

    2007-01-01

    The use of blackout cloth to create long night photoperiods for the induction of dormancy in certain conifer species has been an established practice for a long time. Its use was suggested by Tinus and McDonald (1979) as an effective technique, and the practice has been commonly used in Canadian forest nurseries for a number of years. Cal-Forest Nursery installed its...

  5. Accident at the Fukushima Dai-ichi Nuclear Power Stations of TEPCO —Outline & lessons learned—

    Science.gov (United States)

    TANAKA, Shun-ichi

    2012-01-01

    The severe accident that broke out at Fukushima Dai-ichi nuclear power stations on March 11, 2011, caused seemingly infinite damage to the daily life of residents. Serious and wide-spread contamination of the environment occurred due to radioactive materials discharged from nuclear power stations (NPSs). At the same time, many issues were highlighted concerning countermeasures to severe nuclear accidents. The accident is outlined, and lessons learned are extracted with respect to the safety of NPSs, as well as radiation protection of residents under the emergency involving the accident. The materials of the current paper are those released by governmental agencies, academic societies, interim reports of committees under the government, and others. PMID:23138450

  6. Investigation on accident management measures for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Tusheva, P.; Schaefer, F.; Rohde, U.; Reinke, N.

    2009-01-01

    A consequence of a total loss of AC power supply (station blackout) leading to unavailability of major active safety systems which could not perform their safety functions is that the safety criteria ensuring a secure operation of the nuclear power plant would be violated and a consequent core heat-up with possible core degradation would occur. Currently, a study which examines the thermal-hydraulic behaviour of the plant during the early phase of the scenario is being performed. This paper focuses on the possibilities for delay or mitigation of the accident sequence to progress into a severe one by applying Accident Management Measures (AMM). The strategy 'Primary circuit depressurization' as a basic strategy, which is realized in the management of severe accidents is being investigated. By reducing the load over the vessel under severe accident conditions, prerequisites for maintaining the integrity of the primary circuit are being created. The time-margins for operators' intervention as key issues are being also assessed. The task is accomplished by applying the GRS thermal-hydraulic system code ATHLET. In addition, a comparative analysis of the accident progression for a station blackout event for both a reference German PWR and a reference VVER-1000, taking into account the plant specifics, is being performed. (authors)

  7. Swiss Solutions for Providing Electrical Power in Cases of Long-Term Black-Out of the Grid

    International Nuclear Information System (INIS)

    Altkind, Franz; Schmid, Daniel

    2015-01-01

    A better understanding of nuclear power plant electrical system robustness and defence-in-depth may be derived from comparing design and operating practices in member countries. In pursuing this goal, the current paper will focus on Switzerland. It will present in general the protective measures implemented in the Swiss nuclear power plants to ensure power supply, which comply with the 'Defence-in-depth' principle by means of several layers of protection. In particular it will present the measures taken in case of a total station blackout. The different layers supplying electricity may be summed up as follows. The first layer consists of the external main grid, which the plant generators feed into. The second layer is the auxiliary power supply when the power plant is in island mode in case of a failure of the main grid. A third layer is provided by the external reserve grid in case of both a failure of the external main grid and of the auxiliary power supply in island mode. As a fourth layer there exists an emergency electrical power supply. This is supplied either from an emergency diesel generator or a direct feed from a hydroelectric power plant. In the fifth layer, the special emergency electrical power supply from bunkered emergency diesel generators power the special emergency safety system and is activated upon the loss of all external feeds. A sixth layer consists of accident management equipment. Since the Fukushima event, the sixth layer has been reinforced and a seventh layer with off-site accident management equipment has been newly added. The Swiss nuclear safety regulator has analysed the accident. It reviewed the Swiss plants' protection against earthquakes as well as flooding and demanded increased precautionary measures from the Swiss operators in the hypothetical case of a total station blackout, when all the first five layers of supply would fail. In the immediate, a centralized storage with severe accident management equipment

  8. Earthworm populations in soils contaminated by the Chernobyl atomic power station accident, 1986-1988

    International Nuclear Information System (INIS)

    Krivolutzkii, D.A.; Pokarzhevskii, A.D.; Viktorov, A.G.

    1992-01-01

    A study of earthworm populations in the 30 km zone around the Chernobyl atomic power station was carried out in 1986-1988. Significant differences in earthworm population numbers were found between highly contaminated and control plots in summer and autumn 1986 and in April 1987. But in the autumn of 1988 the earthworm population numbers in contaminated plots were higher than in the control plots. The ratio of mature to immature specimens was higher in 1986 in the contaminated plots in comparison with the control plots. Only one species of earthworms, Dendrobaena octahedra, was found in contaminated forest plots during the first 2 yr following the accident but in the control forest plots Apporectodea caliginosa was also found. (author)

  9. Comparison of Severe Accident Results Among SCDAP/RELAP5, MAAP, and MELCOR Codes

    International Nuclear Information System (INIS)

    Wang, T.-C.; Wang, S.-J.; Teng, J.-T.

    2005-01-01

    This paper demonstrates a large-break loss-of-coolant accident (LOCA) sequence of the Kuosheng nuclear power plant (NPP) and station blackout sequence of the Maanshan NPP with the SCDAP/RELAP5 (SR5), Modular Accident Analysis Program (MAAP), and MELCOR codes. The large-break sequence initiated with double-ended rupture of a recirculation loop. The main steam isolation valves (MSIVs) closed, the feedwater pump tripped, the reactor scrammed, and the assumed high-pressure and low-pressure spray systems of the emergency core cooling system (ECCS) were not functional. Therefore, all coolant systems to quench the core were lost. MAAP predicts a longer vessel failure time, and MELCOR predicts a shorter vessel failure time for the large-break LOCA sequence. The station blackout sequence initiated with a loss of all alternating-current (ac) power. The MSIVs closed, the feedwater pump tripped, and the reactor scrammed. The motor-driven auxiliary feedwater system and the high-pressure and low-pressure injection systems of the ECCS were lost because of the loss of all ac power. It was also assumed that the turbine-driven auxiliary feedwater pump was not functional. Therefore, the coolant system to quench the core was also lost. MAAP predicts a longer time of steam generator dryout, time interval between top of active fuel and bottom of active fuel, and vessel failure time than those of the SR5 and MELCOR predictions for the station blackout sequence. The three codes give similar results for important phenomena during the accidents, including SG dryout, core uncovery, cladding oxidation, cladding failure, molten pool formulation, debris relocation to the lower plenum, and vessel head failure. This paper successfully demonstrates the large-break LOCA sequence of the Kuosheng NPP and the station blackout sequence of the Maanshan NPP

  10. Radiation dose estimation from foods due to the accident of TEPCO Fukushima Daiichi Nuclear Power Station

    International Nuclear Information System (INIS)

    Yamaguchi, Ichiro

    2012-01-01

    Explained are the purpose of dose assessment, its methods, actual radionuclide levels in food, amounts of food intake, dose estimated hitherto, dose in the future, dose estimated by total food studies, and problems of assessing the dose from food, all of which Tokyo Electric Power Company (TEPCO) Power Station Accident has raised. Dose derived from food can be estimated by the radioactivity measured in each food material and in its combined amounts or in actually cooked food. Amounts of radioactive materials ingested in the body can be measured externally or by bioassay. Japan MHLW published levels of radioactivity in vegetables', fruits, marine products and meats from Mar. 2011, of which time course pattern has been found different each other within and between month(s). Dose due to early exposure in the Accident can be estimated by the radioactivity levels above and data concerning the amounts of food intake summarized by National Institute of Health and Nutrition in 2010 and other institutions. For instance, the thyroid tissue equivalent dose by I-131 in a 1 year old child is estimated to be 1.1-5 mSv depending on the assumed data for calculation, in the first month after the Accident when ICRP tissue equivalent dose coefficient 3.7 x 10-6 Sv/Bq is used. In the future (later than Apr. 2012), new standard limits of radiocesium levels in milk/its products and foods for infant and in other general foods are to be defined 50 and 100 Bq/kg, respectively. The distribution of committed effective doses by radiocesium (mSv/y food intake) are presented as an instance, where it is estimated by 1 million stochastic simulations using 2 covariates of Cs-134, -137 levels (as representative nuclides under regulation) in food and of daily food intake. In dose prediction, conjecturing the behavior of environmental radionuclides and the time of resume of primary industries would be necessary. (T.T.)

  11. 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

  12. Chernobyl reactor accident. A documentation submitted by the Deutsche Welle radio station. Der Fall Tschernobyl. Eine Dokumentation der Deutschen Welle

    Energy Technology Data Exchange (ETDEWEB)

    1986-01-01

    The documentation abstracted contains a complete survey of the broadcasts transmitted by the Russian wire service of the Deutsche Welle radio station between April 28 and May 15, 1986 on the occasion of the Chernobyl reactor accident. Access is given to extracts of the remarkable eastern and western echoes on the broadcasts of the Deutsche Welle.

  13. The emergency medical programs of japan and foreign countries for radiation accidents in nuclear power stations

    International Nuclear Information System (INIS)

    Aoki, Yoshiro

    1994-01-01

    In our country, the medical emergency programs for the people living near nuclear power stations are well organized, however, preparation of medical staffs who are well trained is considered to be not sufficient. In the USA, on call 24 hours response to a radiological emergency is provided and funded by Department of Energy(DOE) or electric companies. Especially, REAC/TS is a part of DOE response network, in which there are provided well-trained physicians, nurses, health physicists, coordinators and support personnels. In United Kingdom, National Radiological Protection Board(NRPB) is responsible to a radiological emergency program. Each nuclear power station has its own emergency program consisting of a team of physicians, nurses and health physicists. In France, French Atomic Energy Commission (CEA) is a responsible agency for a radiological emergency program. On call 24 hours response to a radiological emergency is provided in Fontenay-aux Roses Institute and Curie Institute. Curie Institute also responds to radiological emergencies in other countries at the request of WHO. In Germany(West Germany), compulsory assurance system covers a radiological emergency program and a radiological protection. There are seven centers in West Germany, in which well-trained medical staffs are provided against radiological injuries. In this report, I tried to propose a new concept about emergency medical programs for nuclear power station accidents in Japan. I think it is a very urgent theme to provide on call 24 hours radiological emergency program, in which patients suffered from acute radiation sickness with internal contamination or contaminated radiation burns will be treated without any trouble. We have to make our best efforts to complete basic or clinical research about radiation injuries including bone marrow transplantation, radioprotectors, chelating agents and radiation burns etc. (J.P.N.)

  14. Experience and lessons learned from emergency disposal of Fukushima nuclear power station accident

    International Nuclear Information System (INIS)

    Xu Xiegu; Zhen Bei; Yang Xiaoming; Chen Xiaohua

    2012-01-01

    After Fukushima nuclear accident, we visited the related medical aid agencies for nuclear accidents and conducted investigations in disaster-affected areas in Japan. This article summarizes the problems with emergency disposal of Fukushima nuclear accident while disclosing problems should be solved during the emergency force construction for nuclear accidents. (authors)

  15. RBMK-1500 accident management for loss of long-term core cooling

    International Nuclear Information System (INIS)

    Uspuras, E.; Kaliatka, A.

    2001-01-01

    Results of the Level 1 probabilistic safety assessment of the Ignalina NPP has shown that in topography of the risk, transients dominate above the accidents with LOCAs and failure of the core long-term cooling are the main factors to frequency of the core damage. Previous analyses have shown, that after initial event, as a rule, the reactivity control, as well as short-term and intermediate cooling are provided. However, the acceptance criteria of the long-term cooling are not always carried out. It means that from this point of view the most dangerous accident scenarios are the scenarios related to loss of the core long-term cooling. On the other hand, the transition to the core condition due to loss of the long-term cooling specifies potential opportunities for the management of the accident consequences. Hence, accident management for the mitigation of the accident consequences should be considered and developed. The most likely initiating event, which probably leads to the loss of long term cooling accident, is station blackout. The station blackout is the loss of normal electrical power supply for local needs with an additional failure on start-up of all diesel generators. In the case of loss of electrical power supply MCPs, the circulating pumps of the service water system and MFWPs are switched-off. At the same time, TCV of both turbines are closed. Failure of diesel generators leads to the non-operability of the ECCS long-term cooling subsystem. It means the impossibility to feed MCC by water. The analysis of the station blackout for Ignalina NPP was performed using RELAP5 code. (author)

  16. Accident sequences simulated at the Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1998-01-01

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

  17. Accident of the Fukushima-Daiichi nuclear power station. Situation two years after the event - IRSN file

    International Nuclear Information System (INIS)

    2013-03-01

    Two years after the Fukushima accident, this report proposes a review of the situation in Japan, and of the European and international actions aimed at preventing the occurrence of another nuclear accident and its radiological consequences. It is based on information available at the end of January or February 2013. After a recall of the situation in Japan and Europe in 2011 (recall of the accident, of the different simulation, calculation and information actions undertaken by the IRSN, launching of a program of additional safety assessments and of European stress tests), the report addresses the situation in Japan two years after the accident: evolution of the nuclear risk management governance, status of the Fukushima-Daiichi power station, health and environmental impact and management of the post-accidental phase, actions undertaken by the IRSN (dose assessment, cooperation in the field of severe accidents, participation to the Fukushima Dialogue). The next part details the contribution of the IRSN to the strengthening of safety and radiation protection at the international level (in relationship with international organizations: IAEA, UNSCEAR and WHO). Additional technical information is provided in appendix, as well as a report on the environmental impact of the accident, and a report on the post-accidental management of the accident

  18. How should we handle Fukushima Daiichi Nuclear Power Station accident with engineer ethics? What is seen by encountering resilience engineering

    International Nuclear Information System (INIS)

    Oba, Kyoko

    2017-01-01

    Many of lectures on 'engineer ethics' being held at universities, etc. positively incorporate case study. This paper introduced the Fukushima Daiichi Nuclear Power Station accident (1F accident) not as a mere failure case, but with broader view and broader manner based on practice. Resilience engineering cites anticipating, monitoring, responding, and learning as four core capabilities required to realize safety that people and organizations should aim at. The authors tried to analyze 1F accident using these four core capabilities and the elements required to demonstrating these core capabilities. In the responding of TEPCO to earthquake and tsunami cases before the accident, tsunami countermeasure responding required to prevent 1F accident was not demonstrated as a result. Good examples seen in other responding are the construction of the seismic isolation important building and the deployment of fire engines to the whole nuclear power plant. Accident reports so far took viewpoints of why the accident occurred, and why it led to hydrogen explosion. However, from resilience engineering, why catastrophic conditions could be avoided and how water injection in nuclear facilities could be assured become viewpoints. (A.O.)

  19. Identification of NPP accidents using support vector classification

    Energy Technology Data Exchange (ETDEWEB)

    Back, Ju Hyun; Yoo, Kwae Hwan; Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2016-10-15

    In case of the accidents that happens in a nuclear power plants (NPPs), it is very important to identify its accidents for the operator. Therefore, in order to effectively manage the accidents, the initial short time trends of major parameters have to be observed and NPP accidents have to accurately be identified to provide its information to operators and technicians. In this regard, the objective of this study is to identify the accidents when the accidents happen in NPPs. In this study, we applied the support vector classification (SVC) model to classify the initiating events of critical accidents such as loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), station blackout (SBO), and steam generator tube rupture (SGTR). Input variables were used as the initial integral value of the signal measured in the reactor coolant system (RCS), steam generator, and containment vessel after reactor trip. The proposed SVC model is verified by using the simulation data of the modular accident analysis program (MAAP4) code. In this study, the proposed SVC model is verified by using the simulation data of the modular accident analysis program (MAAP4) code. We used an initial integral value of the simulated sensor signals to identify the NPP accidents. The training data was used to train the SVC model. And, the trained model was confirmed using the test data. As a result, it was known that it can accurately classify five events.

  20. Whole-body counting of Fukushima residents after the TEPCO Fukushima Daiichi Nuclear Power Station accident

    Energy Technology Data Exchange (ETDEWEB)

    Momose, Takumaro; Takada, Chie; Nakagawa, Takahiro; Nakai, Katsuta; Kurihara, Osamu; Tsujimura, Norio; Furuta, Sadaaki [Japan Atomic Energy Agency, Nuclear Fuel Cycle Engineering Laboratories, Tokai, Ibaraki (Japan); Ohi, Yoshihiro; Murayama, Takashi; Suzuki, Takashi [Japan Atomic Energy Agency, Nuclear Science Research Institute, Tokai, Ibaraki (Japan); Uezu, Yasuhiro [Japan Atomic Energy Agency, Fukushima Environmental Safety Center, Fukushima, Fukushima (Japan)

    2012-11-15

    At the request of the Fukushima government, the Japan Atomic Energy Agency (JAEA) started whole-body counting of residents on July 11, 2011, to assess radiation exposure after the TEPCO Fukushima Daiichi Nuclear Power Station accident. JAEA has examined residents in Iitate, Kawamata, Namie, and eight other local communities. The measurement capacity of the whole-body counting device is approximately 100 persons per day, and the total number of people to measure reached 9,927 by the end of January 2012. Two types of whole-body counter (WBC), each of which has two large-sized NaI(Tl) detectors, were used to perform the measurements. Physical phantoms (a Canberra RMC-II transfer phantom or a water-filled block phantom developed by JAEA) were used to perform the efficiency calibration of the WBCs, and results of this calibration were verified by comparing them with different-sized BOttle Mannequin ABsorber (BOMAB) phantoms imitating an adult male, a 10-year-old child, and a 4-year-old child. Short half-life radionuclides (e.g., {sup 131}I) originating from the accident were not detected in the present work, because the measurements were made starting four months after the accident. These measurements showed that about 80% of the residents were below a minimum detectable amount (MDA) of the measured radionuclides in the whole-body content. No artificial nuclides other than {sup 134}Cs and {sup 137}Cs were detected in the present whole-body counting. The maximum whole-body content of radiocesium ({sup 134}Cs and {sup 137}Cs in total) was 2.7 kBq for children (< 8 years old) and 14 kBq for adults. The radioactivity ratio ({sup 137}Cs/{sup 134}Cs) was estimated to be from 1.12 to 1.26. An acute intake scenario via inhalation of radiocesium was assumed in the present dose estimation for all the residents who were measured by the end of January 2012. The committed effective dose (CED) to 99.8% of the residents was found to be below 1 mSv. There were only 25 subjects with a

  1. Radiation risk after Fukushima Nuclear Power Station accident and recognition of society

    International Nuclear Information System (INIS)

    Yamashita, Shunichi

    2017-01-01

    In the Fukushima Nuclear Power Station accident, the confusion caused by inconsistency of risk assessment among scientists cast new challenges for communication between science and society. In response to the way of crisis communication in the future, the Japan Society for the Promotion of Science (JSPS) is required to extract specific subjects and to make efforts to solve them. The Committee for 'Radiation Risk and Crisis Communication' had been set up in the Leading R and D Committee of JSPS over three years since October 2013. This paper introduced the outline of the Committee for 'Radiation Risk and Crisis Communication,' with a focus on the activity system of three subcommittees, activity guidelines and contents of each subcommittee, and the outcomes of activities of each subcommittee. The themes of the subcommittees are as follows. The subcommittee 1 is to collect, analyze, and organize information on the effects of radiation based on the latest findings, the subcommittee 2 is to study the formation of consensus within the scientific community as well as information disclosure methods, and the subcommittee 3 is to survey and study the information disclosure means of radiation measurement results under crisis situation. (A.O.)

  2. Effect of air condition on AP-1000 containment cooling performance in station black out accident

    International Nuclear Information System (INIS)

    Hendro Tjahjono

    2015-01-01

    AP1000 reactor is a nuclear power plant generation III+ 1000 MWe which apply passive cooling concept to anticipate accidents triggered by the extinction of the entire supply of electrical power or Station Black Out (SBO). In the AP1000 reactor, decay heat disposal mechanism conducted passively through the PRHR-IRWST and subsequently forwarded to the reactor containment. Containment externally cooled through natural convection in the air gap and through evaporation cooling water poured on the outer surface of the containment wall. The mechanism of evaporation of water into the air outside is strongly influenced by the conditions of humidity and air temperature. The purpose of this study was to determine the extent of the influence of the air condition on cooling capabilities of the AP1000 containment. The method used is to perform simulations using Matlab-based analytical calculation model capable of estimating the power of heat transferred. The simulation results showed a decrease in power up to 5% for relative humidity rose from 10% to 95%, while the variation of air temperature of 10°C to 40°C, the power will decrease up to 15%. It can be concluded that the effect of air temperature increase is much more significant in lowering the containment cooling ability compared with the increase of humidity. (author)

  3. Screening survey on thyroid exposure for children after the Fukushima Daiichi Nuclear Power Station accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eunjoo; Kurihara, Osamu; Suzuki, Toshikazu; Matsumoto, Masaki; Fukutsu, Kumiko; Yamada, Yuji; Sugiura, Nobuyuki; Akashi, Makoto [National Institute of Radiological Sciences, Research Center for Radiation Emergency Medicine, Chiba, Chiba (Japan)

    2012-11-15

    In response to a serious accident at the Fukushima Daiichi Nuclear Power Station, National Institute of Radiological Sciences (NIRS) established a protocol for a screening survey of thyroid exposure for children living in areas where thyroid doses were predicted to be high. The aim of the screening survey was to implement measurements for a large number of subjects with conventional NaI(Tl) scintillation survey meters. This protocol was applied to the screening survey of 1,149 children at five locations in three municipalities (Kawamata Town, Iitate Village and Iwaki City). Among 1,080 children (excluding 69 subjects from evaluation), there were no subjects who exceeded a screening level (0.2 {mu}Sv h{sup -1}) corresponding to a thyroid equivalent dose of 100 mSv (for the age group of 1-y-old as of March 24). No significant signals were detected in 55.4% of these subjects and the maximum dose was found to be 43 mSv. This paper presents details on the protocol as well as results of the screening survey. (author)

  4. Communication activity for residents to understand radiation after the accident of Fukushima Daiichi Nuclear Power Station

    International Nuclear Information System (INIS)

    Itabashi, Kiyoshi; Tagawa, Akihiro; Sugiyama, Kenji; Yamamoto, Tomoyo

    2015-01-01

    'Question-and-Answer Session on Radiation and Health' ('Kotaeru-kai' in Japanese) has started in July 2011 in Fukushima Prefecture, which was influenced by the accident of Fukushima Daiichi Nuclear Power Station following the Great East Japan Earthquake and Tsunami on 11 March 2011. The purpose of the Session is to have mainly parents and teachers (kindergartens, schools etc.) understand correctly about radiation and its influence on health. At the requests of the teachers in Fukushima Prefecture, about 4 staff members made a team, and visited Fukushima. The members of the team were selected from 500 JAEA staffs nominated beforehand. The members explained about radiation and its influence on health by using illustrations and metaphors. After the lecture, they answered the questions asked in advance at schools. Also they answered the questions asked in the Session. In the Session, the members placed much value on the communication with participants. Until the end of December 2014, the Question-and-Answer Sessions on Radiation and Health have been held 241 times for about twenty thousand participants. According to 7,613 participants' questionnaires, which were collected from July 2011 to the end of 2012, it seems that participants were able to understand well about radiation and its influence on health. Besides parents and teachers, some of the junior high schools requested to explain for students. JAEA will continue this communication activity in order to meet these expectations and requirements. (author)

  5. An outline of the interim report of the investigation committee on the accident at Fukushima Nuclear Power Stations

    International Nuclear Information System (INIS)

    Yoshioka, Hitoshi

    2012-01-01

    Interim report of the Investigation Committee of the Accident at Fukushima Nuclear Power Stations (NPSs) was published in December 26, 2011. The Japanese cabinet approved ten committee members including the author in May 2011. The committee interviewed more than 400 people over a total of 900 hours of hearings with about 40 staffs consisting of administrative team and three investigation teams of social system, root causes of the accident and countermeasures to prevent damage expansion of the accident. Interim report concluded 'the accident at Fukushima NPSs was caused by failures of every provision against reactor severe accident'. The failures appeared on (1) function of supervisory system for emergency response, (2) Fukushima Daiichi NPSs on-site disaster response especially related with operation of isolation condenser of unit 1 and high-pressure coolant injection system of unit 3, (3) Fukushima Daiichi NPSs off-site disaster response such the government failed to make use of data on the radioactive plumes released from the plant for evacuations, and (4) preparedness against tsunami and severe accident management. Possible worst or best simulation cases were also discussed. With no human support available on-site, workers might not have been able to prevent the meltdowns. Final report was due at the end of July 2012. (T. Tanaka)

  6. Severe accident sequence assessment for boiling water reactors: program overview

    International Nuclear Information System (INIS)

    Fontana, M.H.

    1980-10-01

    The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case

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

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

    2015-01-01

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  8. Report from investigation committee on the accident at the Fukushima Nuclear Power Stations of Tokyo Electric Power Company

    International Nuclear Information System (INIS)

    Koshizuka, Seiichi

    2012-01-01

    Government's Investigation Committee on the Accident at Fukushima Nuclear Power Stations of Tokyo Electric Power Company published its final report on July 23, 2012. Results of investigation combined final report and interim report published on December 26, 2011. The author was head of accident accuse investigation team mostly in charge of site response, prior measure and plant behavior. This article reported author related technical investigation results focusing on site response and prior measures against tsunamis of units 1-3 of Fukushima Nuclear Power Stations. Misunderstanding of working state of isolation condenser of unit 1, unsuitability of alternative water injection at manual stop of high-pressure coolant injection (HPCI) system of unit 3 and improper prior measure against tsunami and severe accident were pointed out in interim report. Improper monitoring of suppression chamber of unit 2 and again unsuitable work for HPCI system of unit 3 were reported in final report. Thorough technical investigation was more encouraged to update safety measures of nuclear power stations. (T. Tanaka)

  9. The Chernobyl accident: Causes and consequences

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1987-01-01

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

  10. Suppositional analysis for the nuclear accidents at the Fukushima Daiichi Nuclear Power Station

    International Nuclear Information System (INIS)

    Yoshida, Yoshitaka

    2012-01-01

    The Great East Japan Earthquake which occurred on March 11, 2011 set off a series of events that have led to a serious nuclear disaster. The Fukushima Daiichi Nuclear Power Station suffered damage due to the huge tsunami waves triggered by the earthquake, which lead to release of large amounts of radioactive materials into the environment. Using the limited information presented by the government and others within several days after the accident, this study examined the amount of radioactive materials released into the atmosphere predicted using the incident progress prediction system (IPPS) and the radioactive release, radiation dose and radiological protection area prediction system (R-Cubic) each developed by INSS, and compared those findings with the amount of radioactive materials evaluated by the government and the Tokyo Electric Power Company. The following points were seen (1) The step-by-step public protection measures taken by the government were consistent with the amount of released radioactive materials by R-Cubic. (2) The IPPS underestimated the amount of released radioactive materials when the irradiation levels in the nuclear reactors were analyzed according to the observed nuclear reactor water levels, although these were well in agreement with the observed parameters. (3) The amounts of released radioactive materials predicted by R-Cubic agreed well with those of the Nuclear Safety Commission and the Nuclear and Industrial Safety Agency. (4) The 0.5MeV equivalent value of noble gases in the amount of radioactive materials released into the atmosphere shown by TEPCO on May 2012 was considered to be an underestimate. (author)

  11. Development and application of a radioactivity evaluation technique the to obtain radiation exposure dose of radioactivity evaluation technique when a severe accident occurs in the a power station of a severe accident. Accident management guidelines of knowledge-based maintenance

    International Nuclear Information System (INIS)

    Kawasaki, Ikuo; Yoshida, Yoshitaka

    2013-01-01

    As a One of the lessons learned from the nuclear accident at the Fukushima Daiichi Nuclear Power Stations of Tokyo Electric Power Company, the was the need for improvement of accident management guidelines is required. In this report study, we developed and applied a dose evaluation technique to evaluated the radiation dose in a nuclear power plant assuming three conditions: employees were evacuation evacuated at the time of a severe accident occurrence; operators carried out the accident management operation; of the operators, and the repair work was carried out for of the trouble damaged apparatuses in a the nuclear power plant using a dose evaluation system. The following knowledge findings were obtained and should to be reflected to in the knowledge base of the guidelines was obtained. (1) By making clearly identifying an areas beforehand becoming the that would receive high radiation doses at the time of a severe accident definitely beforehand, we can employees can be moved to the evacuation places through an areas having of low dose rate and it is also known it how much we long employees can safely stay in the evacuation places. (2) When they circulate CV containment vessel recirculation sump water is recirculated by for the accident management operation and the restoration of safety in the facilities, because the plumbing piping and the apparatuses become radioactive radioactivity sources, the dose evaluation of the shortest access route and detour access routes with should be made for effective the accident management operation is effective. Because the area where a dose rate rises changes which as safety apparatuses are restored, in consideration of a plant state, it is necessary to judge the rightness or wrongness of the work continuation from the spot radioactive dose of the actual apparatus area, with based on precedence of the need to restore with precedence, and to choose a system to be used for accident management. (author)

  12. Management, administrative and operational causes of the accident: Chernobyl nuclear power station

    International Nuclear Information System (INIS)

    Anastas, G.

    1996-01-01

    Full text: The Chernobyl accident, which occurred in April 1986, was the result of management, administrative, operational, technical and design flaws. The accident released millions of curies of mixed fission products (including 70-100 P Bq of 137 Cs). The results of this study strongly suggest that the cultural, political, managerial and operational attributes of the Soviet 'system' performed in a synergistic manner to significantly contribute to the initiation of the accident. At the time of the accident, science, engineering and safety in the former Soviet Union were dominated by an atmosphere of politics, group think and 'dingoes tending the sheep'

  13. Management, administrative and operational causes of the accident: Chernobyl nuclear power station

    International Nuclear Information System (INIS)

    Anastas, G.

    1996-01-01

    The Chernobyl accident, which occurred in April 1986, was the result of management, administrative, operational, technical and design flaws. The accident released millions of curies of mixed fission products including 70-100 PBq of 137 Cs. At the time of the accident, science, engineering and safety in the former Soviet Union were dominated by an atmosphere of politics, group think and 'dingoes tending the sheep'. This corrupted safety culture exacerbated the poor design of the reactor. The results of this study strongly suggest that the cultural, political, managerial and operational attributes of the Soviet 'system' performed in a synergistic manner to significantly contribute to the initiation of the accident. (authors)

  14. Study on self organized criticality of China power grid blackouts

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Xingyong; Zhang, Xiubin; He, Bin [Department of Electrical Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Minhang District, Shanghai 200240 (China)

    2009-03-15

    Based on the complex system theory and the concept of self organized criticality (SOC) theory, the mechanism of China power grid blackout is studied by analyzing the blackout data in the China power system from 1981 to 2002. The probability distribution functions of various measures of blackout size have a power tail. The analysis of scaled window variance and rescaled range statistics of the time series show moderate long time correlations. The blackout data seem consistent with SOC; the results obtained show that SOC dynamics may play an important role in the dynamics of power systems blackouts. It would be possible to propose novel approaches for understanding and controlling power systems blackouts. (author)

  15. Study on self organized criticality of China power grid blackouts

    Energy Technology Data Exchange (ETDEWEB)

    Zhao Xingyong [Department of Electrical Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Minhang District, Shanghai 200240 (China)], E-mail: zhaoxingyong@sjtu.edu.cn; Zhang Xiubin; He Bin [Department of Electrical Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Minhang District, Shanghai 200240 (China)

    2009-03-15

    Based on the complex system theory and the concept of self organized criticality (SOC) theory, the mechanism of China power grid blackout is studied by analyzing the blackout data in the China power system from 1981 to 2002. The probability distribution functions of various measures of blackout size have a power tail. The analysis of scaled window variance and rescaled range statistics of the time series show moderate long time correlations. The blackout data seem consistent with SOC; the results obtained show that SOC dynamics may play an important role in the dynamics of power systems blackouts. It would be possible to propose novel approaches for understanding and controlling power systems blackouts.

  16. Severe accident analysis for level 2 PSA of SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Yong; Lee, Jeong Hun; Kim, Jong Uk; Yoo, Tae Geun; Chung, Soon Il; Kim, Min Gi [FNC Technology Co., Seoul (Korea, Republic of)

    2010-12-15

    The objectives of this study are to produce data for level 2 PSA and evaluation results of severe accident by analyzing severe accident sequence of transient events, producing fault tree of containment systems and evaluating direct containment heating of the SMART. In this project, severe accident analysis results were produced for general transient, loss of feedwater, station blackout, and steam line break events, and based on the results, design safety of SMART was verified. Also, direct containment heating phenomenon of the SMART was evaluated using TCE methodology. For level 2 PSA, fault tree of the containment isolation system, reactor cavity flooding system, plant chilled water system, and reactor containment building HVAC system was produced and analyzed

  17. Study on risk factors of PWR accidents beyond design basis

    International Nuclear Information System (INIS)

    Ahn, Seung Hoon; Nah, W. J.; Bang, Y. S.; Oh, D. Y.; Oh, S. H.

    2005-01-01

    Development of the regulatory guidelines for Beyond Design Basis Accidents (BDBA) with high risk requires a detailed investigation of major factors contributing to the event risk. In this study, each event was classified by the level of risk, based on the probabilistic safety assessment results, so that BDBA with high risk could be selected, with consideration of foreign and domestic regulations, and operating experiences. The regulatory requirements and technical backgrounds for the selected accidents were investigated, and effective regulatory approaches for risk reduction of the accidents. The following conclusions were drawn from this study: - Selected high risk BDBA is station blackout, anticipated without scram, total loss of feedwater. - Major contributors to the risk of selected events were investigated, and appropriate assessment of them was recommended for development of the regulatory guidelines

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  19. Nuclear-station post-accident liquid-sampling system: developed by Duke Power Company

    International Nuclear Information System (INIS)

    Burton, D.A.; Birch, M.L.; Orth, W.C.

    1981-01-01

    The accident at Three Mile Island showed that means must be provided to determine the radioactivity levels in high activity liquid and gaseous systems of a nuclear power plant without undue radiation exposure to personnel. The Duke Power Post Accident Liquid Sampling System provides the means for obtaining diluted liquid samples and diluted dissolved gas samples following a reactor accident involving substantial core damage. Their approach yields a straightforward engineering solution at a fraction of the cost of other systems. A description of the system, general design criteria, and color coded flow diagrams are included

  20. Post-accident cleanup of radioactivity at the Three Mile Island Nuclear Power Station

    International Nuclear Information System (INIS)

    Brooksbank, R.E.; Armento, W.J.

    1980-02-01

    The technical staff of the President's Commission on the Accident at Three Mile Island (TMI) requested that Oak Ridge National Laboratory (ORNL) prepare documentation concerned with the cleanup of radioactivity on the Three Mile Island site following the March 28, 1979 accident. The objective of this report is to provide information in a summarized form, which will be of direct usefulness to the commissioners. The information contained herein includes discussion of on-site assistance and accomplishments following the accident, flowsheet development for the TMI recovery team (by the Technical Advisory Group), and the numerous reports already generated on the TMI cleanup and recovery

  1. Addenda to the second update of the Fukushima Daiichi Nuclear Power Station accident. June 1 to August 31, 2011

    International Nuclear Information System (INIS)

    2011-01-01

    These addenda provide the figures and tables for helping readers to understand the article titled 'the second update of the Fukushima Daiichi Nuclear Power Station (NPS) accident' by SHIBUTANI Yu. These figures and tables are mainly referred from 'Additional Report of the Japanese Government to the IAEA - The Accident at the Tokyo Electric Power Company Inc. (TEPCO) Fukushima Daiichi NPS - September 2011, Nuclear Emergency Response Headquarters Government of Japan' and the website of Prime Minster of Japan and His Cabinet, Nuclear and Industrial Safety Agency (NISA), Ministry of Education, Culture, Sports, Science and Technology (MEXT), TEPCO and Japan Atomic Industrial Forum Inc. (JAIF). The contents of this addenda cover (1) summary of 28 learned lessons, (2) status of each unit of Fukushima Daiichi NPS, (3) alternative core cooling system, (4) spent fuel pool alternative cooling system, (5) outline of waste water storage and treatment system, (6) prevention of environmental release of radioactive materials and monitoring, (7) environmental effect caused by the accident, and (8) influence of Fukushima Daiichi accident on electricity supply in Japan. (author)

  2. Probabilistic risk assessment (PRA) update in light of the accident at Fukushima Daiichi Nuclear Power Station - 15461

    International Nuclear Information System (INIS)

    Maeda, K.; Abe, H.; Hirokawa, N.; Satou, C.

    2015-01-01

    We have performed internal and external event probabilistic risk assessments (PRA) for boiling water reactor power nuclear plants to identify the important accident sequence groups and to evaluate the effectiveness of the additional severe accident measures, regarding to the new regulatory requirements implemented after the accident at Fukushima Daiichi Nuclear Power Station in Japan in 2011. In addition, we will further update our PRA by extracting problems and improvements from the current PRA, by catching up the state-of-the-art knowledge, modern PRA methodologies in order to contribute voluntarily to safety improvement as well as to comply with regulations. In this document, prior to the extensive PRA updates, we would describe technical contents and qualitative results about PRA updates that have been performed preliminary so far, especially about the external event (seismic) PRA and how to model the additionally deployed severe accident measures (e.g. power supply car, fire engine) so that they can be function external hazards, such as component failure rate of equipment, human reliability 'out of control room', and mission time extension. (authors)

  3. Modelling of blackout sequence at Atucha-1 using the MARCH3 code

    International Nuclear Information System (INIS)

    Baron, J.; Bastianelli, B.

    1997-01-01

    This paper presents the modelling of a complete blackout at the Atucha-1 NPP as preliminary phase for a Level II safety probabilistic analysis. The MARCH3 code of the STCP (Source Term Code Package) is used, based on a plant model made in accordance with particularities of the plant design. The analysis covers all the severe accident phases. The results allow to view the time sequence of the events, and provide the basis for source term studies. (author). 6 refs., 2 figs

  4. A study on the implementation effect of accident management strategies on safety

    International Nuclear Information System (INIS)

    Jae, Moo Sung; Kim, Dong Ha; Jin, Young Ho

    1996-01-01

    This paper presents a new approach for assessing accident management strategies using containment event trees(CETs) developed during an individual plant examination (IPE) for a reference plant (CE type, 950 MWe PWR). Various accident management strategies to reduce risk have been proposed through IPE. Three strategies for the station blackout sequence are used as an example: 1) reactor cavity flooding only, 2) primary system depressurization only, and 3) doing both. These strategies are assumed to be initiated at about the time of core uncovery. The station blackout (SBO) sequence is selected in this paper since it is identified as one of the most threatening sequences to safety of the reference plant. The effectiveness and adverse effects of each accident management strategy are considered synthetically in the CETs. A best estimate assessment for the developed CETs using data obtained from NUREG-1150, other PRA results, and the MAAP code calculations is performed. The strategies are ranked with respect to minimizing the frequencies of various containment failure modes. The proposed approach is demonstrated to be very flexible in that it can be applied to any kind of accident management strategy for any sequence. 9 refs., 3 figs., 2 tabs. (author)

  5. Simulation and transient analyses of a complete passive heat removal system in a downward cooling pool-type material testing reactor against a complete station blackout and long-term natural convection mode using the RELAP5/3.2 code

    Directory of Open Access Journals (Sweden)

    Afshin Hedayat

    2017-08-01

    Full Text Available In this paper, a complete station blackout (SBO or complete loss of electrical power supplies is simulated and analyzed in a downward cooling 5-MW pool-type Material Testing Reactor (MTR. The scenario is traced in the absence of active cooling systems and operators. The code nodalization is successfully benchmarked against experimental data of the reactor's operating parameters. The passive heat removal system includes downward water cooling after pump breakdown by the force of gravity (where the coolant streams down to the unfilled portion of the holdup tank, safety flapper opening, flow reversal from a downward to an upward cooling direction, and then the upward free convection heat removal throughout the flapper safety valve, lower plenum, and fuel assemblies. Both short-term and long-term natural core cooling conditions are simulated and investigated using the RELAP5 code. Short-term analyses focus on the safety flapper valve operation and flow reversal mode. Long-term analyses include simulation of both complete SBO and long-term operation of the free convection mode. Results are promising for pool-type MTRs because this allows operators to investigate RELAP code abilities for MTR thermal–hydraulic simulations without any oscillation; moreover, the Tehran Research Reactor is conservatively safe against the complete SBO and long-term free convection operation.

  6. Simulation and transient analyses of a complete passive heat removal system in a downward cooling pool-type material testing reactor against a complete station blackout and long-term natural convection mode using the RELAP5/3.2 code

    Energy Technology Data Exchange (ETDEWEB)

    Hedayat, Afshin [Reactor and Nuclear Safety School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

    2017-08-15

    In this paper, a complete station blackout (SBO) or complete loss of electrical power supplies is simulated and analyzed in a downward cooling 5-MW pool-type Material Testing Reactor (MTR). The scenario is traced in the absence of active cooling systems and operators. The code nodalization is successfully benchmarked against experimental data of the reactor's operating parameters. The passive heat removal system includes downward water cooling after pump breakdown by the force of gravity (where the coolant streams down to the unfilled portion of the holdup tank), safety flapper opening, flow reversal from a downward to an upward cooling direction, and then the upward free convection heat removal throughout the flapper safety valve, lower plenum, and fuel assemblies. Both short-term and long-term natural core cooling conditions are simulated and investigated using the RELAP5 code. Short-term analyses focus on the safety flapper valve operation and flow reversal mode. Long-term analyses include simulation of both complete SBO and long-term operation of the free convection mode. Results are promising for pool-type MTRs because this allows operators to investigate RELAP code abilities for MTR thermal–hydraulic simulations without any oscillation; moreover, the Tehran Research Reactor is conservatively safe against the complete SBO and long-term free convection operation.

  7. 47 CFR 76.111 - Cable sports blackout.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 4 2010-10-01 2010-10-01 false Cable sports blackout. 76.111 Section 76.111... CABLE TELEVISION SERVICE Network Non-duplication Protection, Syndicated Exclusivity and Sports Blackout § 76.111 Cable sports blackout. (a) No community unit located in whole or in part within the specified...

  8. 47 CFR 76.127 - Satellite sports blackout.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 4 2010-10-01 2010-10-01 false Satellite sports blackout. 76.127 Section 76... Sports Blackout § 76.127 Satellite sports blackout. (a) Upon the request of the holder of the broadcast rights to a sports event, or its agent, no satellite carrier shall retransmit to subscribers within the...

  9. Overview of BWR Severe Accident Sequence Analyses at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1983-01-01

    Since its inception in October 1980, the Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory (ORNL) has completed four studies including Station Blackout, Scram Discharge Volume Break, Loss of Decay Heat Removal, and Loss of Injection accident sequences for the Browns Ferry Nuclear Plant. The accident analyses incorporated in a SASA study provide much greater detail than that practically achievable in a Probabilistic Risk Assessment (PRA). When applied to the candidate dominant accident sequences identified by a PRA, the detailed SASA results determine if factors neglected by the PRA would have a significant effect on the order of dominant sequences. Ongoing SASA work at ORNL involves the analysis of Anticipated Transients Without Scram (ATWS) sequences for Browns Ferry

  10. Analysis of accident progression in the TEPCO Fukushima Daiichi Nuclear Power Station

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    One of the objectives of this study is to investigate the early stage of the TEPCO Fukushima Daiichi accident and to check the validity of the countermeasures against the accident. Last year the early stage of the accident was analyzed with use of RELAP5 code, and the longer term analysis was done by MELCOR code. This year, the simulation of reactor water level instrumentation behavior by MELCOR code was performed. Another objective of this study is to analyze of the long term cooling after the Fukushima Daiichi accident by TRACE5 code. In order to simulate the cooling conditions in Fukushima plants after the accident, the parametric calculations were done on the assumption of the existence of steam/liquid leak in Reactor Pressure Vessel (RPV) and Pressure Containment Vessel (PCV) and the variety of debris distribution in RPV and PCV. As a result, the debris distribution in RPV and PCV was estimated by referring plant parameter such as reactor pressure and temperature. (author)

  11. Hypersonic Cruise and Re-Entry Radio Frequency Blackout Mitigation: Alleviating the Communications Blackout Problem

    Science.gov (United States)

    Manning, Robert M.

    2017-01-01

    The work presented here will be a review of a NASA effort to provide a method to transmit and receive RF communications and telemetry through a re-entry plasma thus alleviating the classical RF blackout phenomenon.

  12. Three Mile Island nuclear generating station accident of March 28, 1979

    International Nuclear Information System (INIS)

    Aitken, J.H.; Johnson, A.C.; Kelly, R.J.; Wong, K.Y.

    1979-04-01

    The government of Ontario dispatched a Scientific Assessment Team to the Three Mile Island nuclear power plant to assess the consequences of an accident which occurred at unit 2 on 1979 March 28. The team's objectives were to acquire up-to-the-minute information concerning the accident, study the potential for environmental impacts on Ontario, observe the outcome of off-site emergency procedures, and offer assisance from Ontario should it appear of value. The findings, observations, and impressions of the team are summarized. (OST)

  13. Activities of radionuclides in the Pacific coastal area of Fukushima since the TEPCO Fukushima Daiichi Nuclear Power Station accident - Activities of radionuclides in the coast area off Fukushima after TEPCO's Fukushima Daiichi Nuclear Power Station accident

    Energy Technology Data Exchange (ETDEWEB)

    Aono, Tatsuo; Fukuda, Miho; Yoshida, Satoshi [National Institute of Radiological Sciences, 263-8555, Chiba (Japan); Sohtome, Tadahiro; Mizuno, Takuji [Fukushima Prefecture Fisheries Experimental Station, 970-0316, Fukushima (Japan); Igarashi, Satoshi [Fukushima Prefecture Fisheries Experimental Station, 970-0316, Fukushima (Japan); Fukushima Prefecture Sea-Farming Association, 970-8044, Fukushima (Japan); Ito, Yukari; Kanda, Jota; Ishimaru, Takashi [Tokyo University of Marine Science and Technology, 108-0075, Tokyo (Japan)

    2014-07-01

    The accident of TEPCO's Fukushima Daiichi Nuclear Power Station (FDNPS) has caused the release of the huge quantities of radionuclide by the 2011 Great East Japan Earthquake and Tsunami, and then the serious problems gave rise to pollution in marine environment widely in the coast area off Fukushima. Monitoring of radioactivity in seawater and biota are important for understanding the dispersion of radionuclides and the effects of radioecology in the marine environment around the coast of Fukushima and the Pacific. The activities of Cs-134 and Cs-137 in seawater decreased exponentially and then were almost same levels before the accident around off Fukushima after about three years from the accident. However, the high activities of radio caesium (Cs) have been monitored in marine biota off Fukushima. The aims of the present study were to examine the temporal changes in radioactivity and to clarify the variation factor and the effect of radioecology in marine biota. Cs-134 and Cs-137, and Ag-110m, were released by this accident, determined in biota sample such as the plankton, fish and benthos, although it is well-known that molluscs and crustaceans concentrate silver in visceral parts. However, Sr-90 was not detected and the activities of plutonium were almost same level before this accident in the marine biota around off Fukushima. Concentration ratios of Cs (CR-Cs) in marine organism were from 2.6 E+1 in the muscle part of squid to 1.0 E+4 in the viscera of clam. The large differences in CR-Cs by the parts of marine organism were not observed. It is suggested that rapid change in the activities of radio Cs and silver in seawater, resuspension of particles from sediments and food chain effects led to high radionuclide activities in marine biota after this accident. CR-Cs in plankton was also calculated with the activities in seawater, which were collected around sampling area during this monitoring period. These resulting values ranged from 5.8 E+1 to 7.8 E+2

  14. Effects of the Chernobyl and Fukushima nuclear accidents on atmospheric electricity parameters recorded at Polish observation stations

    Science.gov (United States)

    Kubicki, Marek; Baranski, Piotr; Odzimek, Anna; Michnowski, Stanislaw; Myslek-Laurikainen, Bogna

    2013-04-01

    We analyse the atmospheric electricity parameters, measured at Polish geophysical stations in Swider, Poland, and Hornsund, Spitsbergen, in connection with the radioactive incident in Fukushima, Japan, beginning on 11 March 2011, following the 9.0 earthquake and tsunami. We compare our results with the situation during and after the Chernobyl disaster on April 26, 1986, when the radioactive fallout detected at Swider increased in the last week of April 1986, from 4.111 to 238.7 Bq/m2 and up to 967.0 Bq/m2 in the second week of May 1986 - what was more than 235 times greater than the values measured prior to that accident. Besides the electric field especially the electric conductivity is very sensitive to the radioactive contamination of the air. Thus we postulate that these two measurements should be run at geophysical stations over the world and used as a relatively simple and low-cost tool for continuous monitoring of possible hazard caused by nuclear power plant accidents.

  15. Lessons learned from early direct measurements at Fukushima Medical University after the Fukushima Nuclear Power Station accident

    Energy Technology Data Exchange (ETDEWEB)

    Miyazaki, Makoto; Ohba, Takashi; Ohtsuru, Akira [Fukushima Medical Univ., Dept. of Radiation Health Management, Fukushima, Fukushima (Japan)

    2012-11-15

    The Fukushima Daiichi Nuclear Power Station (FDNPS) accident resulted in a month-long discharge of radioactive materials into the environment. These radioactive materials were detected at Fukushima Medical University (FMU), which is 57 km northwest of the FDNPS. Significant levels of six nuclides (i.e., {sup 131}I, {sup 132}Te, {sup 132}I, {sup 133}Xe, {sup 134}Cs, and {sup 137}Cs) were detected by a whole body counter (WBC) on March 15, 2011 when the ambient dose rate was suddenly elevated for the first time. This WBC has a dual detector system consisting of two NaI(Tl) detectors and two Ge detectors. We conducted periodical measurements of 32 humans and the background using the WBC. Because the three nuclides {sup 131}I, {sup 134}Cs and {sup 137}Cs were still detected in the background by the WBC a few months after the accident, accurate WBC measurements were difficult. Here we describe the limitations of our measurements conducted in the early stage of the FDNPS accident. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

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

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

    International Nuclear Information System (INIS)

    Nomoto, Tsutomu

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-15

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

  19. Analysis of Fukushima Daiichi Nuclear Power Station severe accident using MAAP4.05

    International Nuclear Information System (INIS)

    Yoo, Jae S.; Suh, Kune Y.; Kim, Dong M.

    2011-01-01

    Rather extensive reactor core meltdown and partial melt-through that took place at the Fukushima Daiichi nuclear power plants (NPPs) on March 11, 2011 had been caused by a massive earthquake followed by tsunami rarely seen in history. The happening had turned into an unprecedented serious accident since the Chernobyl Unit 4 in 1986 that extended over multiple reactors simultaneously. A previous documentary survey, NUREG-1150 report provides significant insights into how a severe accident might develop at a boiling water reactor (BWR) and the range of consequences. NUREG-1150 did identify the importance of loss of power accidents for a BWR. This paper describes recent analyses of the Fukushima Daiichi NPPs severe accident. Calculations were performed with the MAAP4.05 code by modifying the parameter file for the Peach Bottom Unit 2, a BWR 4 type in the Mark-I containment. Generally this is the same type of reactor as Fukushima Daiichi Units 2 and 3. This resulted in good understanding of the response of this type of early BWRs to prolonged loss of diesel generators and batteries. There is clearly vulnerability with this early type of BWRs which would be less onerous than later existing plant and new build designs. This analysis, however, was based on rather limited amount of information obtained at the time of preparing this report and adopted various estimates and assumptions for conditions necessary to run the analysis. Hence there persists considerable uncertainty in the results. MAAP4.05 was run pursuant to the observed data and chronology of Fukushima Daiichi Units 1 through 3 reported by the Tokyo Electric Power Company (TEPCO) as well as the Japanese Government. Severe accident scenarios have not only gone far beyond the design basis, but also exceeded the extent of multiple breakdowns assumed in the preparation for such accident management measures as the malfunction or loss of all the emergency core cooling system (ECCS) combined with the extended loss of

  20. Modelling and analysis of severe accidents for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Tusheva, Polina

    2012-01-01

    Accident conditions involving significant core degradation are termed severe accidents /IAEA: NS-G-2.15/. Despite the low probability of occurrence of such events, the investigation of severe accident scenarios is an important part of the nuclear safety research. Considering a hypothetical core melt down scenario in a VVER-1000 light water reactor, the early in-vessel phase focusing on the thermal-hydraulic phenomena, and the late in-vessel phase focusing on the melt relocation into the reactor pressure vessel (RPV) lower head, are investigated. The objective of this work is the assessment of severe accident management procedures for VVER-1000 reactors, i.e. the estimation of the maximum period of time available for taking appropriate measures and particular decisions by the plant personnel. During high pressure severe accident sequences it is of prime importance to depressurize the primary circuit in order to allow for effective injection from the emergency core cooling systems and to avoid reactor pressure vessel failure at high pressure that could cause direct containment heating and subsequent challenge to the containment structure. Therefore different accident management measures were investigated for the in-vessel phase of a hypothetical station blackout accident using the severe accident code ASTEC, the mechanistic code ATHLET and the multi-purpose code system ANSYS. The analyses performed on the PHEBUS ISP-46 experiment, as well as simulations of small break loss of coolant accident and station blackout scenarios were used to contribute to the validation and improvement of the integral severe accident code ASTEC. Investigations on the applicability and the effectiveness of accident management procedures in the preventive domain, as well as detailed analyses on the thermal-hydraulic phenomena during the early in-vessel phase of a station blackout accident have been performed with the mechanistic code ATHLET. The results of the simulations show, that the

  1. Accidents in the operation of nuclear power stations. Action for damages against foreign operators

    International Nuclear Information System (INIS)

    Willner, K.

    1986-01-01

    On the occasion of a lecture evening of the Leo Goodman Library in Munich questions of civil liability of foreign reactor operators in cases of nuclear accidents were discussed by participants of various universities. Special subjects were i.a. problems of civil procedural and insurance law, absolute liability according to sec. 25 Atomic Energy Act as well as questions of applicable law. (WG) [de

  2. The third update of the Fukushima Daiichi Nuclear Station accident. September 1 through November 30, 2011

    International Nuclear Information System (INIS)

    Shibutani, Yu

    2011-01-01

    This article provides the third update of the Fukushima Daiichi accident that occurred on March 11, 2011. In the report of the first update of the Fukushima Daiichi accident on March 11 through May 31, the situation was reported on both on-site and off-site of the Fukushima Daiichi, including; failed cooldown of decay heat and meltdown of stricken reactors; emergency evacuation of local residents, radioactive contamination, spread of biased rumors by the information closure by government, regional cooperation with China, Taiwan and Korea, and visit of IAEA investigation team to Japan. The report of the second update on June 1 through August 31 reported the issues of, harsh public criticism against government and electric power companies, results of the public opinion poll, a sort of gentlemen's agreements between nuclear power companies and local governments which would be peculiar tradition in Japan, the first revision of the road map to cold shutdown of stricken reactors, and submission of report on Fukushima Daiichi accident to International Atomic Energy Agency (IAEA). This article provides the third update from September 1 through November 30, particularly on the issues of the second revision of the road map where the “cold shutdown” state should be reached before the end of 2011, the overview of governmental organization on the overall energy and nuclear policy, and the establishment of the nuclear disaster response headquarters (HQ) in the Prime Minister's Office. The HQ in collaboration with Tokyo Electric Power Company (TEPCO) decided the framework of road map plans, provision of various assistance and compensation for the residents affected by the nuclear incident, redefinition by the Nuclear Safety Commission for evacuation areas, recovery process of radioactive decontaminated areas, investigation and verification of the Fukushima Daiichi accident, reorganization of TEPCO management and financial system, establishment of damage compensation scheme

  3. Learning non-technical skill lessons from testimony given in the investigation of the nuclear accident at the Fukushima Nuclear Power Stations

    International Nuclear Information System (INIS)

    Hikono, Masaru; Sakuda, Hiroshi; Matsui, Yuko; Goto, Manabu; Kanayama, Masaki

    2016-01-01

    The Government Investigation Committee on the Accident at the Fukushima Nuclear Power Stations interviewed individuals concerned. The hearing records, published in 2014, are considered to have valuable lessons for power station managers who encounter severe accidents. In this study, descriptions from the hearing records were extracted as lessons for managers. The extractions were classified by the subject (for whom the lessons are intended), and the category of the non-technical skills. The results showed the possibility of pointing out the lessons in accordance with responsibilities. (author)

  4. Long-term security of electrical and control engineering equipment in nuclear power stations to withstand a loss of coolant accident

    International Nuclear Information System (INIS)

    Mueller, H.

    1996-01-01

    Electrical and control engineering equipment, which has to function even after many years of operation in the event of a fault in a saturated steam atmosphere of 160 C maximum, is essential in nuclear power stations in order to control a loss of coolant accident. The nuclear power station operators have, for this purpose, developed verification strategies for groups of components, by means of which it is ensured that the electrical and control engineering components are capable of dealing with a loss of coolant accident even at the end of their planned operating life. (orig.) [de

  5. Assessment of radiation doses to the public in areas contaminated by the Fukushima Daiichi Nuclear Power Station accident

    International Nuclear Information System (INIS)

    Takahara, Shogo; Iijima, Masashi; Shimada, Kazumasa; Kimura, Masanori; Homma, Toshimitsu

    2013-01-01

    In the areas contaminated by radioactive materials due to the Fukushima Daiichi Nuclear Power Station accident, many residents are exposed to radiation through various exposure pathways. Dose assessment is important for providing appropriate protection to the people and clarifying the impact of the accident. The aim of this study is to provide preliminary results of the assessment of radiation doses received by the inhabitants of Fukushima Prefecture. To assess the doses realistically and comprehensively, a probabilistic approach was adopted using data that reflected realistic environmental trends and lifestyle habits in Fukushima Prefecture. In the first year after the contamination, the 95th percentile of the annual effective dose received by the inhabitants evacuated from the evacuation areas and the deliberate evacuation areas was mainly in the 1-10 mSv dose band. However, the 95th percentile of the dose received by some outdoor workers and inhabitants evacuated from highly contaminated areas was in the 10-50 mSv dose band. The doses due to external exposure to deposited radionuclides were the dominant exposure pathway, and their contributions were about 90% under prevailing contamination conditions in Fukushima Prefecture. In addition, 20%-30% of the lifetime effective dose was delivered during the first year after the contamination. (author)

  6. Severe accident analysis of a steam generator tube rupture accident using MAAP-CANDU to support level 2 PSA for the Point Lepreau Generating Station Refurbishment Project

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Brown, M.J. [Canadian Nuclear Laboratories, Chalk River, ON (Canada)

    2015-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station. The MAAP-CANDU code was used to simulate the progression of postulated severe core damage accidents and fission product releases. This paper discusses the results for the reference case of the Steam Generator Tube Rupture initiating event. The reference case, dictated by the Level 2 Probabilistic Safety Assessment, was extreme and assumed most safety-related plant systems were not available: all steam generator feedwater; the emergency water supply; the moderator, shield and shutdown cooling systems; and all stages of emergency core cooling. The reference case also did not credit any post Fukushima lessons or any emergency mitigating equipment. The reference simulation predicted severe core damage beginning at 3.7 h, containment failure at 6.4 h, moderator boil off by 8.2 h, and calandria vessel failure at 42 h. A total release of 5.3% of the initial inventory of radioactive isotopes of Cs, Rb and I was predicted by the end of the simulation (139 h). Almost all noble gas fission products were released to the environment, primarily after the containment failure. No hydrogen/carbon monoxide burning was predicted. (author)

  7. Severe accident consequence mitigation by filtered containment venting at Canadian nuclear power plants

    International Nuclear Information System (INIS)

    Lebel, Luke S.; Morreale, Andrew C.; Korolevych, Volodymyr; Brown, Morgan J.; Gyepi-Garbrah, Sam

    2017-01-01

    Highlights: • Use of filtered containment venting during a severe accident assessed. • Severe accident simulations performed using MAAP-CANDU and ADDAM. • Flow capacity, initiation protocols, efficiency, mass and thermal loading evaluated. • Efficient, robust system drastically reduces accident consequences. - Abstract: Having the capability to use filtered containment venting during a severe nuclear accident can significantly reduce its overall consequences. This study employs the MAAP-CANDU severe accident analysis code and the ADDAM atmospheric dispersion code to study the progression of: an unmitigated station blackout accident at a generic pressurized heavy water reactor, the release of radioactive material into the environment, the subsequent dispersion of the fission products through the atmosphere and the subsequent consequences (evacuation radius). The goal is to evaluate the application of filtered venting as an accident mitigation technology. Several aspects of filtered containment venting system design, like flow capacity, initiation protocols, filter efficiency, mass loading, and thermal loading are considered. An efficient and robust filtered containment venting system can reduce the amount of radiological materials emitted during an accident by 25 times or more, and as a result considerably reduce the off-site consequences of an accident.

  8. The use of influence diagrams for evaluating severe accident management strategies

    International Nuclear Information System (INIS)

    Jae, M.; Apostolakis, G.E.

    1992-01-01

    In this paper, the influence diagram, a new analytical tool for developing and evaluating severe accident management strategies, is presented. Influence diagrams are much simpler than decision trees because they do not lead to the large number of branches that are generated when decision trees are used in realistic problems; furthermore, they show explicitly the dependencies between the variables of the problem. One of the accident management strategies proposed for light water reactors, flooding the reactor cavity as a means of preventing vessel breach during a short-term station blackout sequence, is presented. The influence diagram associated with this strategy is constructed. Finally, the advantages of using influence diagrams in accident management are explored

  9. Screening and analysis of beyond design basis accident of 49-2 SPR

    International Nuclear Information System (INIS)

    Zhang Yadong; Guo Yue; Wu Yuanyuan; Zou Yao

    2015-01-01

    The beyond design basis accident was analyzed to ensure safe operation of 49-2 Swimming Pool Reactor (SPR) after design life. Because it's difficult to use PSA method, the unconditional assumed severe accidents were adopted to obtain a conservative result. The main conclusions were obtained by analyzing anticipated transients without scram in station blackout (SBO ATWS), horizontal channel rupture, core uncovering after shutdown and emergency response capacity. The results show that the core is safe in SBO ATWS, and the fuel elements will not melt as long as the core are not exposed in 2.5 h in loss of coolant accident caused by horizontal channel rupture and other factors. The passive siphon breaker function and various ways of emergency core makeup can ensure that the core is not exposed. (authors)

  10. Fractal Characteristics Analysis of Blackouts in Interconnected Power Grid

    DEFF Research Database (Denmark)

    Wang, Feng; Li, Lijuan; Li, Canbing

    2018-01-01

    The power failure models are a key to understand the mechanism of large scale blackouts. In this letter, the similarity of blackouts in interconnected power grids (IPGs) and their sub-grids is discovered by the fractal characteristics analysis to simplify the failure models of the IPG. The distri......The power failure models are a key to understand the mechanism of large scale blackouts. In this letter, the similarity of blackouts in interconnected power grids (IPGs) and their sub-grids is discovered by the fractal characteristics analysis to simplify the failure models of the IPG....... The distribution characteristics of blackouts in various sub-grids are demonstrated based on the Kolmogorov-Smirnov (KS) test. The fractal dimensions (FDs) of the IPG and its sub-grids are then obtained by using the KS test and the maximum likelihood estimation (MLE). The blackouts data in China were used...

  11. The second update of the Fukushima Daiichi Nuclear Station accident. June 1 through August 31, 2011

    International Nuclear Information System (INIS)

    Shibutani, Yu

    2011-01-01

    The Fukushima Daiichi accident occurred on March 11, 2011. The Japanese government and Tokyo Electric Power Company (TEPCO) encountered great deal of difficulties to stabilize the nuclear power reactor. Although the maximum radiation level has decreased, it still accounts for one-fifth of the level detected as of end of August 2011. There are currently growing concerns regarding the complexities of nuclear energy that are faced by the society of Japan. Owing to the occurrence of the Fukushima Daiichi accident, the firm belief on absolute safety of Japan's nuclear power plants has suddenly collapsed. Over the past decades, people tended to trust the government and electric power companies, which assured the safety and ensured about the necessity of nuclear power. However the recent disaster that took place in Fukushima has triggered public anger and distrusts that lead to negative campaigns against nuclear energy, in addition to oppositions against the government and nuclear power companies. The harsh public criticism has forced the government to separate the Nuclear and Industrial Safety Agency (NISA) from affiliation with the Ministry of Economy, Trade and Industry (METI). This action was taken since NISA had expressed contradictions with METI, which is the promoting agency for nuclear energy, The nation-wide opinion poll and surveys undertaken by Asahi Shimbun, a leading Japanese daily newspaper, has announced that 74% of respondents supported the policy to phase-out all nuclear power plants with the final goal to completely end the operation of nuclear plants. A mere 14% of the respondents disagree. There are approximately 100,000 or more residents that evacuated from the dangerous Fukushima Daiichi zone and were suffering from immediate threats such as radiation exposure, health problems, sanitary contamination, products shipment ban, business slump, groundless rumors, etc. These conditions have severely weakened their capability to completely recover and

  12. Update of severe accident management guidelines at Point Lepreau Generating Station

    Energy Technology Data Exchange (ETDEWEB)

    Mullin, D.; Robertson, S., E-mail: dmullin@nbpower.com, E-mail: srobertson@nbpower.com [New Brunswick Power Corporation, Point Lepreau Generating Station, NB (Canada); Wilson, D.J., E-mail: wilsondj@nbnet.nb.ca [Atlantic Nuclear Inc., Fredericton, NB (Canada)

    2015-07-01

    In response to the events at Fukushima Dai'ichi, the CANDU Owners Group (COG) embarked on Joint Project JP4426 to update the overall framework for Severe Accident Management Guidelines (SAMG) to reflect lessons learned. This includes enhancements to consider SAMG in the shutdown/low power state, more extensive consideration of Spent (Irradiated) Fuel Bay (SFB) events and to address plant habitability issues. Point Lepreau is in the process of applying the COG methodologies in developing site-specific basis documents and revisions to SAM guidance documents. The paper will also describe SAMG for the Canister site, emergency procedure revisions to deal with extreme natural events and recommendations for evacuation of protective zones based on plant state criteria. (author)

  13. [Experience in organization of joint actions of expert divisions during the accident at P.S. Podporozniy Sayano-Shushenskaya hydroelectric power station].

    Science.gov (United States)

    Kolkutin, V V; Ivanov, P L; Fetisov, V A; Afanas'ev, S A; Dorozhkin, O A; Vognerubov, R N; Kuznetsov, T L

    2010-01-01

    The authors illustrate positive experience in organization and coordination of joint actions of expert divisions of different sectors during the accident at P.S. Podporozniy Sayano-Shushenskaya hydroelectric power station in August 2009. Special emphasis is laid on the participation of experts of quick-reaction teams formed by territorial forensic medical bureaus, mobile and supporting forces from the adjacent regions.

  14. Simulation of accident and restrained transients in PWR nuclear power plant with RELAP 5/MOD 1 computer code

    International Nuclear Information System (INIS)

    Silva Filho, E.

    1986-01-01

    The computer code RELAP5/MOD1 has been utilized to investigate the thermal-hydraulic behaviour of a standard 1300 Mwe pressurized water reactor plant of the KWU design during a station blackout and during a loss-of-coolant accident involving 2% break in the cross-sectional area the cold leg in one of the four loops and located between the pump and the reactor pressure vessel. During the simulations the reactor scram system and the emergency coolant system were considered inactive. (Author) [pt

  15. Volumes of radionuclide into the basins of water while the accident at the Chernobyl nuclear power station and a specifics of radiation situation development in the post-accidents periods

    International Nuclear Information System (INIS)

    Standritchuk, O.Z.; Maksin, V.I.; Goncharuk, V.V.

    1996-01-01

    There was stated total content of radionuclide pollution, rejected to the environment in consequence of the accident at the Chernobyl nuclear power station, specifics of qualitative and quantitative change which supposes the division of post-accident period into five conventional post-accident periods. There were given the data about the levels of main fragmentation radionuclide activity in river water, atmospheric precipitation and sewage of the objects of sanitary treatment in May 1986. According to these data there were estimated the volumes of radioactive pollution rejection to the Kiev basins of water (1.56 centre dot 10 10 Ku, that is equal to 144,57 kg of radionuclides or 3,67 % of their mass in reactor) and their going into the Dnieper river. There was shown an interconnection of all season state of water basins which are near to Chernobyl nuclear power station, with specific development of radiation situation in them after the accident. There was proposed a probated variant of improvement of the traditional technology of drinking water preparation from the open water source within 1-2 post-accident periods

  16. Thermal Hydraulic design parameters study for severe accidents using neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Chang Hyun; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chang, Keun Sun [Sunmoon University, Asan (Korea, Republic of)

    1997-12-31

    To provide the information on severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore, was performed to investigate the effect of thermal hydraulic design parameters on severe accident progression of pressurized water reactors (PWRs). Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among nine parameters. For training, different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3 and 4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout (SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to the other six parameters. 9 refs., 5 tabs. (Author)

  17. Thermal Hydraulic design parameters study for severe accidents using neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Chang Hyun; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chang, Keun Sun [Sunmoon University, Asan (Korea, Republic of)

    1998-12-31

    To provide the information on severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore, was performed to investigate the effect of thermal hydraulic design parameters on severe accident progression of pressurized water reactors (PWRs). Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among nine parameters. For training, different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3 and 4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout (SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to the other six parameters. 9 refs., 5 tabs. (Author)

  18. Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Colman, J.; Hsu, C.J.

    1983-01-01

    In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization

  19. The accident at TEPCO's Fukushima-Daiichi Nuclear Power Station: What went wrong and what lessons are universal?

    Science.gov (United States)

    Omoto, Akira

    2013-12-01

    After a short summary of the nuclear accident at the Fukushima Daiichi Nuclear Power Station, this paper discusses “what went wrong” by illustrating the problems of the specific layers of defense-in-depth (basic strategy for assuring nuclear safety) and “what lessons are universal.” Breaches in the multiple layers of defense were particularly significant in respective protection (a) against natural disasters (first layer of defense) as well as (b) against severe conditions, specifically in this case, a complete loss of AC/DC power and isolation from the primary heat sink (fourth layer of defense). Confusion in crisis management by the government and insufficient implementation of offsite emergency plans revealed problems in the fifth layer of defense. By taking into consideration managerial and safety culture that might have relevance to this accident, in the author's view, universal lessons are as follows: Resilience: the need to enhance organizational capabilities to respond, monitor, anticipate, and learn in changing conditions, especially to prepare for the unexpected. This includes increasing distance to cliff edge by knowing where it exists and how to increase safety margin. Responsibility: the operator is primarily responsible for safety, and the government is responsible for protecting public health and environment. For both, their right decisions are supported by competence, knowledge, and an understanding of the technology, as well as humble attitudes toward the limitations of what we know and what we can learn from others. Social license to operate: the need to avoid, as much as possible regardless of its probability of occurrence, the reasonably anticipated environmental impact (such as land contamination), as well as to build public confidence/trust and a renewed liability scheme.

  20. The accident at TEPCO's Fukushima-Daiichi Nuclear Power Station: What went wrong and what lessons are universal?

    International Nuclear Information System (INIS)

    Omoto, Akira

    2013-01-01

    After a short summary of the nuclear accident at the Fukushima Daiichi Nuclear Power Station, this paper discusses “what went wrong” by illustrating the problems of the specific layers of defense-in-depth (basic strategy for assuring nuclear safety) and “what lessons are universal.” Breaches in the multiple layers of defense were particularly significant in respective protection (a) against natural disasters (first layer of defense) as well as (b) against severe conditions, specifically in this case, a complete loss of AC/DC power and isolation from the primary heat sink (fourth layer of defense). Confusion in crisis management by the government and insufficient implementation of offsite emergency plans revealed problems in the fifth layer of defense. By taking into consideration managerial and safety culture that might have relevance to this accident, in the author's view, universal lessons are as follows: a)Resilience: the need to enhance organizational capabilities to respond, monitor, anticipate, and learn in changing conditions, especially to prepare for the unexpected. This includes increasing distance to cliff edge by knowing where it exists and how to increase safety margin. b)Responsibility: the operator is primarily responsible for safety, and the government is responsible for protecting public health and environment. For both, their right decisions are supported by competence, knowledge, and an understanding of the technology, as well as humble attitudes toward the limitations of what we know and what we can learn from others. c)Social license to operate: the need to avoid, as much as possible regardless of its probability of occurrence, the reasonably anticipated environmental impact (such as land contamination), as well as to build public confidence/trust and a renewed liability scheme

  1. Off-shore contamination by I-131 and Cs-137 from the Fukushima Daiichi nuclear power station accident

    International Nuclear Information System (INIS)

    Pereira, Wagner de Souza; Kelecom, Patrick Vicent; Miyashita, Erika; Universidade Federal Fluminense; Kelecom, Alphonse

    2011-01-01

    On March 11, 2011 the biggest earthquake ever registered in Japan severed off-site power supply to the Fukushima Daiichi Nuclear Power Station. Backup diesel generators began providing electricity to pumps circulating coolant to the reactors, but were knocked out by a large tsunami and the nuclear site lost the ability to maintain proper reactor cooling. This was the beginning of a huge nuclear accident that was assigned an INES maximum rating of 7. On March 21, Japanese authorities reported that the Tokyo Electric Power Company (TEPCO) had detected radioactive materials in seawater. Radioactivity started to be measured by the Japan Atomic Energy Agency every two days in sea water from eight locations, 30km from the coastline. I-131 and Cs-137 were analyzed among other radionuclides. It is the aim of this paper to gather all this information and to discuss the evolution of the radioactive marine contamination during the first month of the accident. Results indicate for surface seawater concentrations ranging from 24.9 to 161.0 Bq/L for I-131 and 11.2 to 186.0 Bq/L for Cs-137, and for deep waters of 1.59-15.0 Bq/L (I-131) and 0.0-11.4 Bq/L (Cs-137). The I-131 concentrations in superficial waters were at or above Japanese regulatory limits in the first days, then lowered during one week to increase again above limits when TEPCO released contaminated water into the ocean, to finally reach not detectable values the last week of April. With the exception of point 4, on April 15, the Cs-137 levels were always well below regulatory limits. (author)

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

    International Nuclear Information System (INIS)

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

    2017-02-01

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

  3. Direct measurements of employees involved in the Fukushima Daiichi Nuclear Power Station accident for internal dose estimates. JAEA's experiences

    Energy Technology Data Exchange (ETDEWEB)

    Kurihara, Osamu; Kanai, Katsuta; Nakagawa, Takahiro; Takada, Chie; Momose, Takumaro; Furuta, Sadaaki [Japan Atomic Energy Agency, Nuclear Fuel Cycle Engineering Laboratories, Tokai, Ibaraki (Japan)

    2012-11-15

    Japan Atomic Energy Agency (JAEA) performed internal dose measurements of employees involved in the Fukushima Daiichi nuclear power station accident. Nuclear Fuel Cycle Engineering Laboratories (NFCEL), one of the JAEA's core centers, examined 560 of these employees by direct (in vivo) measurements during the period from April 20 to August 5 in 2011. These measurements consisted of whole-body counting for radiocesium and thyroid counting for radioiodine. The whole-body counting was conducted with two types of whole-body counters (WBCs): a standing-type WBC with two large NaI(Tl) detectors (Fastscan{sup TM}, Canberra Inc.) and a chair-type WBC with HPGe detectors (GC5021, Canberra Inc.) installed in a shielded chamber made of 20-cm-thick steel. The thyroid counting was mainly performed using one of the two HPGe detectors equipped with the chair-type WBC. The subjects examined in this work were divided into two groups: Group 1 was the first 39 subjects who were measured up to June 17, 2011 and Group 2 was the remaining 521 subjects who were measured on and after June 18, 2011. The performance of our direct measurements was validated by comparing measurement results of the Group 1 subjects using two different methods (e.g., the standing-type WBC vs. the chair-type WBC). Tentative internal dose estimates of the subjects of Group 1 were also performed based on the assumption of a single intake scenario on either March 12, when the first hydrogen explosion occurred at the station or the first day of work after the accident. It was found that the contribution of {sup 131}I to the total internal dose greatly exceeded those of {sup 134}Cs and {sup 137}Cs, the other major nuclides detected in the measurements. The maximum committed effective dose (CED) was found in a male subject whose thyroid content of {sup 131}I was 9760 Bq on May 23, 2011; the CED of this subject was estimated to be 600 mSv including a small contribution of {sup 134}Cs and {sup 137}Cs. The typical

  4. Re-evaluation of Station Blackout in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Eunchan; Shin, Taeyoung [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    This paper proposes a reduction of the uncertainty due to the small number of LOOP events and an estimation of the non-recovery probability after a LOOP event where the operators fail to energize a safety bus using the offsite power recovery during an SBO with recent operating experience. In addition, in this analysis, the CDF is re-evaluated through reflecting the enhancement of the Class-1E battery capacity. For newly constructed KHNP plants, the LOOP frequency and non-recovery probability after a LOOP during an SBO were re-evaluated through integrating the KHNP events into generic data containing broader experiences for PSA. For an initiating event frequency, a new LOOP frequency was calculated through a Bayesian update of the KHNP LOOP frequency using NUREG/CR-6890, which reflects the recent trends and has a large data size. For the non-recovery probability estimation, domestic data were added to the American experiences in the NUREG/CR-6890; these data were fitted to a lognormal distribution in order to reduce the uncertainty due to the small size of the KHNP data. Regarding the battery capacity enhancement, the success criteria during an SBO were re-evaluated considering the longer battery duty time. The CDF was recalculated using the resultant available time for operator action. The changed CDF was reduced by approximately 50% compared with the value before battery improvement. In conclusion, it was quantitatively proven that enlarging the battery capacity to manage SBOs positively affected plant safety. In addition, methods to improve data uncertainty due to the small number of experiences were selected in order to evaluate the LOOP frequency and non-recovery probability after a LOOP for future plants. These efforts contribute to obtaining a realistic risk profile and to prioritizing countermeasures and improvements of vulnerabilities for safety.

  5. Radioactivity monitoring by the official monitoring stations in North-Rhine Westphalia and the Juelich Nuclear Research Centre after the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    1986-01-01

    This official report presents a governmental declaration of the prime minister of NRW, Mr. Rau, concerning the reactor accident at Chernobyl, and a joint declaration of ministers of NRW, concerning the impact of the accident on the Land NRW. These statements are completed by six official reports on radioactivity measurements carried out by the official monitoring stations of the Land and by the KFA Juelich. These reports inform about methods, scope, and results of the measuring campaigns accomplished by the Zentralstelle fuer Sicherheitstechnik (ZFS), the public materials testing office (MPA), the Chemisches Untersuchungsamt, the Landesamt fuer Wasser und Abfall, and the KFA Juelich. (DG) [de

  6. Fukushima. The accident sequence and important causes. Pt. 2/3; Fukushima. Unfallablauf und wesentliche Ursachen. T. 2/3

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph [Oeko-Institut e.V., Darmstadt (Germany). Bereich Nukleartechnik und Anlagensicherheit

    2013-07-01

    In this part on the accident sequence in the NPP Fukushima Daiichi on March 11, 2011 the important safety systems of a nuclear power plant are described, including the design of a nuclear boiling water reactor with Mark-II type containment, the high-pressure injection system and the systems for afterheat removal. The chronology of the accident progress in the NPP units 1-3 is described. The units 4-6 were shutdown due to revision work. Due to the earthquake an electric power transformation station close to the NPP site and the power poles were destroyed, the redundant power supply of the neighboring electricity supplier Tohoku did not work. All emergency diesel generators were flooded and destroyed resulting in the so-called station blackout. Firefighting trucks and materials for radiation protection and the infrastructure at the NPP site were destroyed. The release of radioactivity induced a severe contamination of the reactor site.

  7. Characteristics of debris in the lower head of a BWR in different severe accident scenarios

    International Nuclear Information System (INIS)

    Phung, Viet-Anh; Galushin, Sergey; Raub, Sebastian; Goronovski, Andrei; Villanueva, Walter; Kööp, Kaspar; Grishchenko, Dmitry; Kudinov, Pavel

    2016-01-01

    Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small ( 100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small variations of the input

  8. Characteristics of debris in the lower head of a BWR in different severe accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Phung, Viet-Anh, E-mail: vaphung@kth.se; Galushin, Sergey, E-mail: galushin@kth.se; Raub, Sebastian, E-mail: raub@kth.se; Goronovski, Andrei, E-mail: andreig@kth.se; Villanueva, Walter, E-mail: walterv@kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

    2016-08-15

    Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20 tons) and large (>100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small

  9. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

  10. Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

    International Nuclear Information System (INIS)

    Park, Soo Yong; Ahn, Kwang Il

    2015-01-01

    Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO

  11. Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

  12. Social Capital Enhanced Disaster Preparedness and Health Consultations after the 2011 Great East Japan Earthquake and Nuclear Power Station Accident.

    Science.gov (United States)

    Hasegawa, Makoto; Murakami, Michio; Takebayashi, Yoshitake; Suzuki, Satoshi; Ohto, Hitoshi

    2018-03-14

    After the Great East Japan Earthquake and the subsequent Fukushima Daiichi Nuclear Power Station accident in 2011, there was a strong demand to promote disaster preparedness approaches and health checkups for the prevention of lifestyle diseases. This study examined the yearly change in the percentage of those who prepared for disasters and who utilized health checkups in Fukushima Prefecture, and identified the factors governing disaster preparedness and utilization of health checkups. We used the public opinion survey from 2011 to 2015 ( n = 677-779 each year) on prefectural policies that is conducted every year by the Fukushima Prefecture government Public Consultation Unit. We found that the percentage of those who prepare for disasters decreased, while that for health checkups did not significantly change. With regard to disaster preparedness, experiences of disaster enhance disaster preparedness, while bonds with other local people help to maintain preparedness. For health checkups, familiarity with the welfare service was the most important factor governing such consultations. The findings suggest that social capital should be promoted in order to improve disaster preparedness. The findings also suggest that residents' accessibility to medical and welfare services is also important in promoting the utilization of health checkups.

  13. Geographical distribution of radioactive nuclides released from the Fukushima Daiichi Nuclear Power Station accident in eastern Japan

    International Nuclear Information System (INIS)

    Ishida, Masanobu; Umetsu, Kohei; Sugimoto, Miyabi; Yamaguchi, Yuta; Yamazaki, Hideo; Nakagawa, Ryota

    2013-01-01

    The geographical distribution of radioactive nuclides released from the Fukushima Daiichi Nuclear Power Station accident in metropolitan areas located in eastern Japan was investigated. The radioactive contamination of environmental samples, including soil and biological materials, was analyzed. The concentrations of 131 I, 134 Cs, and 137 Cs in the soil samples collected from Fukushima City were 122000, 11500 and 14000 Bq/kg on 19th March 2011 and 129000, 11000 and 13700 Bq/kg on 26th March 2011, for the three nuclides respectively. The concentrations of 131 I, 134 Cs and 137 Cs in the soil samples collected from March-June 2011 from study sites ranged from 240 to 101000, 28 to 26200, and 14 to 33700 Bq/kg, respectively. In Higashiosaka City, it began to detect those radioactive nuclides in the atmospheric airborne dust from 25th March. Radioactive fission products 95 Zr- 95 Nb were detected on 18th April 2011. Biological samples collected from Tokyo Bay were studied. The maximum concentrations of 134 Cs and 137 Cs detected in the biological samples were 12.2 and 19.2 Bq/kg, which were measured in goby. 131 I was not detected in the biological samples however, trace amounts of the short half-life nuclide 110m Ag were found in the shellfish samples. (author)

  14. Radioactivity leakage accidents in the feed water heater and the general drainage of the Tsuruga Nuclear Power Station of Japan Atomic Power Company

    International Nuclear Information System (INIS)

    1981-01-01

    In the Tsuruga Nuclear Power Station, JAPC on the shell on extracted-steam side in B system of No. 4 feed water heater, drain water leakage occurred twice in January, 1981. Then, 61 pCi/g cobalt-60 and 10 pCi/g manganese-54 were detected in soil at the outlet of general drainage on April 17, 1981. The cause was found to be the overflow of radioactive liquid waste in the filter sludge storage tank on March 8, the same year. On-the-spot inspection was subsequently made by the Agency of Natural Resources and Energy on both leakage accidents. The results of inspections are described as follows: the course of leakage accident, and also the measures taken to JAPC in connection with the two leakage accidents. (J.P.N.)

  15. WHEN MODEL MEETS REALITY – A REVIEW OF SPAR LEVEL 2 MODEL AGAINST FUKUSHIMA ACCIDENT

    Energy Technology Data Exchange (ETDEWEB)

    Zhegang Ma

    2013-09-01

    The Standardized Plant Analysis Risk (SPAR) models are a set of probabilistic risk assessment (PRA) models used by the Nuclear Regulatory Commission (NRC) to evaluate the risk of operations at U.S. nuclear power plants and provide inputs to risk informed regulatory process. A small number of SPAR Level 2 models have been developed mostly for feasibility study purpose. They extend the Level 1 models to include containment systems, group plant damage states, and model containment phenomenology and accident progression in containment event trees. A severe earthquake and tsunami hit the eastern coast of Japan in March 2011 and caused significant damages on the reactors in Fukushima Daiichi site. Station blackout (SBO), core damage, containment damage, hydrogen explosion, and intensive radioactivity release, which have been previous analyzed and assumed as postulated accident progression in PRA models, now occurred with various degrees in the multi-units Fukushima Daiichi site. This paper reviews and compares a typical BWR SPAR Level 2 model with the “real” accident progressions and sequences occurred in Fukushima Daiichi Units 1, 2, and 3. It shows that the SPAR Level 2 model is a robust PRA model that could very reasonably describe the accident progression for a real and complicated nuclear accident in the world. On the other hand, the comparison shows that the SPAR model could be enhanced by incorporating some accident characteristics for better representation of severe accident progression.

  16. Large blackouts in North America: Historical trends and policy implications

    International Nuclear Information System (INIS)

    Hines, Paul; Apt, Jay; Talukdar, Sarosh

    2009-01-01

    Using data from the North American Electric Reliability Council (NERC) for 1984-2006, we find several trends. We find that the frequency of large blackouts in the United States has not decreased over time, that there is a statistically significant increase in blackout frequency during peak hours of the day and during late summer and mid-winter months (although non-storm-related risk is nearly constant through the year) and that there is strong statistical support for the previously observed power-law statistical relationship between blackout size and frequency. We do not find that blackout sizes and blackout durations are significantly correlated. These trends hold even after controlling for increasing demand and population and after eliminating small events, for which the data may be skewed by spotty reporting. Trends in blackout occurrences, such as those observed in the North American data, have important implications for those who make investment and policy decisions in the electricity industry. We provide a number of examples that illustrate how these trends can inform benefit-cost analysis calculations. Also, following procedures used in natural disaster planning we use the observed statistical trends to calculate the size of the 100-year blackout, which for North America is 186,000 MW.

  17. Human reliability analysis for venting a BWR Mark I during a severe accident

    International Nuclear Information System (INIS)

    Nelson, W.R.; Blackman, H.S.

    1986-01-01

    A Human Reliability Analysis (HRA) was performed for the operator actions necessary to achieve containment venting for the Peach Bottom Atomic Power Station. This study was funded by the United States Nuclear Regulatory Commission (USNRC) and performed by the Idaho National Engineering Laboratory (INEL). The goal of the analysis was to estimate Human Error Probabilities (HEPs) to determine the likelihood that operators would fail to complete the venting process. The analysis was performed for two generic accident sequences: anticipated transient without scram (ATWS) and station blackout. Two major methods were used to estimate the HEPs: Technique for Human Error rate Prediction (THERP) and Success Likelihood Index Methodology (SLIM). For the ATWS scenarios analyzed, the calculated HEPs ranged from 0.23 to 0.35, depending on the number of vent paths that are required to reduce the containment pressure. It should be noted that the confidence bounds around these HEPs are large, However, even when considering the large confidence range, the failure probabilities are larger than what is typical for normal operator actions. For station blackout, the HEP is 1.0, resulting from the dangerous environmental conditions that are present, assuming that plant management would not deliberately expose personnel to a potentially fatal environment. These results are based on the analysis of draft procedures for containment venting. It is probable that careful revision of the procedures could reduce the human error probabilities

  18. Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code; Verbundprojekt WASA-BOSS: Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt B: Druckwasserreaktor-Stoerfallanalysen unter Verwendung des Severe-Accident-Codes ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-15

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

  19. Recent SCDAP/RELAP5 improvements for BWR severe accident simulations

    International Nuclear Information System (INIS)

    Griffin, F.P.

    1995-01-01

    A new model for the SCDAP/RELAP5 severe accident analysis code that represents the control blade and channel box structures in a boiling water reactor (BWR) has been under development since 1991. This model accounts for oxidation, melting, and relocation of these structures, including the effects of material interactions between B 4 C, stainless steel, and Zircaloy. This paper describes improvements that have been made to the BWR control blade/channel box model during 1994 and 1995. These improvements include new capabilities that represent the relocation of molten material in a more realistic manner and modifications that improve the usability of the code by reducing the frequency of code failures. This paper also describes a SCDAP/RELAP5 assessment calculation for the Browns Ferry Nuclear Plant design based upon a short-term station blackout accident sequence

  20. An electromagnetic method for removing the communication blackout with a space vehicle upon re-entry into the atmosphere

    Science.gov (United States)

    Cheng, Jianjun; Jin, Ke; Kou, Yong; Hu, Ruifeng; Zheng, Xiaojing

    2017-03-01

    When a hypersonic vehicle travels in the Earth and Mars atmosphere, the surface of the vehicle is surrounded by a plasma layer, which is an envelope of ionized air, created from the compression and heat of the atmosphere by the shock wave. The vehicles will lose contact with ground stations known as the reentry communication blackout. Based on the magnetohydrodynamic framework and electromagnetic wave propagation theory, an analytical model is proposed to describe the effect of the effectiveness of electromagnetic mitigation scheme on removing the reentry communication blackout. C and Global Positioning System (GPS) bands, two commonly used radio bands for communication, are taken as the cases to discuss the effectiveness of the electromagnetic field mitigation scheme. The results show that the electron density near the antenna of vehicles can be reduced by the electromagnetic field, and the required external magnetic field strength is far below the one in the magnetic window method. The directions of the external electric field and magnetic field have a significant impact on the effectiveness of the mitigation scheme. Furthermore, the effect of electron collisions on the required applied electromagnetic field is discussed, and the result indicates that electron collisions are a key factor to analyze the electromagnetic mitigation scheme. Finally, the feasible regions of the applied electromagnetic field for eliminating blackout are given. These investigations could have a significant benefit on the design and optimization of electromagnetic mitigation scheme for the blackout problem.

  1. Thermalydraulic processes in the reactor coolant system of a BWR under severe accident conditions

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1990-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermal hydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom atomic power station is presented

  2. Evaluation of a cavity flooding strategy for the prevention of reactor vessel failure in a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Je, Moo Sung; Park, Chang Kyoo [Korea Atomic Energy Research Institute, TaeJon (Korea, Republic of)

    1994-10-01

    As a part of the evaluation of accident management strategies for severe accident prevention or mitigation in a station blackout scenario for YGN 3 and 4, an external vessel cooling strategy for the prevention of reactor vessel failure has been estimated using the MAAP4 computer code. The sensitivity studies have been performed such as actuating timings and the number of spray pumps used. To explore external vessel cooling strategies, containment spray pumps were actuated by varying time spanning core uncovery, core melting and relocation of molten core material. It was shown that flooding of the reactor cavity using the containment spray system may prevent reactor vessel failure but may not prevent the failure of the relocation of molten core material during the station blackout sequence of YGN 3 and 4. Reactor vessel failure can be prevented by external vessel cooling using condensed water from the operation of two containment spray pumps at the time of core melting and using water from the operation of one containment spray pumps at the time of core melting and using water from the operation of one containment spray pump at the time of core uncovery. (Author) 46 refs., 26 figs., 5 tabs.

  3. Optimal operation of hybrid-SITs under a SBO accident

    International Nuclear Information System (INIS)

    Jeon, In Seop; Heo, Sun; Kang, Hyun Gook

    2016-01-01

    Highlights: • Operation strategy of hybrid-SIT (H-SIT) in station blackout (SBO) is developed. • There are five main factors which have to be carefully treated in the development of the operation strategy. • Optimal value of each main factor is investigated analytically and then through thermal-hydraulic analysis using computer code. • The optimum operation strategy is suggested based on the optimal value of the main factors. - Abstract: A hybrid safety injection tank (H-SIT) is designed to enhance the capability of pressurized water reactors against high-pressure accidents which might be caused by the combined accidents accompanied by station blackout (SBO), and is suggested as a useful alternative to electricity-driven motor injection pumps. The main purpose of the H-SIT is to provide coolant to the core so that core safety can be maintained for a longer period. As H-SITs have a limited inventory, their efficient use in cooling down the core is paramount to maximize the available time for long-term cooling component restoration. Therefore, an optimum operation strategy must be developed to support the operators for the most efficient H-SIT use. In this study, the main factors which have to be carefully treated in the development of an operation strategy are first identified. Then the optimal value of each main factor is investigated analytically, a process useful to get the basis of the global optimum points. Based on these analytical optimum points, a thermal-hydraulic analysis using MARS code is performed to get more accurate values and to verify the results of the analytical study. The available time for long-term cooling component restoration is also estimated. Finally, an integrated optimum operation strategy for H-SITs in SBO is suggested.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  5. Investigation of plasma–surface interaction effects on pulsed electrostatic manipulation for reentry blackout alleviation

    International Nuclear Information System (INIS)

    Krishnamoorthy, S; Close, S

    2017-01-01

    The reentry blackout phenomenon affects most spacecraft entering a dense planetary atmosphere from space, due to the presence of a plasma layer that surrounds the spacecraft. This plasma layer is created by ionization of ambient air due to shock and frictional heating, and in some cases is further enhanced due to contamination by ablation products. This layer causes a strong attenuation of incoming and outgoing electromagnetic waves including those used for command and control, communication and telemetry over a period referred to as the ‘blackout period’. The blackout period may last up to several minutes and is a major contributor to the landing error ellipse at best, and a serious safety hazard in the worst case, especially in the context of human spaceflight. In this work, we present a possible method for alleviation of reentry blackout using electronegative DC pulses applied from insulated electrodes on the reentry vehicle’s surface. We study the reentry plasma’s interaction with a DC pulse using a particle-in-cell (PIC) model. Detailed models of plasma–insulator interaction are included in our simulations. The absorption and scattering of ions and electrons at the plasma–dielectric interface are taken into account. Secondary emission from the insulating surface is also considered, and its implications on various design issues is studied. Furthermore, we explore the effect of changing the applied voltage and the impact of surface physics on the creation and stabilization of communication windows. The primary aim of this analysis is to examine the possibility of restoring L- and S-band communication from the spacecraft to a ground station. Our results provide insight into the effect of key design variables on the response of the plasma to the applied voltage pulse. Simulations show the creation of pockets where electron density in the plasma layer is reduced three orders of magnitude or more in the vicinity of the electrodes. These pockets extend to

  6. Organization of action to be taken in the event of a nuclear accident in the nuclear power stations of Electricite de France

    International Nuclear Information System (INIS)

    Martin, J.J.

    1977-01-01

    Depending on the magnitude of the accident in a nuclear power station, the organization of action provides for calling in only the members of the teams on emergency duty if the accident is localized or for putting into effect a structured emergency plan if the accident goes much beyond the confines of the monitored zone. The emergency plan calls for establishing four command posts, the functions of which are specified. Radiological data are centralized at a monitoring command post whose task is to analyse the situation for the management command post. The latter takes decisions which apply to the interior of the site and provides information to the civil defense services at the prefecture level. Any action to be taken outside the site lies within the competence of the prefecture services, which can draw on the resources of the Ministry of Health and the Commissariat a l'energie atomique. Analysis of some incidents in EDF power stations, which are described briefly, has shown the need to facilitate the work of the responsible officials by the preparation of simple charts or schemes which can be used for making an estimate quickly but on the higher side of the potential dose commitment for the public. (author)

  7. Morphological peculiarities of duodenal peptic ulcer and leucocytes functional activity in the persons who were present in the zone of the accident at Chernobyl Atomic Power Station

    International Nuclear Information System (INIS)

    Babak, O.Ya.; Kushnyir, Yi.E.; Bobro, L.M.; Karamishev, D.V.

    1994-01-01

    36 persons with duodenal peptic ulcer (DPU) who were in the zone of the accident at Chernobyl Atomic Power Station (experimental group) and 20 patients who were not exposed to small doses of ionizing radiation were examined to study morphological peculiarities of DPU and blood leucocytes functional activity in the persons who were present in the zone of the accident. The finding have shown that in the persons, exposed to small doses of ionizing radiation, peptic ulcer is often accompanied by erosive changes of gastric and duodenal mucosa. Disturbance of mucus formation in myocytes and secret evacuation from the cells, epithelium large-intestine-type metaplasia, were revealed. Shift of cellular correlation balance in inflammatory infiltrate to the side of monocytes number increase as well as decrease of leucocytes functional activity, manifesting itself by slowing a granulocytes migration to the focus of inflammation, were noted, which is necessary to take into account at administration of effective peptic ulcer therapy in the persons who were in the zone of the accident at Chernobyl Atomic Power Station

  8. Radiation-related anxiety among public health nurses in the Fukushima Prefecture after the accident at the Fukushima Daiichi Nuclear Power Station: a cross-sectional study

    Science.gov (United States)

    Yoshida, Koji; Orita, Makiko; Goto, Aya; Kumagai, Atsushi; Yasui, Kiyotaka; Ohtsuru, Akira; Hayashida, Naomi; Kudo, Takashi; Yamashita, Shunichi; Takamura, Noboru

    2016-01-01

    Objective In Japan, public health nurses (PHNs) play important roles in managing the health of local residents, especially after a disaster. In this study, we assessed radiation anxiety and the stress processing capacity of PHNs in the Fukushima Prefecture in Japan, after the accident at the Fukushima Daiichi Nuclear Power Station (FDNPS). Methods We conducted a questionnaire survey among the PHNs (n=430) in July of 2015 via postal mail. The questions included demographic factors (sex, age and employment position), knowledge about radiation, degree of anxiety about radiation at the time of the FDNPS accident (and at present), by asking them to answer questions about radiation and the Sense of Coherence-13 (SOC-13). We classified the low and high levels of anxiety by asking them to answer questions about radiation, and compared the anxiety-negative (−) group with the anxiety-positive (+) group. Results Of the PHNs, 269 (62.6%) were classified in the anxiety (−) group and 161 (37.4%) were in the anxiety (+) group. When the multivariate logistic regression analysis was conducted, the PHNs at the time of the accident (OR: 2.37, p=0.007), current general anxieties about radiation (OR: 3.56, pChernobyl accident (OR: 1.69, p=0.035) were significantly associated with anxiety after the FDNPS accident. The mean SOC-13 was 43.0±7.7, with no significant difference between the anxiety (−) group and anxiety (+) group (p=0.47). Conclusions Our study suggested that anxiety about radiation was associated with materials and knowledge about radiation in the PHNs in the Fukushima Prefecture 4 years after the FDNPS accident. It is important for PHNs to obtain knowledge and teaching materials about radiation, and radiation education programmes for PHNs must be established in areas that have nuclear facilities. PMID:27798037

  9. Pu isotopes in the western North Pacific Ocean before the accident at Fukushima Dai-ichi Nuclear Power Station

    Science.gov (United States)

    Yamada, M.; Zheng, J.; Aono, T.

    2011-12-01

    Anthropogenic radionuclides such as Pu-239 (half-life: 24100 yr), Pu-240 (half-life: 6560 yr) and Pu-241 (half-life: 14.325 yr) mainly have been released into the environment as the result of atmospheric nuclear weapons testing. In the North Pacific Ocean, two distinct sources of Pu isotopes can be identified; i.e., the global stratospheric fallout and close-in tropospheric fallout from nuclear weapons testing at the Pacific Proving Grounds in the Marshall Islands. The atom ratio of Pu-240/Pu-239 is a powerful fingerprint to identify the sources of Pu in the ocean. The Pu-240/Pu-239 atom ratios in seawater and marine sediment samples collected in the western North Pacific before the accident at Fukushima Dai-ichi Nuclear Power Station will provide useful background data for understanding the process controlling Pu transport and for distinguishing future Pu sources. The atom ratios of Pu-240/Pu-239 in water columns from the Yamato and Tsushima Basins in the Japan Sea were significantly higher than the mean global fallout ratio of 0.18; however, there were no temporal variation of atom ratios during the period from 1984 to 1993 in the Japan Sea. The total Pu-239+240 inventories in the whole water columns were approximately doubled during the period from 1984 to 1993 in the two basins. The atom ratio of Pu-240/Pu-239 in surface water from Sagami Bay, western North Pacific Ocean, was 0.224 and showed no notable variation from the surface to the bottom with the mean atom ratio being 0.234. The atom ratios for the Pacific coast, near the Rokkasho nuclear fuel reprocessing plant, were approximately the same as the 0.224 ratio obtained from Sagami Bay, western North Pacific margin. The atom ratios in the surficial sediments from Sagami Bay ranged from 0.229 to 0.247. The mean atom ratio in the sediment columns in the East China Sea ranged from 0.248 for the Changjiang estuary to 0.268 for the shelf edge. The observed atom ratios were significantly higher than the mean

  10. Report of the US Department of Energy's team analyses of the Chernobyl-4 Atomic Energy Station accident sequence

    International Nuclear Information System (INIS)

    1986-11-01

    In an effort to better understand the Chernobyl-4 accident of April 26, 1986, the US Department of Energy (DOE) formed a team of experts from the National Laboratories including Argonne National Laboratory, Brookhaven National Laboratory, Oak Ridge National Laboratory, and Pacific Northwest Laboratory. The DOE Team provided the analytical support to the US delegation for the August meeting of the International Atomic Energy Agency (IAEA), and to subsequent international meetings. The DOE Team has analyzed the accident in detail, assessed the plausibility and completeness of the information provided by the Soviets, and performed studies relevant to understanding the accident. The results of these studies are presented in this report

  11. Non-radiological consequences to the aquatic biota and fisheries of the Susquehanna River from the 1979 accident at Three Mile Island Nuclear Station

    International Nuclear Information System (INIS)

    Hickey, C.R. Jr.; Samworth, R.B.

    1979-11-01

    The non-radiological consequences to the aquatic biota and fishes of the Susquehanna River from the March 28, 1979 accident at Three Mile Island Nuclear Station were assessed through the post-accident period of July 1979. Thermal and chemical discharges during the period did not exceed required effluent limitations. Several million gallons of treated industrial waste effluents were released into the river which were not of unusual volumes compared with normal operation and were a very small proportion of the seasonally high river flows. The extent and relative location of the effluent plume were defined and the fisheries known to have been under its immediate influence were identified, including rough, forage, and predator/sport fishery species

  12. The NUREG-1150 probabilistic risk assessment for the Grand Gulf nuclear station

    International Nuclear Information System (INIS)

    Brown, T.D.; Breeding, R.J.; Jow, H.N.; Higgins, S.J.; Shiver, A.W.; Helton, J.C.

    1992-01-01

    This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Grand Gulf Nuclear Station performed in support of NUREG-1150. The emphasis is on the 'back-end' analyses, that is, the acident progression, source term, consequence analsyes, and risk results obtained when the results of these analyses are combined with the accident frequency analysis. The offsite risk from internal initiating events was found to be quite low, both with respect to the safety goals and to the other plants analyzed in NUREG-1150. The offsite risk is dominated by short-term station blackout plant damage states. The long-term blackout group and the anticiptated transients without scram (ATWS) group contribute considerably less to risk. Transients in which the power conversion system is unavailable are very minor contributors to risk. The low values for risk can be attributed to low core damage frequency, good emergency response, and plant features that reduce the potential source term. (orig.)

  13. Preliminary accident analysis of Flexblue® underwater reactor

    Directory of Open Access Journals (Sweden)

    Haratyk Geoffrey

    2015-01-01

    Full Text Available Flexblue® is a subsea-based, transportable, small modular reactor delivering 160 MWe. Immersion provides the reactor with an infinite heat sink – the ocean – around the metallic hull. The reference design includes a loop-type PWR with two horizontal steam generators. The safety systems are designed to operate passively; safety functions are fulfilled without operator action and external electrical input. Residual heat is removed through four natural circulation loops: two primary heat exchangers immersed in safety tanks cooled by seawater and two emergency condensers immersed in seawater. In case of a primary piping break, a two-train safety injection system is actuated. Each train includes a core makeup tank, an accumulator and a safety tank at low pressure. To assess the capability of these features to remove residual heat, the reactor and its safety systems have been modelled using thermal-hydraulics code ATHLET with conservative assumptions. The results of simulated transients for three typical PWR accidents are presented: a turbine trip with station blackout, a large break loss of coolant accident and a small break loss of coolant accident. The analyses show that the safety criteria are respected and that the reactor quickly reaches a safe shutdown state without operator action and external power.

  14. Evaluation of High-Pressure RCS Natural Circulations Under Severe Accident Conditions

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Bang, Young Suk; Suh, Nam Duk

    2006-01-01

    Since TMI-2 accident, the occurrence of severe accident natural circulations inside RCS during entire in-vessel core melt progressions before the reactor vessel breach had been emphasized and tried to clarify its thermal-hydraulic characteristics. As one of consolidated outcomes of these efforts, sophisticated models have been presented to explain the effects of a variety of engineering and phenomenological factors involved during severe accident mitigation on the integrity of RCS pressure boundaries, i.e. reactor pressure vessel(RPV), RCS coolant pipe and steam generator tubes. In general, natural circulation occurs due to density differences, which for single phase flow, is typically generated by temperature differences. Three natural circulation flows can be formed during severe accidents: in-vessel, hot leg countercurrent flow and flow through the coolant loops. Each of these flows may be present during high-pressure transients such as station blackout (SBO) and total loss of feedwater (TLOFW). As a part of research works in order to contribute on the completeness of severe accident management guidance (SAMG) in domestic plants by quantitatively assessing the RCS natural circulations on its integrity, this study presents basic approach for this work and some preliminary results of these efforts with development of appropriately detailed RCS model using MELCOR computer code

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

    International Nuclear Information System (INIS)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor

    2013-01-01

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

  16. First retrieval of hourly atmospheric radionuclides just after the Fukushima accident by analyzing filter-tapes of operational air pollution monitoring stations.

    Science.gov (United States)

    Tsuruta, Haruo; Oura, Yasuji; Ebihara, Mitsuru; Ohara, Toshimasa; Nakajima, Teruyuki

    2014-10-22

    No observed data have been found in the Fukushima Prefecture (FP) for the time-series of atmospheric radionuclides concentrations just after the Fukushima Daiichi Nuclear Power Plant (FD1NPP) accident. Accordingly, current estimates of internal radiation doses from inhalation, and atmospheric radionuclide concentrations by atmospheric transport models are highly uncertain. Here, we present a new method for retrieving the hourly atmospheric (137)Cs concentrations by measuring the radioactivity of suspended particulate matter (SPM) collected on filter tapes in SPM monitors which were operated even after the accident. This new dataset focused on the period of March 12-23, 2011 just after the accident, when massive radioactive materials were released from the FD1NPP to the atmosphere. Overall, 40 sites of the more than 400 sites in the air quality monitoring stations in eastern Japan were studied. For the first time, we show the spatio-temporal variation of atmospheric (137)Cs concentrations in the FP and the Tokyo Metropolitan Area (TMA) located more than 170 km southwest of the FD1NPP. The comprehensive dataset revealed how the polluted air masses were transported to the FP and TMA, and can be used to re-evaluate internal exposure, time-series radionuclides release rates, and atmospheric transport models.

  17. First retrieval of hourly atmospheric radionuclides just after the Fukushima accident by analyzing filter-tapes of operational air pollution monitoring stations

    Science.gov (United States)

    Tsuruta, Haruo; Oura, Yasuji; Ebihara, Mitsuru; Ohara, Toshimasa; Nakajima, Teruyuki

    2014-01-01

    No observed data have been found in the Fukushima Prefecture (FP) for the time-series of atmospheric radionuclides concentrations just after the Fukushima Daiichi Nuclear Power Plant (FD1NPP) accident. Accordingly, current estimates of internal radiation doses from inhalation, and atmospheric radionuclide concentrations by atmospheric transport models are highly uncertain. Here, we present a new method for retrieving the hourly atmospheric 137Cs concentrations by measuring the radioactivity of suspended particulate matter (SPM) collected on filter tapes in SPM monitors which were operated even after the accident. This new dataset focused on the period of March 12–23, 2011 just after the accident, when massive radioactive materials were released from the FD1NPP to the atmosphere. Overall, 40 sites of the more than 400 sites in the air quality monitoring stations in eastern Japan were studied. For the first time, we show the spatio-temporal variation of atmospheric 137Cs concentrations in the FP and the Tokyo Metropolitan Area (TMA) located more than 170 km southwest of the FD1NPP. The comprehensive dataset revealed how the polluted air masses were transported to the FP and TMA, and can be used to re-evaluate internal exposure, time-series radionuclides release rates, and atmospheric transport models. PMID:25335435

  18. TEPCO's risk communication activities in Fukushima Prefecture in light of the lessons learned from the Fukushima Daiichi Nuclear Power Station accident

    International Nuclear Information System (INIS)

    Sagasaki, Yoshitoyo; Yamamoto, Takashi

    2015-01-01

    This paper introduces the risk communication activities of the Tokyo Electric Power Company (TEPCO) in Fukushima Prefecture. It analyzed the organizational cause as the background for the Fukushima Daiichi Nuclear Power Station Accident, and concluded that the root cause of the accident is the thought that 'safety has already been secured, and operation rate and the like are important management issues, which incurred the insufficient preparedness for accident.' It has taken six measures as nuclear safety reform plans. One of these is the 'enhancement of risk communication activities.' The nuclear power leader take the initiative to disclose risk under the idea that 'there is no absolute safety (zero risk) in nuclear power,' and promote risk communication for continuously obtaining the understanding of the regional community and society about safety measures, etc. To implement risk communication, 'risk communicators' are installed, and they propose for the management and nuclear leader, about the risk perception and measures associated with public disclosure and its limit, and perform risk communication in accordance with the policy. As the examples of these initiatives, this paper introduces the cases of Fukushima Prefecture, questionnaire study, and evaluations by international organizations. (A.O.)

  19. Simulations of the design basis accident at conditions of power increase and the o transient of MSIV at overpressure conditions of the Laguna Verde Power Station

    International Nuclear Information System (INIS)

    Araiza M, E.; Nunez C, A.

    2001-01-01

    This document presents the analysis of the simulation of the loss of coolant accident at uprate power conditions, that is 2027 MWt (105% of the current rated power of 1931MWt). This power was reached allowing an increase in the turbine steam flow rate without changing the steam dome pressure value at its rated conditions (1020 psiaJ. There are also presented the results of the simulation of the main steam isolation va/ve transient at overpressure conditions 1065 psia and 1067 MWt), for Laguna Verde Nuclear Power Station. Both simulations were performed with the best estimate computer code TRA C BF1. The results obtained in the loss of coolant accident show that the emergency core coolant systems can recover the water level in the core before fuel temperature increases excessively, and that the peak pressure reached in the drywell is always below its design pressure. Therefore it is concluded that the integrity of the containment is not challenged during a loss of coolant accident at uprate power conditions.The analysis of the main steam isolation valve transients at overpressure conditions, and the analysis of the particular cases of the failure of one to six safety relief valves to open, show that the vessel peak pressures are below the design pressure and have no significant effect on vessel integrity. (Author)

  20. Blackout sequence modeling for Atucha-I with MARCH3 code

    International Nuclear Information System (INIS)

    Baron, J.; Bastianelli, B.

    1997-01-01

    The modeling of a blackout sequence in Atucha I nuclear power plant is presented in this paper, as a preliminary phase for a level II probabilistic safety assessment. Such sequence is analyzed with the code MARCH3 from STCP (Source Term Code Package), based on a specific model developed for Atucha, that takes into accounts it peculiarities. The analysis includes all the severe accident phases, from the initial transient (loss of heat sink), loss of coolant through the safety valves, core uncovered, heatup, metal-water reaction, melting and relocation, heatup and failure of the pressure vessel, core-concrete interaction in the reactor cavity, heatup and failure of the containment building (multi-compartmented) due to quasi-static overpressurization. The results obtained permit to visualize the time sequence of these events, as well as provide the basis for source term studies. (author) [es

  1. Containment event analysis for postulated severe accidents: Peach Bottom Atomic Power Station, Unit 2. Draft report for comment

    Energy Technology Data Exchange (ETDEWEB)

    Amos, C N [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States); Griesmeyer, J M [Sandia National Laboratories, Albuquerque, NM (United States); Kolaczkowski, A M [Science Applications International Corporation, Albuquerque, NM (United States)

    1987-05-01

    A study has been performed as part of the Severe Accident Risk Reduction Program (SARRP) to investigate the response of a particular boiling water reactor with a Mark I containment (Peach Bottom Unit 2) to postulated severe accidents. A detailed containment event tree for the Peach Bottom plant has been developed to describe the various possible accident pathways that can lead to radioactive releases from containment. Data and analyses from a large number of NRC and industry-sponsored programs have been reviewed and used as a basis for quantifying the event tree, i.e., determining the likelihood of the pathways at each branch point for a variety of accident sequence initiators. A generalized containment event tree code, called EVNTRE, has been developed to facilitate the quantification. The uncertainty in the results has been examined by performing the quantification three times, using a different set of input each time to represent the variation of opinion in the reactor safety community. In the so-called 'central' estimate, the likelihood of early containment failure (occurring before or within a short time after reactor vessel breach) was found to be significant because of the possible occurrence of the following phenomena that can threaten containment integrity: (1) meltthrough of the drywell shell caused by thermal attack from core debris, and (2) drywell overpressurization caused by rapid depressurization of the reactor vessel in combination with other events such as direct heating. However, uncertainties surrounding these issues could cause the early failure likelihood to be significantly lower than in the central estimate. This work supports NRC's assessment of severe accident risks to be published in NUREG-1150. (author)

  2. Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ji-Han, E-mail: chunjh@kaeri.re.kr; Lim, Sung-Won; Chung, Bub-Dong; Lee, Won-Jae

    2015-08-15

    Highlights: • Thermal conductivity model of the FCM fuel was developed and adopted in the MARS. • Scoping analysis for candidate FCM FAs was performed to select feasible FA. • Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up. • Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. - Abstract: The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates. A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  4. Evaluation of severe accident risks and the potential for risk reduction: Surry Power Station, Unit 1: Draft report for comment

    International Nuclear Information System (INIS)

    Benjamin, A.S.; Boyd, G.J.; Kunsman, D.M.; Murfin, W.B.; Williams, D.C.

    1987-02-01

    The Severe Accident Risk Reduction Program (SARRP) has completed a rebaselining of the risks to the public from a particular pressurized water reactor with a subatmospheric containment (Surry, Unit 1). Emphasis was placed on determining the magnitude and character of the uncertainties, rather than focusing on a point estimate. The risk-reduction potential of a set of proposed safety option backfits was also studied, and their costs and benefits were also evaluated. It was found that the risks from internal events are generally lower than previously evaluated in the Reactor Safety Study (RSS). However, certain unresolved issues (such as direct containment heating) caused the top of the uncertainty band to appear at a level that is comparable with the RSS point estimate. None of the postulated safety options appears to be cost effective for the Surry power plant. This work supports the Nuclear Regulatory Commission's assessment of severe accidents in NUREG-1150

  5. Evaluation of severe accident risks and the potential for risk reduction: Surry Power Station, Unit 1: Draft report for comment

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin, A.S.; Boyd, G.J.; Kunsman, D.M.; Murfin, W.B.; Williams, D.C.

    1987-02-01

    The Severe Accident Risk Reduction Program (SARRP) has completed a rebaselining of the risks to the public from a particular pressurized water reactor with a subatmospheric containment (Surry, Unit 1). Emphasis was placed on determining the magnitude and character of the uncertainties, rather than focusing on a point estimate. The risk-reduction potential of a set of proposed safety option backfits was also studied, and their costs and benefits were also evaluated. It was found that the risks from internal events are generally lower than previously evaluated in the Reactor Safety Study (RSS). However, certain unresolved issues (such as direct containment heating) caused the top of the uncertainty band to appear at a level that is comparable with the RSS point estimate. None of the postulated safety options appears to be cost effective for the Surry power plant. This work supports the Nuclear Regulatory Commission's assessment of severe accidents in NUREG-1150.

  6. The Jules Horowitz reactor: complementary safety evaluation in the light of the Fukushima 1 nuclear power station accident

    International Nuclear Information System (INIS)

    2011-01-01

    This report proposes a complementary safety evaluation of the Jules Horowitz reactor in Cadarache (INB 172), one of the French basic nuclear installations (BNI, in French INB) in the light of the Fukushima accident. This evaluation takes the following risks into account: risks of flooding, earthquake, loss of power supply and loss of cooling, in addition to operational management of accident situations. It presents the main characteristics of the installation, identifies the risks of a cliff effect and the main structures and equipment, evaluates the seismic risk (installation sizing, installation conformity, margin evaluation), evaluates the flooding risk (installation sizing, installation conformity, margin evaluation), briefly examines other extreme natural phenomena (extreme meteorological conditions related to flooding, earthquake or flooding with a higher level than that for which the installation is designed). It analyzes the risk of a loss of power supply and of cooling (loss of external and internal electric sources, loss of the ultimate cooling system). It analyzes the management of severe accidents: crisis management organization, available intervention means, robustness of available means. It discusses the conditions of the use of subcontractors

  7. Mitigating reentry radio blackout by using a traveling magnetic field

    Directory of Open Access Journals (Sweden)

    Hui Zhou

    2017-10-01

    Full Text Available A hypersonic flight or a reentry vehicle is surrounded by a plasma layer that prevents electromagnetic wave transmission, which results in radio blackout. The magnetic-window method is considered a promising means to mitigate reentry communication blackout. However, the real application of this method is limited because of the need for strong magnetic fields. To reduce the required magnetic field strength, a novel method that applies a traveling magnetic field (TMF is proposed in this study. A mathematical model based on magneto-hydrodynamic theory is adopted to analyze the effect of TMF on plasma. The mitigating effects of the TMF on the blackout of typical frequency bands, including L-, S-, and C-bands, are demonstrated. Results indicate that a significant reduction of plasma density occurs in the magnetic-window region by applying a TMF, and the reduction ratio is positively correlated with the velocity of the TMF. The required traveling velocities for eliminating the blackout of the Global Positioning System (GPS and the typical telemetry system are also discussed. Compared with the constant magnetic-window method, the TMF method needs lower magnetic field strength and is easier to realize in the engineering field.

  8. Mitigating reentry radio blackout by using a traveling magnetic field

    Science.gov (United States)

    Zhou, Hui; Li, Xiaoping; Xie, Kai; Liu, Yanming; Yu, Yuanyuan

    2017-10-01

    A hypersonic flight or a reentry vehicle is surrounded by a plasma layer that prevents electromagnetic wave transmission, which results in radio blackout. The magnetic-window method is considered a promising means to mitigate reentry communication blackout. However, the real application of this method is limited because of the need for strong magnetic fields. To reduce the required magnetic field strength, a novel method that applies a traveling magnetic field (TMF) is proposed in this study. A mathematical model based on magneto-hydrodynamic theory is adopted to analyze the effect of TMF on plasma. The mitigating effects of the TMF on the blackout of typical frequency bands, including L-, S-, and C-bands, are demonstrated. Results indicate that a significant reduction of plasma density occurs in the magnetic-window region by applying a TMF, and the reduction ratio is positively correlated with the velocity of the TMF. The required traveling velocities for eliminating the blackout of the Global Positioning System (GPS) and the typical telemetry system are also discussed. Compared with the constant magnetic-window method, the TMF method needs lower magnetic field strength and is easier to realize in the engineering field.

  9. Neutrophil and lymphocyte dose curves in whole-body relatively homogeneous human γ-irradiation (on the basis of the materials of the accident at the Chernobyl Nuclear Power Station)

    International Nuclear Information System (INIS)

    Konchalovskij, M.V.; Baranov, A.E.; Solov'ev, V.Yu.

    1991-01-01

    The experience in a study of regularties of the bone marrow syndrome in persons exposed to rather homogeneous γ-beam irradiation during the accident at the Chernobyl Nuclear Power Station (127 cases) were summed up. Hematological data were processed by computer, and emperic dose curves of neutrophils and lymphocytes were obtained within the range of 0.5-12 Gy by regressive analysis. New data were obtained on the nature of a course of a granulocyte recovery phase at a dose level over 5 Gy. Some features of the time course of lymphocytes in persons exposed to radiation during the accident at the Chernobyl Nuclear Power Station, were considered

  10. Containment response and radiological release for a TMLB' accident sequence in a large dry containment

    International Nuclear Information System (INIS)

    Gasser, R.D.; Bieniarz, P.P.; Tills, J.L.

    1987-01-01

    An analysis has been performed for the Bellefonte Pressurized Water Reactor (PWR) Unit 1 to determine the containment loading and the radiological releases into the environment from a station blackout accident. A number of issues have been addressed in this analysis, which include the effects of direct heating on containment loading and the effects of fission product heating and natural convection on releases from the primary system. The results indicate that direct heating, which involves more than about 50% of the core, may fail the Bellefonte containment, but natural convection in the Reactor Coolant System (RCS) may lead to overheating and failure of the primary system piping before core slump, thus, eliminating or mitigating direct heating. Releases from the primary system are significantly increased before vessel breach, due to natural circulation, and after vessel breach, due to reevolution of retained fission products by fission product heating of RCS structures. (orig.)

  11. A MARS and MIDAS Linked Accident Simulation for Large LOCA in APR1400

    International Nuclear Information System (INIS)

    Choi, Young; Kim, K. R.; Kim, D. H.; Chung, B. D.

    2006-01-01

    A linked calculation utilizing the design-basis code MARS and the severe accident code MIDAS has been accomplished for a station blackout simulation in APR1400. The MARS code was developed by using the RELAP3/MOD3 and COBRA-TF codes, while the MIDAS code is currently under a development process using the MELCOR code. The objectives of this paper are to explain how to identify the MAR-MIDAS linked calculation outlines and the technical problems, including the MARS data transfer method, the MIDAS input generation works and so on. For the performance verification of the MARS-MIDAS linked calculation, the MARS, MIDAS and their linkage system are run independently for the same initiating event, so that their data can be compared with each other after the selection of proper variables

  12. Detailed evaluation of RCS boundary rupture during high-pressure severe accident sequences

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Hong, Seong-Wan

    2011-01-01

    A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR 1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture. (author)

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  14. Insights from Severe Accident Analyses for Verification of VVER SAMG

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A. J.; Rao, R. S.; Gupta, A.; Obaidurrahaman, K., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

    2014-10-15

    The severe accident analyses of simultaneous rupture of all four steam lines (case-a), simultaneous occurrence of LOCA with SBO (case-b) and Station blackout (case-c) were performed with the computer code ASTEC V2r2 for a typical VVER-1000. The results obtained will be used for verification of sever accident provisions and Severe Accident Management Guidelines (SAMG). Auxiliary feed water and emergency core cooling systems are modelled as boundary conditions. The ICARE module is used to simulate the reactor core, which is divided into five radial regions by grouping similarly powered fuel assemblies together. Initially, CESAR module computes thermal hydraulics in primary and secondary circuits. As soon as core uncovery begins, the ICARE module is actuated based on certain parameters, and after this, ICARE module computes the thermal hydraulics in the core, bypass, downcomer and the lower plenum. CESAR handles the remaining components in the primary and secondary loops. CPA module is used to simulate the containment and to predict the thermal-hydraulic and hydrogen behaviour in the containment. The accident sequences were selected in such a way that they cover low/high pressure and slow/fast core damage progression events. Events simulated included slow progression events with high pressure and fast accident progression with low primary pressure. Analysis was also carried out for the case of SBO with the opening of the PORVs when core exit temperature exceeds certain value as part of SAMG. Time step sensitivity study was carried out for LOCA with SBO. In general the trends and magnitude of the parameters are as expected. The key results of the above analyses are presented in this paper. (author)

  15. OSSA. A second generation of severe accident management

    International Nuclear Information System (INIS)

    Sauvage, E.C.; Musoyan, G.; Ducros, V.D.

    2009-01-01

    diagnostic directs the users to a series of system guidelines. The system approach allows a quick and general overview of the plant status. Following the course of an accident scenario evolving to a severe accident, the reader will be taken through the advanced concepts that are developed within the frame of the OSSA. Step by step, in place of the operators or the technical support group, he will experiment the OSSA. The selected severe accident scenario presented is a station blackout during which the late recovery will lead to severe accident. (author)

  16. Population dose and health impact of the accident at the Three Mile Island Nuclear Station. Preliminary estimates for the period March 28, 1979--April 7, 1979

    International Nuclear Information System (INIS)

    Battist, L.; Congel, F.; Buchanan, J.; Peterson, H.

    1979-05-01

    This report contains a preliminary assessment of the radiation dose and potential health impact of the accident on March 28, 1979 at the Three Mile Island Nuclear Station. This assessment was prepared by a task group composed of technical staff members from The Environmental Protection Agency, The Department of Health, Education and Welfare, and The Nuclear Regulatory Commission. The estimated dose that might have been received by an individual is less than 100 mrem. The collective dose received by the 2,164,000 people estimated to live within 50 miles of the reactor site is calculated to be 3,300 person-rem (with a range of 1600 to 5300 person-rem). This corresponds to an average dose of approximately 1.5 mrem. The potential number of fatal cancers that is projected to occur as a result of the accident is less than 1. This potential impact would be undetectable compared to the 325,000 cancer deaths that would normally be expected to occur in a population of 2,164,000. The estimated total health impact, including fatal and non-fatal cancers and genetic effects to all future generations is approximately 2 health effects

  17. The 1st NIRS symposium on reconstruction of early internal dose in the TEPCO Fukushima Daiichi Nuclear Power Station accident. Proceedings

    International Nuclear Information System (INIS)

    Kurihara, Osamu; Akahane, Keiichi; Fukuda, Shigekazu; Miyahara, Nobuyuki; Yonai, Shunsuke

    2012-11-01

    The 2011 earthquake off the Pacific coast of Tohoku district (northern Japan) and the massive tsunamis generated by the earthquake wreaked the most catastrophic damage Japan has experienced in recent centuries. About twenty thousand people were killed or went missing in this natural disaster. This disaster also caused an unprecedented accident at the Fukushima Daiichi Nuclear Power Station operated by Tokyo Electric Power Company. Three reactors in operation were automatically scrammed right after the earthquake; however, these reactors ultimately reached core melt-down by the loss of their cooling systems regardless of extensive efforts for recovery. An enormous amount of radioactive material was released into the environment due to vent operations and a series of explosive events at reactor buildings. The total amount of released radioactive material has been estimated to be about 900 PBq (in 131 I equivalents), which is around one-tenth of that in the Chernobyl accident. Estimation of the dose to the public in affected areas is essential to assess the possible radiological risks in the accident. The National Institute of Radiological Sciences (NIRS) developed a system for estimating early external doses of residents in Fukushima mainly based on information on individual behavior in combination with ambient dose levels measured at various locations after the accident. NIRS has reported the external doses of about 100 thousand residents as of August 2012, revealing that a majority of the external doses are below a few mSv. However, it is difficult to estimate internal doses because of the limited data from individual monitoring or air sampling, especially in the early stage of the accident when radioiodine with a relatively short-half life would have existed as the largest contributor to the thyroid dose. Our current understanding is that there are only about 1,500 human thyroid data from the public and that the main route of intake in the accident was probably

  18. The 1st NIRS symposium on reconstruction of early internal dose in the TEPCO Fukushima Daiichi Nuclear Power Station accident. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Kurihara, Osamu; Akahane, Keiichi; Fukuda, Shigekazu; Miyahara, Nobuyuki; Yonai, Shunsuke [eds.

    2012-11-15

    The 2011 earthquake off the Pacific coast of Tohoku district (northern Japan) and the massive tsunamis generated by the earthquake wreaked the most catastrophic damage Japan has experienced in recent centuries. About twenty thousand people were killed or went missing in this natural disaster. This disaster also caused an unprecedented accident at the Fukushima Daiichi Nuclear Power Station operated by Tokyo Electric Power Company. Three reactors in operation were automatically scrammed right after the earthquake; however, these reactors ultimately reached core melt-down by the loss of their cooling systems regardless of extensive efforts for recovery. An enormous amount of radioactive material was released into the environment due to vent operations and a series of explosive events at reactor buildings. The total amount of released radioactive material has been estimated to be about 900 PBq (in {sup 131}I equivalents), which is around one-tenth of that in the Chernobyl accident. Estimation of the dose to the public in affected areas is essential to assess the possible radiological risks in the accident. The National Institute of Radiological Sciences (NIRS) developed a system for estimating early external doses of residents in Fukushima mainly based on information on individual behavior in combination with ambient dose levels measured at various locations after the accident. NIRS has reported the external doses of about 100 thousand residents as of August 2012, revealing that a majority of the external doses are below a few mSv. However, it is difficult to estimate internal doses because of the limited data from individual monitoring or air sampling, especially in the early stage of the accident when radioiodine with a relatively short-half life would have existed as the largest contributor to the thyroid dose. Our current understanding is that there are only about 1,500 human thyroid data from the public and that the main route of intake in the accident was

  19. Analysis of two different types of hydrogen combustion during severe accidents in a typical pressurized water reactor

    International Nuclear Information System (INIS)

    Ko Yuchih; Lee Min

    2005-01-01

    Hydrogen combustion is an important phenomenon that may occur during severe accidents of Nuclear Power Plants (NPPs). Depending on the specific plant design, the initiating events, and mitigation actions executed, hydrogen combustion may have distinct characteristics and may damage the plant in various degrees. The worst scenario will be the catastrophic failure of containment. In this study two specific types of hydrogen combustion are analyzed to evaluate their impact on the containment integrity. In this paper, Station Blackout (SBO) and Loss of Coolant Accidents (LOCAs) sequences are analyzed using MAAP4 (Modular Accident Analysis Program) code. The former sequence is used to represent hydrogen combustion phenomenon under the condition that the reactor pressure vessel (RPV) breaches at high pressure and the latter sequence represents the phenomenon that RPV fails at low pressure. Two types of hydrogen combustion are observed in the simulation. The Type I hydrogen combustion represents global and instantaneous hydrogen combustion. Large pressure spike is created during the combustion and represents a threat to containment integrity. Type II hydrogen combustion is localized burn and burn continuously over a time period. There is hardly any impact of this type hydrogen burn on the containment pressurization rate. Both types of hydrogen combustion can occur in the severe accidents without any human intervention. From the accident mitigation point of view, operators should try to bring the containment into conditions that favor the Type II hydrogen combustion. (authors)

  20. Hospitalizations for Accidents and Injuries in the U.S. Navy. I. Duty Station Assignment and Duty Status

    Science.gov (United States)

    1985-01-01

    unspecified. . ŗ %.. . . . . . . . . .. . . . Population data, for all male enlisted Navy personnel and for majur duty stations, were compiled from data files...01 aAll male enlisted personnel b nldsonly ship’s company personnel includes only personnel assigned to actual landing forces •’ •,d Significantly...N.S. 102 0.96 2.5. Construction Battalion 28 0.85 N.S. 111 1.05 M.S. Nuclear Cruiser 22 0.67 N.S. 97 0.92 N.S. Pleet Marinie Force 33 1.00 N.S. 71 0.66d

  1. Program for accident and incident management support, AIMS

    International Nuclear Information System (INIS)

    Putra, M.A.

    1993-12-01

    A prototype of an advisory computer program is presented which could be used in monitoring and analyzing an ongoing incident in a nuclear power plant. The advisory computer program, called the Accident and Incident Management Support (AIMS), focuses on processing a set of data that is to be transmitted from a nuclear power plant to a national or regional emergency center during an incident. The AIMS program will assess the reactor conditions by processing the measured plant parameters. The applied model of the power plant contains a level of complexity that is comparable with the simplified plant model that the power plant operator uses. A standardized decay heat function and a steam water property library is used in the integral balance equations for mass and energy. A simulation of the station blackout accident of the Borssele plant is used to test the program. The program predicts successively: (1) the time of dryout of the steam generators, (2) the time of saturation of the primary system, and (3) the onset of core uncovery. The coolant system with the actual water levels will be displayed on the screen. (orig./HP)

  2. [Age dynamics of cerebral blood flow in liquidators of the accident at the Chernobyl Atomic Energy Station].

    Science.gov (United States)

    Krasilenko, E P

    2002-01-01

    An analysis was done of the state of bloodflow in arteries of the carotid and vertebral-and-basilar (VB) trees (as evaluated by USDG) in 104 liquidators of aftermaths of the Chernobyl accident (ChAALs) and in 73 controls intact to the radiation factor. ChAALs demonstrated a higher tempo of an age-associated decrease in linear systolic velocity of the bloodflow (LBFSV) and significantly lower (compared to controls) LBFSV levels in the studied arteries in different age periods, which fact suggests to us a quickened and heterochronous aging of cerebral vessels in ChAALs. In ChAALs, signs of stenoses in the carotid tree have been found to be present two decades earlier than in controls, with associated affections of arteries of both trees detectable at ages 40 to 69 years, which fact suggests apparent diffuse changes in the cerebral vascular bed and narrowing of the cerebral bloodflow safety range in ChAALs. The identified structural changes in the great arteries of the brain should be regarded as a risk factor for formation of cerebrovascular pathology in ChAALs.

  3. Emergency makeup of nuclear steam generators in blackout conditions

    International Nuclear Information System (INIS)

    Korolev, A.V.; Derevyanko, O.V.

    2014-01-01

    The paper describes an original solution for using steam energy to organize makeup of NPP steam generators in blackout conditions. The proposed solution combines a disk friction turbine and an axial turbine in a single housing to provide a high overall technical effect enabling the replenishment of nuclear steam generators with steam using the pump turbine drive assembly. The application of the design is analyzed and its efficiency and feasibility are shown

  4. The accident at TEPCO's Fukushima-Daiichi Nuclear Power Station: What went wrong and what lessons are universal?

    Energy Technology Data Exchange (ETDEWEB)

    Omoto, Akira, E-mail: akira.omoto@mac.com

    2013-12-11

    After a short summary of the nuclear accident at the Fukushima Daiichi Nuclear Power Station, this paper discusses “what went wrong” by illustrating the problems of the specific layers of defense-in-depth (basic strategy for assuring nuclear safety) and “what lessons are universal.” Breaches in the multiple layers of defense were particularly significant in respective protection (a) against natural disasters (first layer of defense) as well as (b) against severe conditions, specifically in this case, a complete loss of AC/DC power and isolation from the primary heat sink (fourth layer of defense). Confusion in crisis management by the government and insufficient implementation of offsite emergency plans revealed problems in the fifth layer of defense. By taking into consideration managerial and safety culture that might have relevance to this accident, in the author's view, universal lessons are as follows: a)Resilience: the need to enhance organizational capabilities to respond, monitor, anticipate, and learn in changing conditions, especially to prepare for the unexpected. This includes increasing distance to cliff edge by knowing where it exists and how to increase safety margin. b)Responsibility: the operator is primarily responsible for safety, and the government is responsible for protecting public health and environment. For both, their right decisions are supported by competence, knowledge, and an understanding of the technology, as well as humble attitudes toward the limitations of what we know and what we can learn from others. c)Social license to operate: the need to avoid, as much as possible regardless of its probability of occurrence, the reasonably anticipated environmental impact (such as land contamination), as well as to build public confidence/trust and a renewed liability scheme.

  5. Risk communication practice after the Fukushima Daiichi Nuclear Power Station Accident. Interactive explanatory meeting on radiation and its health effects in Ibaraki prefecture

    International Nuclear Information System (INIS)

    Ayame, Junko; Sugiyama, Kenji; Takashita, Hirofumi; Yamamoto, Ryuuichi

    2016-02-01

    Large amounts of radioactive material were released into the environment during the accident at the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company (hereinafter referred to as Fukushima nuclear accident) in March, 2011. The radiation dose rose in a large area of plural prefectures including Fukushima prefecture, and many people had anxiety about radiation and its health effects on their bodies. In such a situation, Japan Atomic Energy Agency (JAEA) received a lot of inquiries and lecture requests about radiation from local residents in Japan. R and D Institutes/Centers of JAEA had explanatory meetings and lectures on radiation and its health effects in response to those requests. Nuclear Fuel Cycle Engineering Laboratories (hereinafter referred to as NCL) of JAEA has held the explanatory meetings in Ibaraki prefecture since May 2011 in order to transmit factual information and reduce the excessive anxiety about radiation risk, based on our experience of risk communication practice and research activities over 10 years. Applying to our past risk communication process to the explanatory meetings, we built a process of interactivity between participants and our staff for the meetings. We incorporated the participants' needs into the meetings, and, as far as possible, we had interactive two-way communication so that the meetings were not one-way and persuasive but promote mutual understanding. According to the opinions and the results of questionnaire survey that were received from the participants, it became evident that the interactive explanatory meetings were effective in reducing participants' anxiety. This report explains the risk communication process for carrying out the explanatory meeting, and shows the activities of the meetings, questions and opinions from the participants, and questionnaire results that NCL implemented. (author)

  6. Areva T and D market opportunities after the US and EU Blackouts

    Energy Technology Data Exchange (ETDEWEB)

    Hakansson, K

    2004-02-01

    This document presents the events on the transmission systems during August 2003 in Usa and in September 2003 in Italy. The author analyzes the causes of the blackouts (small margins in transmission system, not adequate control, weaknesses in interconnections between regions), the market opportunity arising out of the blackouts, the economic regulatory and environmental structure/issues today and developments, the scenario for Areva after the blackout (the market size today and in the future) and Areva strength in relation to blackout. (A.L.B.) opportunities.

  7. Areva T and D market opportunities after the US and EU Blackouts

    International Nuclear Information System (INIS)

    Hakansson, K.

    2004-02-01

    This document presents the events on the transmission systems during August 2003 in Usa and in September 2003 in Italy. The author analyzes the causes of the blackouts (small margins in transmission system, not adequate control, weaknesses in interconnections between regions), the market opportunity arising out of the blackouts, the economic regulatory and environmental structure/issues today and developments, the scenario for Areva after the blackout (the market size today and in the future) and Areva strength in relation to blackout. (A.L.B.) opportunities

  8. Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

    Energy Technology Data Exchange (ETDEWEB)

    Ott, Larry J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate

  9. A study on the loss-of-coolant accidents associated with the lung-men nuclear power station

    International Nuclear Information System (INIS)

    Teng, J.T.; Hsu, C.T.; Wang, T.Q.; Chen, Y.H.; Wang, L.C.; Chung, N.M.; Yuann, R.Y.

    2001-01-01

    This study was intended to evaluate the behavior of the nuclear core of the Lung-Men Nuclear Power Station (LMNPS) under postulated LOCA conditions. The LMNPS construction is now in suspense by the Ministry of Economic Affairs, the Republic of China. The assumptions used in this study were in compliance with the requirements specified in 10CFR50.46 and Appendix K. The methodology used was primarily RELAP5YA, which was a modification to the RELAP5/MOD1 Cycle 18. In the paper, features of the thermo-fluids, neutronics, flow systems, trips, and breaks are discussed. Their assumptions and the resulting implications to the outcome of the analyses are emphasized. Also typical sequences of events, the reactor pressure vessel (RPV) pressure, temperature and water inventory transients, and the ultimate core heat-ups for a number of break sizes, ranging from small- to large-break LOCAs, are delineated. The results of this study indicated that for all cases studied, the peak cladding temperature (PCT) was 699.1 Celsius degrees (1290.4 F). This PCT was much lower than the upper temperature limit of 1204.4 Celsius degrees (2200 F) specified in the acceptance criterion of 10CFR50.46. It is to be noted that for all cases studied, the highest PCTs obtained occurred at 4 s after the initiation of the LOCAs. The reason for the occurrence of these PCTs was the internal pump trip, allowing the pump to coast down and the pump to reverse. The next PCTs, resulted from the LOCA, were observed to occur only for the LOCA cases with feedwater line breaks. It did not happen for the cases with steam-line breaks. (authors)

  10. Programmatic environmental impact statement related to decontamination and disposal of radioactive wastes resulting from March 28, 1979 accident, Three Mile Island Nuclear Station, Unit 2 (Docket No. 50-320): Draft

    International Nuclear Information System (INIS)

    1986-12-01

    In accordance with the National Environmental Policy Act and the Commission's implementing regulations and its April 27, 1981 Statement of Policy, the Programmatic Environmental Impact Statement related to decontamination and disposal of radioactive wastes resulting from March 28, 1979, accident Three Mile Island Nuclear Station, Unit 2 NUREG-0683 (PEIS) is being supplemented. This draft supplement updates the environmental evaluation of accident-generated water disposal alternatives published in the PEIS, utilizing more complete and current information. Also, the draft supplement includes a specific environmental evaluation of the licensee's recently submitted proposal for water disposition

  11. 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)

  12. Statistical analysis of the early phase of SBO accident for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Yaroslav, E-mail: y.kozmenkov@hzdr.de; Jobst, Matthias, E-mail: m.jobst@hzdr.de; Kliem, Soeren, E-mail: s.kliem@hzdr.de; Schaefer, Frank, E-mail: f.schaefer@hzdr.de; Wilhelm, Polina, E-mail: p.wilhelm@hzdr.de

    2017-04-01

    Highlights: • Best estimate model of generic German PWR is used in ATHLET-CD simulations. • Uncertainty and sensitivity analysis of the early phase of SBO accident is presented. • Prediction intervals for occurrence of main events are evaluated. - Abstract: A statistical approach is used to analyse the early phase of station blackout accident for generic German PWR with the best estimate system code ATHLET-CD as a computation tool. The analysis is mainly focused on the timescale uncertainties of the accident events which can be detected at the plant. The developed input deck allows variations of all input uncertainty parameters relevant to the case. The list of identified and quantified input uncertainties includes 30 parameters related to the simulated physical phenomena/processes. Time uncertainties of main events as well as the major contributors to these uncertainties are defined. The uncertainty in decay heat has the highest contribution to the uncertainties of the analysed events. A linear regression analysis is used for predicting times of future events from detected times of occurred/past events. An accuracy of event predictions is estimated and verified. The presented statistical approach could be helpful for assessing and improving existing or elaborating additional emergency operating procedures aimed to prevent severe damage of reactor core.

  13. Depressurization as an accident management strategy to minimize the consequences of direct containment heating

    International Nuclear Information System (INIS)

    Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A.

    1990-10-01

    Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of ∼922 K (1200 degree F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs

  14. Depressurization as an accident management strategy to minimize the consequences of direct containment heating

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P.; Dobbe, C.A. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-10-01

    Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of {approximately}922 K (1200{degree}F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs.

  15. 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.)

  16. Basic concept of the nuclear emergency preparedness and response in Japan after the accident of the Fukushima Dai-ichi Nuclear Power Station. The plain explanation for regional officials and emergency workers

    International Nuclear Information System (INIS)

    Sato, Sohei; Yamamoto, Kazuya

    2013-07-01

    After the accident of TEPCO's Fukushima Dai-ichi Nuclear Power Station occurred on March 11, 2011, actions for controlling the accident and protective actions for the residents like evacuation were taken. In parallel with this, it has been developed to reform the nuclear regulatory systems and the emergency preparedness and response systems in Japan. Especially the Nuclear Regulation Authority's Nuclear Emergency Preparedness and Response Guidelines were adopted with the introducing the basic concepts and the criteria on the basis of the IAEA's safety standards and differed greatly from the prior guidelines. Thus the arrangement of emergency response systems, resources and the operational procedures will be developed complying with according to the guidelines in municipalities around the nuclear power station sites. This work attempts to provide a plain explanation as possible for the regional officials and emergency workers about the basic concepts of the new guidelines. (author)

  17. The 4 november 2006 blackout: an outage of 'technical' democracy?

    International Nuclear Information System (INIS)

    Leteurtrois, J.P.

    2007-01-01

    With its power plants and exportation of electricity to neighboring lands, France imagined that it was sheltered from blackouts. But in the autumn of 2006, five million French households were deprived of electricity due to an error by a German operator. What to think of this? The internationalization of the electricity market, though useful to consumers, should not mean deregulation or the relinquishment of rules and regulations to power companies. Supervision of the grid must be reinforced on behalf of all European users of electricity. (author)

  18. Programmatic Environmental Impact Statement related to decontamination and disposal of radioactive wastes resulting from March 28, 1979 accident at Three Mile Island Nuclear Station, Unit 2 (Docket No. 50-320): Final report

    International Nuclear Information System (INIS)

    1987-06-01

    In accordance with the National Environmental Policy Act, the Commission's implementing regulations, and the Commission's April 27, 1981 Statement of Policy, the Programmatic Environmental Impact Statement related to decontamination and disposal of radioactive wastes resulting from March 28, 1979, accident Three Mile Island Nuclear Station, Unit 2 NUREG-0683 (PEIS) is being supplemented. This supplement updates the environmental evaluation of accident-generated water disposal alternatives published in the PEIS, utilizing more complete and current information. Also, the supplement includes a specific environmental evaluation of the licensee's proposal for water disposition. Although no clearly preferable water disposal alternative was identified, the supplement concluded that a number of alternatives could be implemented without significant environmental impact. The NRC staff has concluded that the licensee's proposed disposal of the accident-generated water by evaporation will not significantly affect the quality of the human environment. Further, any impacts from the disposal program are outweighed by its benefits

  19. Blackouts as a Moderator of Young Adult Veteran Response to Personalized Normative Feedback for Heavy Drinking.

    Science.gov (United States)

    Miller, Mary Beth; DiBello, Angelo M; Carey, Kate B; Pedersen, Eric R

    2018-06-01

    Blackouts-or periods of alcohol-induced amnesia for all or part of a drinking event-have been identified as independent predictors of alcohol-related harm that may be used to identify individuals who would benefit from intervention. However, little is known about the prevalence and impact of blackouts among Veterans. This study examined blackouts as a moderator of young adult veteran response to a brief, online personalized normative feedback (PNF) intervention for heavy drinking. Veterans scoring ≥3/4 (women/men) on the Alcohol Use Disorders Identification Test completed a baseline and 1-month assessment as part of a larger intervention trial (N = 571; 83% male; age M = 28.9, SD = 3.3). Participants were randomized to alcohol PNF (n = 285) or a video game attention control (n = 286). Hierarchical regression was used to examine the interaction between intervention condition and blackouts on alcohol-related outcomes at 1-month follow-up. At baseline, 26% of participants reported loss of memory for drinking events in the past 30 days. The interaction between condition and blackouts was significant, such that PNF participants who had experienced blackouts at baseline reported greater decreases in drinking quantity at 1 month than those who had not, and only PNF participants who had experienced baseline blackouts reported a decrease in alcohol problems at follow-up. PNF appears to be particularly effective for individuals who have experienced alcohol-induced blackout, perhaps because blackouts prime them for feedback on their alcohol use. While other negative consequences may also prime individuals for behavior change, blackouts are posited as a particularly useful screening tool because they are prevalent among young adults, have a strong association with alcohol-related harm, and are assessed in widely used clinical measures. Copyright © 2018 by the Research Society on Alcoholism.

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  1. Simulation and dose analysis of a hypothetical accident in Sanmen nuclear power plant

    International Nuclear Information System (INIS)

    Zhu, Yangmo; Guo, Jianghua; Nie, Chu; Zhou, Youhua

    2014-01-01

    Highlights: • Atmospheric dispersion following a hypothetical accident in Sanmen NPP is simulated. • Japan, North Korea and Russia are slightly influenced in this accident. • In Taiwan and South Korea, population on 100% and 35% of the land should be given information about reducing dose. • In mainland China, about 284 thousand people are likely to get cancer. - Abstract: In November 2013, an AP1000 nuclear power plant (NPP) will be put into commercial operation. An atmospheric dispersion of radionuclides during a severe hypothetical accident in Sanmen NPP, Zhejiang province, China, is simulated with a Lagrangian particle dispersion model FLEXPART. The accident assumes that a station blackout (SBO) accident occurred on August 25, 2011, 55% core was damaged and 49 radionuclides were released into the atmosphere. Our simulation indicates that, during this dispersion, the radioactive plume will cover the mainland China, Taiwan, Japan, North Korea, South Korea and Russia. The radiation dose levels in Japan, North Korea and Russia are the lightest, usually less than 1 mSv. The influenced areas in these countries are 9901 km 2 , 31,736 km 2 and 2,97,524 km 2 , respectively; dose levels in Taiwan and South Korea are moderate, no more than 20 mSv. Information about reducing dose should be given to the public. Total influenced areas in these two countries are 3621 km 2 and 42,370 km 2 , which take up 100% of the land in Taiwan and 35% of the land in South Korea; the worst situation happens in mainland China. The total influenced area is 3 × 106 km 2 and 1,40,000 km 2 in this area has a dose level higher than 20 mSv. Measurement must be taken to reduce the dose. More than 284 thousand residents will face the risk of developing cancer. Furthermore, 96% of this population is mainly concentrated in Zhejiang province, where Sanmen NPP locates

  2. Emergency operating procedures improvement based on the lesson learned from the Fukushima Daiichi accident

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Wen-Hsiung, E-mail: whwu1127@aec.gov.tw [Atomic Energy Council, 2F., No. 80, Sec.1, Chenggong Rd., Yonghe Dist., New Taipei City 234, Taiwan (China); Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Sec. 2, Guangfu Rd., Hsinchu City 300, Taiwan (China); Liao, Lih-Yih, E-mail: lyliao@iner.gov.tw [Institute of Nuclear Energy Research, Atomic Energy Council, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 325, Taiwan (China)

    2016-12-01

    Highlights: • Discuss the problem of EOPs at the time of Fukushima accident to deal with the prolonged SBO. • Elaborate the potential risk accompanied with the emergency depressurization in the SBO. • Describe a special guideline to cope with Fukushima-like accidents and provide its technical basis. • Point out that Fukushima accident might have been prevented if improved EOPs had been used. • Propose key points and suggestions for improving the EOPs. - Abstract: One of the lessons learned from the Fukushima Daiichi accident is the emergency operating procedures (EOPs) have to be improved. The BWR Owners’ Group revised the emergency procedure guidelines and addressed the lesson learned from the Fukushima Daiichi accident in revision 3 in order to avoid loss of turbine-driven makeup water systems during reactor depressurization. However, the improvement deserves much more attention. The existing EOPs at the time of the accident may not be adequate enough for the prolonged station blackout condition, because resources required for performing the EOPs are vastly unavailable or gradually exhausted. The improved EOPs must not only permit early reactor pressure vessel depressurization, but also address the risk accompanied with the emergency depressurization. For this reason, Taiwan Power Company proposed the Ultimate Response Guideline (URG) to cope with Fukushima-like accidents. The main content of the URG is a two-stage depressurization strategy, namely the controlled depressurization and the emergency depressurization. The technical basis of the two-stage depressurization strategy was discussed in this paper. The effectiveness of the URG was verified by using TRAC/RELAP Advanced Computational Engine (TRACE). Besides, the emergency responses performed by Fukushima Daini nuclear power plant (Fukushima Daini NPP) were found to be very similar to the URG. The consequences of Fukushima Daini NPP somehow demonstrate that the URG is effective for Fukushima

  3. Emergency operating procedures improvement based on the lesson learned from the Fukushima Daiichi accident

    International Nuclear Information System (INIS)

    Wu, Wen-Hsiung; Liao, Lih-Yih

    2016-01-01

    Highlights: • Discuss the problem of EOPs at the time of Fukushima accident to deal with the prolonged SBO. • Elaborate the potential risk accompanied with the emergency depressurization in the SBO. • Describe a special guideline to cope with Fukushima-like accidents and provide its technical basis. • Point out that Fukushima accident might have been prevented if improved EOPs had been used. • Propose key points and suggestions for improving the EOPs. - Abstract: One of the lessons learned from the Fukushima Daiichi accident is the emergency operating procedures (EOPs) have to be improved. The BWR Owners’ Group revised the emergency procedure guidelines and addressed the lesson learned from the Fukushima Daiichi accident in revision 3 in order to avoid loss of turbine-driven makeup water systems during reactor depressurization. However, the improvement deserves much more attention. The existing EOPs at the time of the accident may not be adequate enough for the prolonged station blackout condition, because resources required for performing the EOPs are vastly unavailable or gradually exhausted. The improved EOPs must not only permit early reactor pressure vessel depressurization, but also address the risk accompanied with the emergency depressurization. For this reason, Taiwan Power Company proposed the Ultimate Response Guideline (URG) to cope with Fukushima-like accidents. The main content of the URG is a two-stage depressurization strategy, namely the controlled depressurization and the emergency depressurization. The technical basis of the two-stage depressurization strategy was discussed in this paper. The effectiveness of the URG was verified by using TRAC/RELAP Advanced Computational Engine (TRACE). Besides, the emergency responses performed by Fukushima Daini nuclear power plant (Fukushima Daini NPP) were found to be very similar to the URG. The consequences of Fukushima Daini NPP somehow demonstrate that the URG is effective for Fukushima

  4. The countermeasures on Fukushima accident by EU and USA. Report of no need of emergency response according to European intermediate report and US review

    International Nuclear Information System (INIS)

    Mizumachi, Wataru

    2011-01-01

    On September 15, intermediate report of 'stress test' was published from reactor operator of 14 countries introducing nuclear power plants among 27 member states of EU. Based on Fukushima Daiichi accident and with assumption of similar accident occurrence such as (1) earthquake and flood, (2) station blackout and/or loss of final heat sink, (3) accident management for loss of reactor core cooling, loss of cooling function of spent fuel storage pool and loss of integrity of containment vessel, results of computerized simulation were reported. As a result, there existed no nuclear power plant needed for reactor closure. Report would be updated, reviewed by regulatory body, submitted to IAEA by next summer and then final assessment would be performed. If additional improvements were needed in terms of safety margins, additional works would be done during next refueling period. As for Muehlberg reactor in Swiss, intake structure was newly added. In US no 'stress test' was performed like EU and each plant was requested to respond according to NRC's recommendations issued on July 12. As a result, short-term evaluation about Fukushima accident showed US nuclear power plants could operate safely because mitigation measures to reduce possibility of core damage and radioactive material release such as containment vessel venting system had been already taken and decided to reinforce safety measures against outages and others as long-term evaluation. (T. Tanaka)

  5. ESTABLISHMENT OF A SEVERE ACCIDENT MITIGATION STRATEGY FOR AN SBO AT WOLSONG UNIT 1 NUCLEAR POWER PLANT

    Directory of Open Access Journals (Sweden)

    SUNGMIN KIM

    2013-08-01

    Full Text Available During a station blackout (SBO, the initiating event is a loss of Class IV and Class III power, causing the loss of the pumps, used in systems such as the primary heat transporting system (PHTS, moderator cooling, shield cooling, steam generator feed water, and re-circulating cooling water. The reference case of the SBO case does not credit any of these active heat sinks, but only relies on the passive heat sinks, particularly the initial water inventories of the PHTS, moderator, steam generator secondary side, end shields, and reactor vault. The reference analysis is followed by a series of sensitivity cases assuming certain system availabilities, in order to assess their mitigating effects. This paper also establishes the strategies to mitigate SBO accidents. Current studies and strategies use the computer code of the Integrated Severe Accident Analysis Code (ISAAC for Wolsong plants. The analysis results demonstrate that appropriate strategies to mitigate SBO accidents are established and, in addition, the symptoms of the SBO processes are understood.

  6. Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Soo-Yong Park

    2015-10-01

    Full Text Available Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

  7. Evaluation of severe accident risks and the potential for risk reduction: Grand Gulf, Unit 1. Draft for comment, February 1987

    Energy Technology Data Exchange (ETDEWEB)

    Amos, C N [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States); Benjamin, A S; Kunsman, D M; Williams, D C [Sandia National Laboratories, Albuquerque, NM (United States); Boyd, G J; Lewis, S R [Safety and Reliability Optimization Services, Inc., Knoxville, TN (United States); Smith, L N [Science Applications International Corporation, Albuquerque, NM (United States)

    1987-04-01

    The Severe Accident Risk Reduction Program (SARRP) has completed a rebaselining of the risks to the public from a boiling water reactor with a Mark III containment (Grand Gulf, Unit 1). Emphasis was placed on determining the magnitude and character of the uncertainties, rather than focusing on a point estimate. The risk-reduction potential of a set of proposed safety option backfits was also studied, and their costs and benefits were also evaluated. It was found that the risks from internal events are generally low relative to previous studies; for example, most of the uncertainty range is lower than the point estimate of risk for the Peach Bottom plant in the Reactor Safety Study (RSS). However, certain unresolved issues cause the top of the uncertainty band to appear at a level that is comparable with the RSS point estimate. These issues include the diesel generator failure rate, iodine and cesium revolatilization after vessel breach and the possibility of reactor vessel pedestal failure caused by core debris attack. Some of the postulated safety options appear to be potentially cost effective for the Grand Gulf power plant, particularly when onsite accidents costs are included in the evaluation of benefits. Principally these include procedural modifications and relatively inexpensive hardware additions to insure core cooling in the event of a station blackout. This work supports the Nuclear Regulatory Commission's assessment of severe accidents in NUREG-1150. (author)

  8. Psychological distress of residents in Kawauchi village, Fukushima Prefecture after the accident at Fukushima Daiichi Nuclear Power Station: the Fukushima Health Management Survey

    Directory of Open Access Journals (Sweden)

    Koji Yoshida

    2016-08-01

    Full Text Available Background To shed light on the mental health of evacuees after the accident at Fukushima Daiichi Nuclear Power Station (FDNPS, we evaluate the results of the Fukushima Health Management Survey (FHMS of the residents at Kawauchi village in Fukushima, which is located less than 30 km from the FDNPS. Methods We conducted the cross-sectional study within the framework of the FHMS. Exposure values were “anorexia,” “subjective feelings about health,” “feelings about sleep satisfaction,” and “bereavement caused by the disaster,” confounding variables were “age” and “sex,” and outcome variables were “K6 points.” We collected data from the FHMS, and employed the Kessler Psychological Distress Scale (K6 and the posttraumatic stress disorder (PTSD Checklist Stressor-Specific Version (PCL-S to carry out the research. A total of 13 or greater was the cut-off for identifying serious mental illness using the K6 scale. The study subjects included residents (n = 542 of over 30 years of age from Kawauchi village, and data were used from the period of January 1, 2012 to October 31, 2012. Results A total of 474 residents (87.5% scored less than 13 points in the K6 and 68 (12.6% scored 13 points or more. The proportion of elderly residents (over 65 years old among people with K6 score above the cut-off was higher than that among people with K6 score below the cut-off (44.1 vs 31.0%, p < 0.05. In addition, the proportion of residents with anorexia and mental illness among people with K6 score above the cut-off was higher than among people with K6 score below the cut-off (p < 0.001 and p < 0.05, respectively. The amount of residents who scored 44 points or more in the PCL-S among people with K6 score above the cut-off was also considerably higher than among people with K6 score below the cut-off (79.4 vs 12.9%, p < 0.001. Interestingly, the proportion of residents who scored more than among people with K6 score above the cut-off and the

  9. Alcohol-Induced Memory Blackouts as an Indicator of Injury Risk among College Drinkers

    Science.gov (United States)

    Mundt, Marlon P.; Zakletskaia, Larissa I.; Brown, David D.; Fleming, Michael F.

    2011-01-01

    Objective An alcohol-induced memory blackout represents an amnesia to recall events but does not involve a loss of consciousness. Memory blackouts are a common occurrence among college drinkers, but it is not clear if a history of memory blackouts is predictive of future alcohol-related injury above and beyond the risk associated with heavy drinking episodes. This analysis sought to determine if baseline memory blackouts can prospectively identify college students with alcohol-related injury in the next 24 months after controlling for heavy drinking days. Methods Data were analyzed from the College Health Intervention Project Study (CHIPS), a randomized controlled trial of screening and brief physician intervention for problem alcohol use among 796 undergraduate and 158 graduate students at four university sites in the US and one in Canada, conducted from 2004 to 2009. Multivariate analyses used generalized estimating equations (GEE) with the logit link. Results The overall 24-month alcohol-related injury rate was 25.6%, with no significant difference between males and females (p=.51). Alcohol-induced memory blackouts at baseline exhibited a significant dose-response on odds of alcohol-related injury during follow-up, increasing from 1.57 (95% CI: 1.13–2.19) for subjects reporting 1–2 memory blackouts at baseline to 2.64 (95% CI: 1.65–4.21) for students acknowledging 6+ memory blackouts at baseline. The link between memory blackouts and injury was mediated by younger age, prior alcohol-related injury, heavy drinking, and sensation-seeking disposition. Conclusions Memory blackouts are a significant predictor of future alcohol-related injury among college drinkers after adjusting for heavy drinking episodes. PMID:21708813

  10. Ecosystem monitoring of radiocesium redistribution dynamics in a forested catchment in Fukushima after the nuclear power station accident in March 2011

    Science.gov (United States)

    Ohte, N.; Endo, I.; Ohashi, M.; Murakami, M.; Oda, T.; Hotta, N.; Yamanishi, R.; Sugiyama, Y.; Tanoi, K.; Kobayashi, N. I.; Ishii, N.

    2016-12-01

    The accident at the Fukushima Daiichi Nuclear Power Station (F1NPS) in March 2011 emitted 1.2 × 1016 Bq of cesium-137 (137Cs) into the surrounding environment. Radioactive substances, including 137Cs, were deposited onto forested areas in the northeastern region of Japan. To clarify the mechanisms of dispersion and export of 137Cs, within and from a forest ecosystem, we have conducted intensive field observations on hydrological processes and the 137Cs movement and storage in a forested headwater catchment in an area 50 km from F1NNS. Two major pathways of 137Cs transport were focused: 1) through the hydrological processes with dissolved and particulate or colloidal forms, and 2) by dispersion through the food web in the forest-stream ecological continuum. The 137Cs concentrations of stream waters were monitored. Various aquatic and terrestrial organisms were periodically sampled to measure their 137Cs concentrations. The results indicate that the major form of exported 137Cs is via suspended solid in the streamflow. Thus, high flows generated by a storm event accelerated strongly the transportation of 137Cs from the forested catchments. Estimation of 137Cs export from the forested catchments requires precise evaluation of the high water flow during storm events. On the other hand, dissolved form, especially mineral ion form of 137Cs were dominant in througfall and stemflow. Because the biggest pool of 137Cs in the forested ecosystem was the accumulated litters and detritus on the forest floor, 137Cs dispersion through food web was quicker through the detritus food chain than through the grazing food chain. 137Cs concentration of leaves (1.1k Bq/kg), barks (4.4-22.5 kBq/kg) woods (0.1-0.6 kBq/kg) and litters (1.3 kBq/kg) of a dominant deciduous tree (Quercus serrata) suggested that internal cyccling of 137Cs have aready occurred between surface soils including litter layer and plants.

  11. Observing power blackouts from space - A disaster related study

    Science.gov (United States)

    Aubrecht, C.; Elvidge, C. D.; Ziskin, D.; Baugh, K. E.; Tuttle, B.; Erwin, E.; Kerle, N.

    2009-04-01

    In case of emergency disaster managers worldwide require immediate information on affected areas and estimations of the number of affected people. Natural disasters such as earthquakes, hurricanes, tornados, wind and ice storms often involve failures in the electrical power generation system and grid. Near real time identification of power blackouts gives a first impression of the area affected by the event (Elvidge et al. 2007), which can subsequently be linked to population estimations. Power blackouts disrupt societal activities and compound the difficulties associated with search and rescue, clean up, and the provision of food and other supplies following a disastrous event. Locations and spatial extents of power blackouts are key considerations in planning and execution of the primary disaster missions of emergency management organizations. To date only one satellite data source has been used successfully for the detection of power blackouts. Operated by NOAA's National Geophysical Data Center (NGDC) the U.S. Air Force Defense Meteorological Satellite Program (DMSP) Operational Linescan System (OLS) offers a unique capability to observe lights present at the Earth's surface at night. Including a pair of visible and thermal spectral bands and originally designed to detect moonlit clouds, this sensor enables mapping of lights from cities and towns, gas flares and offshore platforms, fires, and heavily lit fishing boats. The low light imaging of the OLS is accomplished using a photomultiplier tube (PMT) which intensifies the visible band signal at night. With 14 orbits collected per day and a 3.000 km swath width, each OLS is capable of collecting a complete set of images of the Earth every 24 hours. NGDC runs the long-term archive for OLS data with the digital version extending back to 1992. OLS data is received by NGDC in near real time (1-2 hours from acquisition) and subscription based services for the near real time data are provided for users all over the

  12. Sinister synergies : how competition for unregulated profit causes blackouts

    International Nuclear Information System (INIS)

    Wilson, J.

    2005-01-01

    This white paper examined the effects of deregulation on electricity system reliability and demonstrated that the pursuit of unregulated profit has increased blackout risk. It was suggested that although deregulation works well in some areas, experts, studies and experience have shown that the deregulation of the electricity system has failed. The make-up of the electricity system was discussed, as well as the importance of the system to security, safety, health and economic well-being. It was suggested that higher costs and the need for greater profits have pushed deregulated power producers to cut costs drastically and to invest where high, short-term returns are more likely, rather than focusing on reasonable long-term returns with reasonable cost savings and reliability. In addition, the complexity of deregulated electricity markets has afforded participants many opportunities to manipulate and cut corners to increase profits. It was suggested that higher costs and the need for higher profits have combined with deregulated market conditions to provide motives and opportunities for a culture of bad behavior. This has cost consumers billions of dollars and resulted in increased blackout risk. It was noted that there have also been significant cut-backs in training, maintenance and rehabilitation, as well as in research. There has been a large increase in the complexity of deregulated systems because of increased numbers of participants, transactions and relationships, which has led to the introduction of new systems without appropriate testing, pilot projects, risk management, gradual implementation and backup procedures. It was concluded that an independent investigation should be carried out, and that a major study is needed to examine deregulated environments. 31 refs

  13. Activities of the interactive public meetings on radiation and its health effect after the TEPCO Fukushima Daiichi Nuclear Power Station accident in the case of Nuclear Fuel Cycle Engineering Laboratories

    International Nuclear Information System (INIS)

    Sugiyama, Kenji; Ayame, Junko; Takashita, Hirofumi; Yamamoto, Ryuuichi

    2016-02-01

    Radioactive materials were widely released into the environment during the TEPCO Fukushima Daiichi Nuclear Power Station accident (hereinafter referred to as the Fukushima nuclear accident), which increased residents' fear of radiation and its health effects on their bodies. Japan Atomic Energy Agency (JAEA) has held public meetings on radiation and its health effects mainly for parents of students in kindergartens, elementary schools, and junior high schools in Fukushima prefecture after the Fukushima nuclear accident. These meetings are held based on our experience of practicing risk communication activities for a decade in JAEA with local residents. Questionnaires were collected after the meetings. The contents of the questionnaire are degree of understanding about the contents of the meetings, anxiety or worry items, opportunity of obtaining radiation and nuclear information before the accident, and some opinions and requests. By analyzing questionnaires, we confirmed that interactive communication is effective in increasing participants' understanding and in decreasing their anxiety. Risk communication study office supported the staff members of the meetings providing information such as participants' questions in the past meetings. To provide information, we made a homepage and held the orientation for the staff members. Questionnaires of the staff members were also collected and analyzed after the public meetings. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

  15. Use of an influence diagram and fuzzy probability for evaluating accident management in a BWR

    International Nuclear Information System (INIS)

    Yu, Donghan; Okrent, D.; Kastenberg, W.E.

    1993-01-01

    This paper develops a new approach for evaluating severe accident management strategies. At first, this approach considers accident management as a decision problem (i.e., ''applying a strategy'' vs. ''do nothing'') and uses influence diagrams. This approach introduces the concept of a ''fuzzy probability'' in the evaluation of an influence diagram. When fuzzy logic is applied, fuzzy probabilities in an influence diagram can be easily propagated to obtain results. In addition, the results obtained provide not only information similar to the classical approach using point-estimate values, but also additional information regarding the impact from imprecise input data. The proposed methodology is applied to the evaluation of the drywell flooding strategy for a long-term station blackout sequence in the Peach Bottom nuclear power plant. The results show that the drywell flooding strategy seems to be beneficial for preventing reactor vessel breach. It is also effective for reducing the probability of the containment failure for both liner melt-through and late overpressurization. Even though there exists uncertainty in the results, ''flooding'' is preferred to ''do nothing'' when evaluated in terms of expected consequences, i.e., early and late fatalities

  16. Correlation of Steam Generator Mixing Parameters for Severe Accident Hot-Leg Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Liao, Yehong; Guentay, Salih [Paul Scherrer Institut, Villigen PSI, CH-5232 (Switzerland)

    2008-07-01

    Steam generator inlet plenum mixing phenomenon with hot-leg counter-current natural circulation during a PWR station blackout severe accident is one of the important processes governing which component will fail first as a result of thermal challenge from the circulating gas with high temperature and pressure. Since steam generator tube failure represents bypass release of fission product from the reactor to environment, study of inlet plenum mixing parameters is important to risk analysis. Probability distribution functions of individual mixing parameter should be obtained from experiments or calculated by analysis. In order to perform sensitivity studies of the synergetic effects of all mixing parameters on the severe accident-induced steam generator tube failure, the distribution and correlation of these mixing parameters must be known to remove undue conservatism in thermal-hydraulic calculations. This paper discusses physical laws governing three mixing parameters in a steady state and setups the correlation among these mixing parameters. The correlation is then applied to obtain the distribution of one of the mixing parameters that has not been given in the previous CFD analysis. Using the distributions and considering the inter-dependence of the three mixing parameters, three sensitivity cases enveloping the mixing parameter uncertainties are recommended for the plant analysis. (authors)

  17. Source Term Analysis for the Nuclear Power Station Goesgen-Daeniken; Quelltermanalysen fuer das Kernkraftwerk Goesgen-Daeniken

    Energy Technology Data Exchange (ETDEWEB)

    Hosemann, J.P.; Megaritis, G.; Guentay, S.; Hirschmann, H.; Luebbesmeyer, D.; Lieber, K.; Jaeckel, B.; Birchley, J.; Duijvestijn, G

    2001-08-01

    Analyses are performed for three accident scenarios postulated to occur in the Goesgen Nuclear Power Plant, a 900 MWe Pressurised Water Reactor of Siemens design. The scenarios investigated comprise a Station Blackout and two separate cases of small break loss-of-coolant accident which lead, respectively, to high, intermediate and low pressure conditions in the reactor system. In each case the accident assumptions are highly pessimistic, so that the sequences span a large range of plant states and a damage phenomena. Thus the plant is evaluated for a diversity of potential safety challenges. A suite of analysis tools are used to examine the reactor coolant system response, the core heat-up, melting, fission product release from the reactor system, the transport and chemical behaviour of those fission products in the containment building, and the release of radioactivity (source term) to the environment. Comparison with reference values used by the licensing authority shows that the use of modern analysis tools and current knowledge can provide substantial reduction in the estimated source term. Of particular interest are insights gained from the analyses which indicate opportunities for operators to reduce or forestall the release. (author)

  18. Research and Activities of OECD/NEA Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Project

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il; Ha, Kwang Soon; Song, JinHo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The severe accident codes were used to analysis the Fukushima Daiichi accident and give valuable information. In addition, the insufficient part of the code could be revised by comparing the calculation result with the measured data. In this circumstance, working plans have been set up to conduct a Benchmark Study of the Accident progression for the Fukushima Daiichi nuclear power plant(BSAF) units 1-3 with the members of the OECD/NEA. The BSAF project was launched in November 2012 with the fifteen organizations of eight countries (France, Germany, Korea, Russia, Spain, Switzerland, the United States, and Japan). The objectives of the project are: - to analyze the Fukushima accident progression. - to raise the understanding of severe accident(SA) phenomena. - to contribute the improvement of methods and models of SA code. - to contribute the status of debris distribution to a future debris removal plan. BSAF phase 2 also has been implemented from April, 2015 and it will be continued to March, 2018. It is more focused on the fission product behavior and source term estimation in phase 2. The phase 2 project implementation period is 3 years from April 2015 to March 2018. The main topic of the BSAF 2 is the fission product behavior in the Fukushima nuclear power plant. In the process of calculation, it is important to know the insufficient models in the code. Furthermore, the model would be applied to the KAERI's severe accident code after making up for the insufficient part.

  19. 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.)

  20. Fukushima nuclear power plant accident was preventable

    Science.gov (United States)

    Kanoglu, Utku; Synolakis, Costas

    2015-04-01

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

  1. Postulated accidents

    International Nuclear Information System (INIS)

    Ullrich, W.

    1980-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  3. Rapid analysis of U isotopes in vegetables using ICP-MS. Application to the emergency U monitoring after the nuclear accident at TEPCO's Fukushima Dai-ichi power station

    International Nuclear Information System (INIS)

    Jian Zheng; Keiko Tagami; Shigeo Uchida

    2012-01-01

    After the nuclear accident at TEPCO's Fukushima Dai-ichi power station in March, hydrogen explosions and reactor building explosion resulted in releases of radionuclides in the environment. Severe radioactive cesium and iodine contaminations have been observed in fallout deposition samples and soils in the East Japan. Radioactive cesium, iodine, uranium, and transuranic radionuclides were set as the monitoring targets in food safety tests. However, so far, only radioactive cesium and iodine were daily measured and reported by the Ministry of Health, Labour, and Welfare. The tedious and time consuming conventional alpha spectrometric method hampered the emergency monitoring U contamination in foods. In this work, we propose a simple and rapid analytical method for 238 U and 235 U/ 238 U isotope ratio analysis in fresh vegetables. This method was applied to the emergency monitoring of radioactive contamination after the nuclear accident at TEPCO's Fukushima Dai-ichi power station. The results showed no U contamination in fresh vegetables collected in Chiba and Ibaraki prefectures in April and May, 2011. (author)

  4. Decontamination and disposal of radioactive wastes resulting from the March 28, 1979 accident, Three-Mile Island Nuclear Station, Unit 2, Pennsylvania-Docket No. 50-320 (final supplement 2 to the final environmental impact statement of March 1981)

    International Nuclear Information System (INIS)

    1987-06-01

    Implementation of actions necessary for decontamination of the facility, defueling of the reactor, and disposition of the radioactive wastes that resulted from the accident on March 28, 1979 at Unit 2 of the Three-Mile Island Nuclear Station in Dauphin County, Pennsylvania are discussed. This second final supplement to the final environmental impact statement, filed in March 1981 on facility decontamination, reevaluates the environmental impacts of accident-generated water disposal alternatives, using more complete and current information. This supplement also includes a specific evaluation of the recently submitted proposal for water disposition. The project would alleviate a radiological hazard that threatens the well-being of the surrounding population and downstream communities. Risks to the general public have been estimated to be very small fractions of the estimated normal incidence of cancer fatalities and genetic disorders. The most significant potential impact is the risk of physical injury associated with transportation accidents. Social impacts during the operation could result in reduced property values, competition between the work force and tourists for temporary housing, and congestion of local traffic arteries. Some psychological stress would experienced by area residents. Economic effects could include increased electricity rates, reduced tourism, and possible resistance to consumption of area goods that consumers might mistakenly think are contaminated

  5. 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.,

  6. The effects of a severe reactor accident at the proposed Sizewell B station upon agriculture and fisheries in the United Kingdom and neighbouring countries

    International Nuclear Information System (INIS)

    Taylor, P.J.

    1984-05-01

    This report is a preliminary assessment of the effects of a degraded core accident at the Sizewell site upon regional, national and international agricultural production and fisheries. Two scales of release are taken, one where the containment works effectively, and one where it is breached within a few hours. These two accidents correspond to the UK-11 and UK-1 releases studied by the NRPB. Hitherto, accident consequence studies have focussed upon numbers of cancers or radiation deaths as a measure of the impact. There have been no detailed studies of the impact of the radioactive fall-out upon agricultural production itself, or upon fisheries if the releases are carried out to sea. This report provides a first assessment of the areas of land contaminated by the two scales of accident under weather conditions representative of the range that occurs normally at the site and for a number of different wind directions. The results show that if the containment remains intact, effects would be confined to the locality, with restrictions extending little further than 30 km. However, in the case of the most severe release, agricultural restrictions would be extensive, depending upon wind direction, over the whole of the British Isles. (author)

  7. Can the complex networks help us in the resolution of the problem of power outages (blackouts) in Brazil?

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Paulo Alexandre de; Souza, Thaianne Lopes de [Universidade Federal de Goias (UFG), Catalao, GO (Brazil)

    2011-07-01

    Full text. What the Brazilian soccer championship, Hollywood actors, the network of the Internet, the spread of viruses and electric distribution network have in common? Until less than two decade ago, the answer would be 'nothing' or 'almost nothing'. However, the answer today to this same question is 'all' or 'almost all'. The answer to these questions and more can be found through a sub-area of statistical physics | called science of complex networks that has been used to approach and study the most diverse natural and non-natural systems, such as systems/social networks, information, technological or biological. In this work we study the distribution network of electric power in Brazil (DEEB), from a perspective of complex networks, where we associate stations and/or substations with a network of vertices and the links between the vertices we associate with the transmission lines. We are doing too a comparative study with the best-known models of complex networks, such as Erdoes-Renyi, Configuration Model and Barabasi-Albert, and then we compare with results obtained in real electrical distribution networks. Based on this information, we do a comparative analysis using the following variables: connectivity distribution, diameter, clustering coefficient, which are frequently used in studies of complex networks. We emphasize that the main objective of this study is to analyze the robustness of the network DEEB, and then propose alternatives for network connectivity, which may contribute to the increase of robustness in maintenance projects and/or expansion of the network, in other words our goal is to make the network to proof the blackouts or improve the endurance the network against the blackouts. For this purpose, we use information from the structural properties of networks, computer modeling and simulation. (author)

  8. Preliminary H{sub 2} Combustion Analysis in the Containment of APR1400 for SBLOCA Accident using a Multi-Dimensional H{sub 2} Analysis System

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Hyung Seok; Kim, Jongtae; Kim, Sang-Baik; Hong, Seong-Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The COM3D analyze an overpressure buildup resulting from a propagation of hydrogen flame along the structure and wall in the containment using the hydrogen distribution result calculated by the GASFLOW. The MAAP evaluates a hydrogen source during a severe accident and transfer it to the GASFLOW. We performed a hydrogen combustion analysis using the multidimensional hydrogen analysis system for a station blackout (SBO) accident under the assumption of 100% metal-water reaction in the reactor vessel. The COM3D results showed that the pressure buildup was about 250 kPa because the flame speed was not increased above 300 m/s and the pressure wave passed through the open spaces in the large containment. To increase the reliability of the COM3D calculation, it is necessary to perform the hydrogen combustion analysis for another accident such as a small break loss of coolant (SBLOCA). KAERI performed a hydrogen combustion analysis for a SBLOCA accident using the multi-dimensional hydrogen analysis system under the assumption of 100% metal-water reaction in the reactor vessel. From the COM3D results, we can know that the pressure buildup was approximately 310 kPa because the flame speed was not increased above 100 m/s owing to the high steam concentration and low oxygen concentration in the hydrogen distributed region of the containment. The predicted maximum overpressure in the SBLOCA accident is similar to that of the COM3D results for the SBO accident. Thus, we found that the maximum overpressure due to the hydrogen combustion in the containment may depend on the amount of hydrogen mass released from the reactor vessel.

  9. Accident at Harrisburg

    International Nuclear Information System (INIS)

    1979-05-01

    The course of events during the accident on 28 March 1979 at Three Mile Island-2 Reactor at Harrisburg, Pennsylvania, is described in detail. The effects (in the environment and within the safety containment) are described. The following points are then discussed: the possibility of a comparable accident occurring in the nuclear power stations in the German Federal Republic; the possibility of any point having been overlooked in the design of nuclear power stations in the Federal Republic; whether previous risk analyses are still valid; and how near the Three Mile Island reactor was to a core meltdown. Some conclusions are drawn. (U.K.)

  10. Use of an influence diagram and fuzzy probability for evaluating accident management in a boiling water reactor

    International Nuclear Information System (INIS)

    Yu, D.; Kastenberg, W.E.; Okrent, D.

    1994-01-01

    A new approach is presented for evaluating the uncertainties inherent in severe accident management strategies. At first, this analysis considers accident management as a decision problem (i.e., applying a strategy compared with do nothing) and uses an influence diagram. To evaluate imprecise node probabilities in the influence diagram, the analysis introduces the concept of a fuzzy probability. When fuzzy logic is applied, fuzzy probabilities are easily propagated to obtain results. In addition, the results obtained provide not only information similar to the classical approach, which uses point-estimate values, but also additional information regarding the impact of using imprecise input data. As an illustrative example, the proposed methodology is applied to the evaluation of the drywell flooding strategy for a long-term station blackout sequence at the Peach Bottom nuclear power plant. The results show that the drywell flooding strategy is beneficial for preventing reactor vessel breach. It is also effective for reducing the probability of containment failure for both liner melt-through and late overpressurization. Even though uncertainty exists in the results, flooding is preferred to do nothing when evaluated in terms of two risk measures: early and late fatalities

  11. MDEP Common Position CP-STC-02. Common Position Addressing Fukushima Daiichi Nuclear Power Accident

    International Nuclear Information System (INIS)

    2016-09-01

    considered are: EPR (1650 MWe pressurized water reactor); AP1000 (1110 MWe pressurized water reactor); APR1400 (1400 MWe pressurized water reactor); VVER (1200 MWe pressurized water reactor); ABWR (boiling water reactor with a power output in the 1350 to 1460 MWe range). The designs were assessed in the following five areas: external hazards reliability of safety functions (addressing station black-out and ultimate heat sinks), accidents with core melt, spent fuel pools, and emergency preparedness in design. Mitigation of extreme external hazards often implies improving the robustness of an NPP. While the MDEP focus in addressing issues related to the Fukushima Daiichi NPP accident has been on safety, it should be noted that improved NPP robustness will also improve the capability of the plant to withstand security related challenges

  12. Highly Reliable Power and Communication System for Essential Instruments under a Severe Accident of NPPs

    International Nuclear Information System (INIS)

    Yoo, S. J.; Choi, B. H.; Jung, S. Y.; Rim, Chun T.

    2013-01-01

    failure. Firstly, the life time of emergency batteries after the station blackout (SBO) was insufficient to operate targeted instruments. Secondly, the absence of proper protection in the containment building for extremely harsh environment after severe accident caused malfunctions. Lastly, since the power or communication cable was cut off, the instruments in the containment building could not transmit information to the outside, or their power source could be lost because most of the equipment in NPPs is wired-system based

  13. Analysis of Hydrogen Concentration Distribution during an SBO Accident for Shin-Ulchin APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jongtae; Hong, Seong Wan [Korea Atomic energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    To prohibit the accumulation of hydrogen, the containment volume is considered to reduce the hydrogen concentration, or hydrogen mitigation devices such as PARs or igniters are installed in the containment. In the case of the Fukushima NPPs, the applied strategy for the hydrogen safety is the use of a containment venting system (CVS). In this way, the hydrogen accumulated in the containment vessel is vented into the environment. One of the causes of the hydrogen explosions occurring in the containment buildings of the Fukushima NPPs is expected to be the failure of the venting system. The hydrogen was therefore easily accumulated in the containment building. It is uncertain what the ignition source for the hydrogen combustion was during the accident. However, it is not too conservative to assume that an ignition source exists at any time and any place in a containment during a core-melt accident. Shin-Ulchin 1 and 2, which are construction plants of an APR 1400, are two of the newest NPPs in Korea. They have many features to enhance the safety margin during a design-based and beyond-design-based accident. One of them is the in-containment refueling water storage tank (IRWST) located inside the containment. It is used as a sink/source for feed-bleed operation. When the core is damaged along an accident progression, the hydrogen generated in the RPV can be released into the IRWST of the APR1400 with steam and water. From a previous study, it was found that a highly concentrated hydrogen/air mixture can be developed if the hydrogen is released into the IRWST. In the case of Shin-Ulchin 1 and 2, the hydrogen mitigation strategy during a high-pressure accident such as a station blackout (SBO) is changed by installing a 3-way valve. When a severe accident management (SAM) for the plant is initiated, the flow path from a pressurizer to the IRWST is changed into a steam-generator (S/G) compartment by turning the 3-wat valve actively (pilot operated). By doing so, it is

  14. Highly Reliable Power and Communication System for Essential Instruments under a Severe Accident of NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, S. J.; Choi, B. H.; Jung, S. Y.; Rim, Chun T. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    failure. Firstly, the life time of emergency batteries after the station blackout (SBO) was insufficient to operate targeted instruments. Secondly, the absence of proper protection in the containment building for extremely harsh environment after severe accident caused malfunctions. Lastly, since the power or communication cable was cut off, the instruments in the containment building could not transmit information to the outside, or their power source could be lost because most of the equipment in NPPs is wired-system based.

  15. MANAGING A PROLONGED STATION BLACKOUT CONDITION IN AHWR BY PASSIVE MEANS

    Directory of Open Access Journals (Sweden)

    MUKESH KUMAR

    2013-10-01

    In view of this, an analysis has been performed for decay heat removal characteristics over several days of an AHWR by ICs. The computer code RELAP5/MOD3.2 was used for this purpose. Results indicate that the ICs can remove the decay heat for more than 10 days without causing any bulk boiling in the GDWP. After that, decay heat can be removed for more than 40 days by boiling off the pool inventory. The pressure inside the containment does not exceed the design pressure even after 10 days by condensation of steam generated from the GDWP on the walls of containment and on the Passive Containment Cooling System (PCCS tubes. If venting is carried out after this period, the decay heat can be removed for more than 50 days without exceeding the design limits.

  16. Tests of Shaft Seal Systems of Circulation Pumps during Station Blackout

    Energy Technology Data Exchange (ETDEWEB)

    Beisiegel, A.; Foppe, F.; Wich, M.

    2014-07-01

    AREVA GmbH operates a unique Thermal-hydraulic plat form in Germany, France and USA. It is recognised as a test body according to ISO 17025. The Deutsche Akkreditierungsstelle GmbH (DAkkS - German Society for Accreditation) has also certified the Thermal-hydraulic platform as an independent inspection body Type C according to ISO 17020. A part of this platform is the Component Laboratory located in Karlstein, Germany which is in operation since more than 50 years. The testing activities cover a wide range as: Critical Heat Flux Tests, Valve Testing and Environmental Qualification for safety related components. Since 2011 the Component Qualification Karlstein extended their testing scope for different types of Shaft Seal Systems. (Author)

  17. Station blackout: Deterministic and probabilistic approach in the field of electrical supply losses by EDF

    International Nuclear Information System (INIS)

    Meslin, T.; Carnino, A.

    1986-01-01

    This example shows the thoroughness of EDF's approach in processing the difficult problems of the loss of electrical power supplies. Efforts are continuing in several directions: continued revision and improvement of operating procedures in the event of loss of electrical power supplies, PWR plant operator training courses devoted to the problems of power supply losses, and continued testing on simulators, and particularly testing under real conditions, including tests lasting several hours made possible by the performance of the new EDF simulators (two-phase code and taking all power losses into account)

  18. A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

    International Nuclear Information System (INIS)

    Mandelli, Diego; Prescott, Steven; Smith, Curtis; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2015-01-01

    In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins

  19. Methodological approaches to diagnosis of immunity state in the persons who took part in elimination of the accident aftereffects at Chernobyl Atomic Power Station

    International Nuclear Information System (INIS)

    Popovs'ka, T.M.; Nikiforova, N.A.; Gubs'kij, V.Yi.; Kalmikova, Yi.Ya.; Sorochan, P.P.; Moskalenko, Yi.P.

    1994-01-01

    Distribution of the types of immune reactions in the patients with cardiovascular diseases and in the healthy persons was shown to be practically identical. Reactions of immune complexes prevailed in the both groups. In obstructive bronchitis the incidence of immune complexes reactions, T-cell mediated and reagan reactions are similar to those in the persons who did not take part in elimination of the accident aftereffects. In contrast tj the other examined patients, reagan and cytotoxic types were not revealed

  20. Utilization of the RELAP4/MOD5/SAS code version in loss of coolant accident in the Angra 1 nuclear power station

    International Nuclear Information System (INIS)

    Sabundjian, G.; Freitas, R.L.

    1991-09-01

    A new version of computer code RELAP4/MOD5 was developed to improve the output. The new version, called RELAP4/MOD5/SAS, prints the main variables in graphical form. In order to check the program, a 36 - volume simulation of the Loss-of-Coolant Accident for Angra - I was performed and the results compared to those of a existing 44 - volume simulation showed a satisfactory agreement with a substantial reduction in computing time. (author)

  1. Impact assessment of the 1977 New York City blackout. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, J. L.; Miles, W. T.

    1978-07-01

    This study was commissioned by the Division of Electric Energy Systems (EES), Department of Energy (DOE) shortly after the July 13, 1977 New York City Blackout. The objectives were two-fold: to assess the availability and collect, where practical, data pertaining to a wide variety of impacts occurring as a result of the blackout; and to broadly define a framework to assess the value of electric power reliability from consideration of the blackout and its effects on individuals, businesses, and institutions. The impacts were complex and included both economic and social costs. In order to systematically classify the most significant of these impacts and provide guidance for data collection, impact classification schemes were developed. Major economic impact categories examined are business; government; utilities (Consolidated Edison); insurance industry; public health services; and other public services. Impacts were classified as either direct or indirect depending upon whether the impact was due to a cessation of electricity or a response to that cessation. The principal economic costs of the blackout are shown. Social impacts, i.e., the changes in social activities and adaptations to these changes were particularly significant in New York due to its unique demographic and geographic characteristics. The looting and arson that accompanied the blackout set aside the NYC experience from other similar power failures. (MCW)

  2. INSTALLATION OF A POST-ACCIDENT CONFINEMENT HIGH-LEVEL RADIATION MONITORING SYSTEM IN THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    Energy Technology Data Exchange (ETDEWEB)

    GREENE,G.A.; GUPPY,J.G.

    1998-09-01

    This is the final report on the INSP project entitled, ``Post-Accident Confinement High-Level Radiation Monitoring System'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.6 (Attachment 1). This project was initiated in February 1993 to assist the Russians in reducing risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Unit 2, through improved accident detection capability, specifically by the installation of a dual train high-level radiation detection system in the confinement of Unit 2 of the Kola NPP. The major technical objective of this project was to provide, install and make operational the necessary hardware inside the confinement of the Kola NPP Unit 2 to provide early and reliable warning of the release of radionuclides from the reactor into the confinement air space as an indication of the occurrence of a severe accident at the plant. In addition, it was intended to provide hands-on experience and training to the Russian plant workers in the installation, operation, calibration and maintenance of the equipment in order that they may use the equipment without continued US assistance as an effective measure to improve reactor safety at the plant.

  3. INSTALLATION OF A POST-ACCIDENT CONFINEMENT HIGH-LEVEL RADIATION MONITORING SYSTEM IN THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    International Nuclear Information System (INIS)

    GREENE, G.A.; GUPPY, J.G.

    1998-01-01

    This is the final report on the INSP project entitled, ''Post-Accident Confinement High-Level Radiation Monitoring System'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.6 (Attachment 1). This project was initiated in February 1993 to assist the Russians in reducing risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Unit 2, through improved accident detection capability, specifically by the installation of a dual train high-level radiation detection system in the confinement of Unit 2 of the Kola NPP. The major technical objective of this project was to provide, install and make operational the necessary hardware inside the confinement of the Kola NPP Unit 2 to provide early and reliable warning of the release of radionuclides from the reactor into the confinement air space as an indication of the occurrence of a severe accident at the plant. In addition, it was intended to provide hands-on experience and training to the Russian plant workers in the installation, operation, calibration and maintenance of the equipment in order that they may use the equipment without continued US assistance as an effective measure to improve reactor safety at the plant

  4. 47 CFR 76.128 - Application of sports blackout rules.

    Science.gov (United States)

    2010-10-01

    ... stations within whose specified zone the community of the community unit or the community within which the... other than major markets as defined in § 76.51, television broadcast stations within whose Grade B... specified zone. [65 FR 68101, Nov. 14, 2000, as amended at 67 FR 68951, Nov. 14, 2002] ...

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

    International Nuclear Information System (INIS)

    Shi, Er-bing; Fang, Cheng-yue; Wang, Chang; Xia, Geng-lei; Zhao, Cui-na

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Sweet, D.W.; Roberts, G.J.

    1994-01-01

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

  7. Occurrence and countermeasures of urban power grid accident

    Science.gov (United States)

    Wei, Wang; Tao, Zhang

    2018-03-01

    With the advance of technology, the development of network communication and the extensive use of power grids, people can get to know power grid accidents around the world through the network timely. Power grid accidents occur frequently. Large-scale power system blackout and casualty accidents caused by electric shock are also fairly commonplace. All of those accidents have seriously endangered the property and personal safety of the country and people, and the development of society and economy is severely affected by power grid accidents. Through the researches on several typical cases of power grid accidents at home and abroad in recent years and taking these accident cases as the research object, this paper will analyze the three major factors that cause power grid accidents at present. At the same time, combining with various factors and impacts caused by power grid accidents, the paper will put forward corresponding solutions and suggestions to prevent the occurrence of the accident and lower the impact of the accident.

  8. Learning from the blackouts. Transmission system security in competitive electricity markets

    Energy Technology Data Exchange (ETDEWEB)

    none

    2005-07-01

    Electricity market reform has fundamentally changed the environment for maintaining reliable and secure power supplies. Growing inter-regional trade has placed new demands on transmission systems, creating a more integrated and dynamic network environment with new real-time challenges for reliable and secure transmission system operation. Despite these fundamental changes, system operating rules and practices remain largely unchanged. The major blackouts of 2003 and 2004 raised searching questions about the appropriateness of these arrangements. Management of system security needs to be transformed to maintain reliable electricity services in this more dynamic operating environment. These challenges raise fundamental issues for policymakers. This publication presents case studies drawn from recent large-scale blackouts in Europe, North America, and Australia. It concludes that a comprehensive, integrated policy response is required to avoid preventable large-scale blackouts in the future.

  9. The Chernobyl reactor accident

    International Nuclear Information System (INIS)

    1986-01-01

    The documentation abstracted contains a complete survey of the broadcasts transmitted by the Russian wire service of the Deutsche Welle radio station between April 28 and Mai 15, 1986 on the occasion of the Chernobyl reactor accident. Access is given to extracts of the remarkable eastern and western echoes on the broadcasts of the Deutsche Welle. (HP) [de

  10. Additional safety assessment of the ATALANTE installation (INB nr 148) with respect to the accident which occurred in the Fukushima-Daiichi power station

    International Nuclear Information System (INIS)

    2012-01-01

    After a presentation of some characteristics of the ATALANTE base nuclear installation (site, activities, buildings, laboratories, materials, status with respect to the additional safety assessment), this document reports the identification of cliff-edge risks and of critical structures and equipment. Then, it addresses the different risks: earthquake (installation sizing, margin assessment, earthquake-induced risks), flooding (different origins, alarm measures), other extreme natural events (lightning, rain falls, extreme earthquake with induced flooding), loss of electric supplies and of cooling systems (separately or combined). The next parts address the organization and means of intervention for severe accident management, and subcontracting practices

  11. Radioactive waste warehousing site installation - INB 56, Additional safety assessment with respect to the accident which occurred in the Fukushima-Daiichi nuclear power station

    International Nuclear Information System (INIS)

    2012-01-01

    After a presentation of some characteristics of the installation located in Cadarache and used for radioactive waste warehousing (description, radioactive and chemical materials, specific risks, present status), this document reports the identification of cliff-edge effect risks and of critical structures and equipment. Then, it addresses the different risks: earthquake (installation sizing, margin assessment), external flooding (installation sizing, margin assessment in relationship with the different flooding origins), other extreme natural events (hail, extreme rainfalls, strong winds, lightning, and earthquake exceeding the design level), loss of electric supplies. The next parts address severe accident management (means and organization for crisis management, robustness of available means), and subcontracting conditions and practices

  12. Programmatic Environmental Impact Statement: related to decontamination and disposal of radioactive wastes resulting from March 28, 1979 accident Three Mile Island Nuclear Station, Unit 2 (Docket No. 50-320). Final supplement dealing with occupational radiation dose. Supplement No. 1

    International Nuclear Information System (INIS)

    1984-10-01

    In accordance with the National Environmental Policy Act, the Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Wastes Resulting from March 28, 1979 Accident Three Mile Island Nuclear Station, Unit 2 has been supplemented. The supplement was required because current information indicates that cleanup may entail substantially more occupational radiation dose to the cleanup work force than originally anticipated. Cleanup was originally estimated to result in from 2000 to 8000 person-rem of occupational radiation dose. Although nearly 2000 person-rem have resulted from cleanup operations performed up to now, current estimates now indicate that between 13,000 and 46,000 person-rem are expected to be required. Alternative cleanup methods considered in the supplement either did not result in appreciable dose savings or were not known to be technically feasible

  13. Final programmatic environmental impact statement related to decontamination and disposal of radioactive wastes resulting from March 28, 1979 accident, Three Mile Island Nuclear Station, Unit 2, Docket No. 50-320

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1981-03-01

    A Final Programmatic Environmental Impact Statement (PEIS) related to the decontamination and disposal of radioactive wastes resulting from the March 28, 1979, accident at Three Mile Island Nuclear Station, Unit 2 (Docket No. 50-320) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission in response to a directive issued by the Commission on November 21, 1979. This statement is an overall study of the activities necessary for decontamination of the facility, defueling, and disposition of the radioactive wastes. The available alternatives considered ranged from implementation of full cleanup to no action other than continuing to maintain the reactor in a safe shutdown condition. Also included are comments of governmental agencies, other organizations, and the general public on the Draft PEIS on this project, and staff responses to these comments. (author)

  14. Programmatic environmental impact statement related to decontamination and disposal of radioactive wastes resulting from March 28, 1979 accident, Three Mile Island Nuclear Station, Unit 2 (Docket No. 50-320). Draft supplement dealing with occupational radiation dose. Supplement No. 1

    International Nuclear Information System (INIS)

    1983-12-01

    In accordance with the National Environmental Policy Act, the Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Waste for the 1979 Accident at Three Mile Island Nuclear Station Unit 2 has been supplemented. The supplement was required because current information indicates that cleanup will entail substantially more occupational radiation dose to the cleanup work force than originally anticipated. Cleanup was originally estimated to result in from 2000 to 8000 person-rem of occupational radiation dose. Although only 1700 person-rem have resulted from cleanup operations performed up to now, current estimates now indicate that between 13,000 and 46,000 person-rem are expected to be required. Alternate cleanup methods considered in the supplement either did not result in appreciable dose savings or were not known to be technically feasible

  15. Accident management

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  16. Airborne cloud tracking measurements during the Three Mile Island nuclear station accident, Middletown, Pennsylvania. Date of survey: March-June 1979

    International Nuclear Information System (INIS)

    Beers, R.H.; Burson, Z.G.; Maguire, T.C.; Shipman, G.R.

    1984-12-01

    The Department of Energy's Aerial Measuring Systems (AMS) flew aerial monitoring missions in response to the Three Mile Island nuclear accident. The first flight was made at 1600 hours EST on 28 March 1979, the first day of the accident. Over 150 flights were made during the period of greatest concern. Measurements included isotope identification, plume direction, and maximum radiation levels inside the plume. The highest exposure rate measured (15 milliroentgen per hour, mR/h) was at 300 feet altitude, 1/4 mile from the plant, at 1030 hours on 30 March 1979. The highest measurement at 1 mile was 4.5 mR/h; this occurred at 1600 hours the same day. The Nuclear Emergency Search Team's (NEST's) communication system and logistic support contributed significantly to the overall DOE coordination of off-site radiation measurements. The AMS also provided real time measurements of radioactive krypton gas during purging operations in June 1980. 13 refs., 11 figs., 1 tab

  17. Unavoidable Accident

    OpenAIRE

    Grady, Mark F.

    2009-01-01

    In negligence law, "unavoidable accident" is the risk that remains when an actor has used due care. The counterpart of unavoidable accident is "negligent harm." Negligence law makes parties immune for unavoidable accident even when they have used less than due care. Courts have developed a number of methods by which they "sort" accidents to unavoidable accident or to negligent harm, holding parties liable only for the latter. These sorting techniques are interesting in their own right and als...

  18. Accidents in perspective

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1989-01-01

    The nuclear industry perspective and the public perspective on big nuclear accidents and leukaemia near nuclear sites are discussed. The industry perspective is that big accidents are so unlikely as to be virtually impossible and that leukaemia is not specifically associated with nuclear installations. Clusters of cancer with statistical significance occur in major cities. The public perspective is coloured by a prejudice and myth: the fear of radiation. The big nuclear accident is seen therefore as much more unacceptable than any other big accident. Risks associated with Sizewell-B nuclear station and the liquid gas depot at Canvey Island are discussed. The facts and figures are presented as tables and graphs. Given conflicting interpretations of the leukaemia problem the public inclines towards the more pessimistic view. (author)

  19. Finite element based stress analysis of BWR internals exposed to accident loads

    Energy Technology Data Exchange (ETDEWEB)

    Altstadt, E; Weiss, F P; Werner, M; Willschuetz, H G

    1998-10-01

    During a hypothetical accident the reactor pressure vessel internals of boiling water reactors can be exposed to considerable loads resulting from temperature gradients and pressure waves. Three dimensional FE models were developed for the core shroud, the upper and the lower core supporting structure, the steam separator pipes and the feed water distributor. The models of core shroud, upper core structure and lower core structure were coupled by means of the substructure technique. All FE models can be used for thermal and for structural mechanical analyses. As an example the FE analysis for the case of a station black-out scenario (loss of power supply for the main circulating pumps) with subsequent emergency core cooling is demonstrated. The transient temperature distributions within the core shroud and within the steam dryer pipes as well were calculated based on the fluid temperatures and the heat transfer coefficients provided by thermo-hydraulic codes. At the maximum temperature gradients in the core shroud, the mechanical stress distribution was computed in a static analysis with the actual temperature field being the load. (orig.)

  20. Evaluation of a severe accident management strategy for boiling water reactors -- Drywell flooding

    International Nuclear Information System (INIS)

    Yu, D.; Xing, L.; Kastenberg, W.E.; Okrent, D.

    1994-01-01

    Flooding of the drywell has been suggested as a strategy to prevent reactor vessel and containment failure in boiling water reactors. To evaluate the candidate strategy, this study considers accident management as a decision problem (''drywell flooding'' versus ''do nothing'') and develops a decision-oriented framework, namely, the influence diagram approach. This analysis chooses the long-term station blackout sequence for a Mark 1 nuclear power plant (Peach Bottom), and an influence diagram with a single decision node is constructed. The node probabilities in the influence diagram are obtained from US Nuclear Regulatory Commission reports or estimated by probabilistic risk assessment methodology. In assessing potential benefits compared with adverse effects, this analysis uses two consequence measures, i.e., early and late fatalities, as decision criteria. The analysis concludes that even though potential adverse effects exist, such as ex-vessel steam explosions and containment isolation failure, the drywell flooding strategy is preferred to ''do nothing'' when evaluated in terms of these consequence measures

  1. AP1000{sup R} severe accident features and post-Fukushima considerations

    Energy Technology Data Exchange (ETDEWEB)

    Scobel, J. H.; Schulz, T. L.; Williams, M. G. [Westinghouse Electric Company, LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at Fukushima without compromising core and containment integrity. The AP1000 plant shuts down the reactor, cools the core, containment and spent fuel pool for more than 3 days using passive systems that do not require AC or DC power or operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. To provide defense-in-depth for design extension conditions, the AP1000 plant has engineered features that mitigate the effects of core damage. Engineered features retain damaged core debris within the reactor vessel as a key feature. Other aspects of the design protect containment integrity during severe accidents, including unique features of the AP1000 design relative to passive containment cooling with water and air, and hydrogen management. (authors)

  2. Conditions for oxygen-deficient combustion during accidents with severe core concrete thermal attack

    International Nuclear Information System (INIS)

    Luangdilok, W.; Elicson, G.T.; Berger, W.E. Jr.

    1993-01-01

    This paper addresses the interactions between MCCI (molten core-concrete interactions)-induced offgas releases, mostly the combustible gases, natural circulation between the cavity and the lower containment based on recent research developments in the area of mixed convection flow (Epstein, et al., 1989; Epstein, 1988; Epstein, 1992) between compartments, and their effects on combustion in PWR containments during prolonged severe accidents. Specifically, large dry PWR containments undergoing severe core-concrete attack during station blackouts where the containment atmosphere is expected to be inerted are objects of this analysis. The purpose of this paper, given the conditions that oxygen can be brought to the cavity, is to demonstrate that consumption of most oxygen present in the containment can be achieved in a reasonable time scale assuming that combustion is not subject to flammability limits due to the high cavity temperatures. The conditions for cavity combustion depend on several factors including good gas flowpaths between the cavity and other containment regions, and combustion processes within the cavity with the hot debris acting as the ignition source

  3. The MCMF installation - INB 53. Additional safety assessment with respect to the accident which occurred in the Fukushima-Daiichi nuclear power station

    International Nuclear Information System (INIS)

    2012-01-01

    After a presentation of some characteristics of the MCMF (fissile materials warehouse) base nuclear installation located in Cadarache (description, activities, radioactive and chemical material inventory, specific risks, and present status), this document reports the identification of cliff-edge effect risks and of critical structures and equipment. Then, it addresses the different risks: earthquake (reference earthquake, installation sizing and compliance, margin assessment for the different structures and equipment), external flooding (installation sizing, compliance, and margin assessment in relationship with the different flooding origins), other extreme natural events (extreme meteorological conditions related to flooding like tempest, rainfalls and so on, earthquake exceeding the design level), and loss of external or internal electric supplies. The next parts address severe accident management (risks related to the industrial environment, means and organization for crisis management, exercises and training, robustness of available means), and subcontracting conditions, modalities and practices

  4. The CEA Cadarache site. Additional safety assessment with respect to the accident which occurred in the Fukushima-Daiichi nuclear power station

    International Nuclear Information System (INIS)

    2012-01-01

    After a presentation of some characteristics of the CEA Cadarache site (internal and external industrial environment, crisis management organization at the CEA level and at the local level), this document reports the identification of structures and equipment concerned by crisis management (site support functions, critical structures and equipment concerned by additional safety assessments). Then, it addresses the different risks: earthquake (sizing of critical structures and equipment, margin assessment), flooding (possible origins, alarm measures), other extreme natural events (extreme meteorological conditions, extreme earthquake with induced flooding, forest fire), and loss of electric supplies and of cooling systems. The last parts address the organization of accident management in situation typically related to additional safety management), and subcontracting conditions and practices

  5. The LECA installation - INB 55. Additional safety assessment with respect to the accident which occurred in the Fukushima-Daiichi nuclear power station

    International Nuclear Information System (INIS)

    2012-01-01

    After a presentation of some characteristics of the Leca base nuclear installation located in Cadarache (description, activities, radioactive and chemical material inventory, specific risks, and present status), this document reports the identification of cliff-edge effect risks and of critical structures and equipment. Then, it addresses the different risks: earthquake (installation sizing, reference earthquake and installation compliance, margin assessment for the different structures and equipment), external flooding (installation sizing, compliance, and margin assessment in relationship with the different flooding origins), other extreme natural events (extreme meteorological conditions, earthquake exceeding the design level) and loss of external or internal electric supplies. The next parts address severe accident management (risks related to the industrial environment, means and organization for crisis management, exercises and training, robustness of available means), and subcontracting conditions and practices

  6. The Pegase installation - INB 22. Additional safety assessment with respect to the accident which occurred in the Fukushima-Daiichi nuclear power station

    International Nuclear Information System (INIS)

    2012-01-01

    After a presentation of some characteristics of the Pegase base nuclear installation located in Cadarache (description and functions, radioactive and chemical materials, specific risks, present status), this document reports the identification of cliff-edge effect risks and of critical structures and equipment. Then, it addresses the different risks: earthquake (installation sizing and compliance, margin assessment), external flooding (installation sizing and compliance, margin assessment in relationship with the different flooding origins), other extreme natural events (hail, extreme rainfalls, strong winds, lightning, and earthquake exceeding the design level), and loss of external or internal electric supplies. The next parts address severe accident management (means and organization for crisis management, robustness of available means), and subcontracting conditions and practices

  7. The Rapsodie installation - INB 25. Additional safety assessment with respect to the accident which occurred in the Fukushima-Daiichi nuclear power station

    International Nuclear Information System (INIS)

    2012-01-01

    After a presentation of some characteristics of the Rapsodie base nuclear installation located in Cadarache (brief description of buildings, radioactive and chemical material inventory, specific risks, and present status), this document reports the identification of cliff-edge effect risks and of critical structures and equipment. Then, it addresses the different risks: earthquake (reference earthquake, installation sizing and compliance issues, margin assessment for the different structures and equipment), external flooding (installation sizing, compliance, and margin assessment in relationship with the different flooding origins), other extreme natural events (extreme meteorological conditions related to flooding, earthquake exceeding the design level), and loss of external or internal electric supplies and of cooling systems. The next parts address severe accident management (risks related to the industrial environment, means and organization for crisis management, exercises and training, robustness of available means), and subcontracting conditions, modalities and practices

  8. The CHICADE installation - INB 156. Additional safety assessment with respect to the accident which occurred in the Fukushima-Daiichi nuclear power station

    International Nuclear Information System (INIS)

    2012-01-01

    After a presentation of some characteristics of the Chicade base nuclear installation located in Cadarache (brief description, radioactive and chemical material inventory, specific risks, and present status), this document reports the identification of cliff-edge effect risks and of critical structures and equipment. Then, it addresses the different risks: earthquake (installation sizing and compliance, margin assessment for the different structures and equipment), external flooding (installation sizing, compliance, and margin assessment in relationship with the different flooding origins), other extreme natural events (extreme meteorological conditions related to flooding like tempest, rainfalls and so on, earthquake exceeding the design level), and loss of external or internal electric supplies. The next parts address severe accident management (risks related to the industrial environment, means and organization for crisis management, exercises and training, robustness of available means), and subcontracting conditions and practices

  9. Acidentes em postos e sistemas retalhistas de combustíveis: subsídios para a vigilância em saúde ambiental Environmental accidents in gas stations and retailling systems

    Directory of Open Access Journals (Sweden)

    Jorge Luiz Nobre Gouveia

    2007-09-01

    Full Text Available Os acidentes em postos e sistemas retalhistas resultam de forma imediata em danos graves à vida de pessoas, na contaminação do solo e da água. O objetivo deste trabalho foi discutir a importância da atuação conjunta dos órgãos ambientais e de saúde e subsidiar a estruturação do Sistema Nacional de Vigilância em Saúde Ambiental. A metodologia consistiu no estudo da literatura nacional e internacional; levantamento dos dados existentes em oito órgãos ambientais estaduais brasileiros sobre atendimentos emergenciais e análise de um caso ocorrido no município de Guarulhos, Estado de São Paulo, em janeiro de 2005. Os resultados revelaram que esses acidentes são importante causa de acidentes ambientais e há grande diferença na atuação dos diversos órgãos ambientais. Também demonstra que a estruturação da vigilância em saúde ambiental deve abordar a interface com outras instituições públicas envolvidas.The accidents in gas stations and retailing systems result in grave damages to the life of people and contamination of the soil and water. The objective is to discuss the importance of the joint action of environmental and health organs and to subsidize the structuring of the National System of Surveillance in Environmental Health. The methodology consisted of the studying of the national and international literature; to compile the existent data on eight environmental organs about accidents, and analysis of one case occurred in Guarulhos, State of São Paulo, in January, 2005. The results revealed that these accidents are an important cause of environmental accidents and that there is a great difference in the actions of the diverse environmental organs and demonstrates that the structure of surveillance in environmental health should consider the interface with other public institutions.

  10. Europe's electrical vulnerability geography : historical interpretations of the 2006 'European blackout'

    NARCIS (Netherlands)

    Vleuten, van der E.B.A.; Lagendijk, V.C.

    2009-01-01

    The so-called "European Blackout" of 4 November 2006 counts as a key example of present day transnational infrastructure vulnerability and an important reference in current debates on transnational electricity infrastructure governance. This is best examplified by the debate itself, where proponents

  11. Control and prediction for blackouts caused by frequency collapse in smart grids.

    Science.gov (United States)

    Wang, Chengwei; Grebogi, Celso; Baptista, Murilo S

    2016-09-01

    The electric power system is one of the cornerstones of modern society. One of its most serious malfunctions is the blackout, a catastrophic event that may disrupt a substantial portion of the system, playing havoc to human life and causing great economic losses. Thus, understanding the mechanisms leading to blackouts and creating a reliable and resilient power grid has been a major issue, attracting the attention of scientists, engineers, and stakeholders. In this paper, we study the blackout problem in power grids by considering a practical phase-oscillator model. This model allows one to simultaneously consider different types of power sources (e.g., traditional AC power plants and renewable power sources connected by DC/AC inverters) and different types of loads (e.g., consumers connected to distribution networks and consumers directly connected to power plants). We propose two new control strategies based on our model, one for traditional power grids and another one for smart grids. The control strategies show the efficient function of the fast-response energy storage systems in preventing and predicting blackouts in smart grids. This work provides innovative ideas which help us to build up a robuster and more economic smart power system.

  12. Acute alcohol effects on narrative recall and contextual memory: an examination of fragmentary blackouts.

    Science.gov (United States)

    Wetherill, Reagan R; Fromme, Kim

    2011-08-01

    The present study examined the effects of alcohol consumption on narrative recall and contextual memory among individuals with and without a history of fragmentary blackouts in an attempt to better understand why some individuals experience alcohol-induced memory impairments whereas others do not, even at comparable blood alcohol concentrations (BACs). Standardized beverage (alcohol and no alcohol) administration procedures and neuropsychological assessments measured narrative recall and context memory performance before and after alcohol consumption in individuals with (n=44) and without (n=44) a history of fragmentary blackouts. Findings indicate that acute alcohol intoxication led to impairments in free recall, but not next-day cued recall. Further, participants showed similar memory performance when sober, but individuals who consumed alcohol and had a positive history of fragmentary blackouts showed greater contextual memory impairments than those who had not previously experienced a fragmentary blackout. Thus, it appears that some individuals may have an inherent vulnerability to alcohol-induced memory impairments due to alcohol's effects on contextual memory processes. Copyright © 2011 Elsevier Ltd. All rights reserved.

  13. Transnational infrastructure vulnerability : the historical shaping of the 2006 European 'blackout'

    NARCIS (Netherlands)

    Vleuten, van der E.B.A.; Lagendijk, V.C.

    2010-01-01

    The "European Blackout" of 4 November 2006 is a key reference in current debates on transnational electricity infrastructure vulnerability and governance. Several commentators have observed that to understand what happened, one must look at history. Our paper answers this call and demonstrates how

  14. The annal of british RPV steel plates for first nuclear power station in Japan (1). Unforseen accidents araised before nuclear power plant open

    International Nuclear Information System (INIS)

    Miyoshi, Shigeru

    2011-01-01

    This article described the author's experiences of reactor vessel steel plates for the first nuclear power station, Tokai-mura reactor. The station is an advanced Calder Hall type. The electrical output is 166 MWe. The reactor vessel was spherical with internal diameter of 189 cm and wall thickness of 83 mm. Material was a fine-grain, aluminum-killed steel. Each part of pressure vessel, bottom cap, belt 1, 2, 3, 4 and top cap, were prefabricated with welding of plates, then lifted into the reactor building and assembled with welding. Steel plates were imported from UK, press formed to spherical segments in Japan and transferred to the site. Ultrasonic testing, magnetic particle testing of groove face (crack detection), sizing of groove and sulfur print tests were performed as an on-site acceptance testing. Inclusions and lamination openings were observed at groove faces due to gas flame cutting. White spot was observed at rupture face of tensile test specimen. At the liquid penetration testing after back gauging of extra seam, a crack-like indication with length of less than 3 mm was observed. Reexamination of groove face by magnetic particles testing showed indications of inclusion cloud or alumina cloud. These would be cracks caused by hydrogen embrittlement. (T. Tanaka)

  15. Assessment of impact of a severe accident at nuclear power plant of Angra dos Reis with release of radionuclides to the atmosphere

    International Nuclear Information System (INIS)

    Aguiar, Andre Silva de

    2015-01-01

    This study had as purpose the assess the impact of a severe accident, and also analyze the dispersion of 131 I in the atmosphere, so that, through concentrating and inhaling dose of the plume, were possible to verify if the results are in accordance with the indicated data by the Plan of Emergency of the CNAAA regarding the Impact Zone and Control. This exercise was performed with the aid of an atmospheric model and a dispersion where to atmospheric modeling we used the data coupling WRF / CALMET and of dispersion, CALPUFF. The suggested accident consists of a Station Blackout at Nuclear Power of Angra (Unit 1), where through the total core involvement, will release 100% of the 131 I to the atmosphere. The value of the total activity in the nucleus to this radionuclide is 7.44 x 1017 Bq, that is relative on the sixth day of burning. This activity will be released through the chimney at a rate in Bq/s in the scenario of 12, 24, 48 and 72 hours of release. Applying the model in the proposed scenario, it is verified that the plume has concentrations of the order of 1020 Bq/m³ and dose of about 108 Sv whose value is beyond of the presented by Eletronuclear in your current emergency plan. (author)

  16. Alcohol-induced blackout as a criminal defense or mitigating factor: an evidence-based review and admissibility as scientific evidence.

    Science.gov (United States)

    Pressman, Mark R; Caudill, David S

    2013-07-01

    Alcohol-related amnesia--alcohol blackout--is a common claim of criminal defendants. The generally held belief is that during an alcohol blackout, other cognitive functioning is severely impaired or absent. The presentation of alcohol blackout as scientific evidence in court requires that the science meets legal reliability standards (Frye, FRE702/Daubert). To determine whether "alcohol blackout" meets these standards, an evidence-based analysis of published scientific studies was conducted. A total of 26 empirical studies were identified including nine in which an alcohol blackout was induced and directly observed. No objective or scientific method to verify the presence of an alcoholic blackout while it is occurring or to confirm its presence retrospectively was identified. Only short-term memory is impaired and other cognitive functions--planning, attention, and social skills--are not impaired. Alcoholic blackouts would not appear to meet standards for scientific evidence and should not be admissible. © 2013 American Academy of Forensic Sciences.

  17. Preventing accidents

    Science.gov (United States)

    2005-08-01

    As the most effective strategy for improving safety is to prevent accidents from occurring at all, the Volpe Center applies a broad range of research techniques and capabilities to determine causes and consequences of accidents and to identify, asses...

  18. Lightning rod ionizing natural ionca - Ionic electrode active trimetallictriac of grounding - Definitive and total solution against 'blackouts' and electrical faults generated by atmospheric charges (lightning)

    Energy Technology Data Exchange (ETDEWEB)

    Cabareda, Luis

    2010-09-15

    The Natural Ionizing System of Electrical Protection conformed by: Lightning Rod Ionizing Natural Ionca and Ionic Electrode Active Trimetallic Triac of Grounding offers Total Protection, Maximum Security and Zero Risk to Clinics, Hospitals, Integral Diagnostic Center, avoiding ''the burning'' of Electronics Cards; Refineries, Tanks and Stations of Fuel Provision; Electrical Substations, Towers and Transmission Lines with transformer protection, motors, elevators, A/C, mechanicals stairs, portable and cooling equipment, electrical plants, others. This New High Technology is the solution to the paradigm of Benjamin Franklin and it's the mechanism to end the 'Blackouts' that produces so many damages and losses throughout the world.

  19. Analysis of a Natural Circulation in the Reactor Coolant System Following a High Pressure Severe Accident at APR1400

    International Nuclear Information System (INIS)

    Kim, Han Chul; Cho, Yong Jin; Park, Jae Hong; Cho, Song Won

    2011-01-01

    Under a high temperature and pressure condition during a severe accident, hot leg pipes or steam generator tubes could fail due to creep rupture following natural circulation in the Reactor Coolant System (RCS) unless depressurization of the system is performed at a proper time. Natural circulation in the RCS can be a multi-dimensional circulation in the reactor vessel, a partial loop circulation of two-phase flow from the core up to steam generators (SGs), or circulation in the total loop. It can delay the reactor vessel failure time by removing heat from the reactor core. This natural phenomenon can be hardly simulated with a single flow path model for the hot spots of the RCS, since it cannot deal with the counter-current flow. Thus it may estimate accident progression faster than reality, which may cause troubles for optimized implementation of severe accident management strategies. An earlier damage in the RCS other than the reactor pressure vessel may make subsequent behaviors of hydrogen or fission products in the containment quite different from the single reactor vessel failure. Therefore, a RCS model which treats natural circulation is needed to evaluate the RCS response and the safety depressurization strategy in a best-estimate way. The aim of this study is to develop a detailed model which allows natural circulation between the reactor vessel and steam generators through hot legs, based on the existing APR1400 RCS model. The station blackout sequence was selected to be the representative high-pressure scenario. Sensitivity study on the effect of node configuration of the upper plenum and addition of cross flow paths from the upper plenum to the hot legs were carried out. This model is described herein and representative calculation results are presented

  20. 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

  1. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of an atmospheric dispersion model with an improved deposition scheme and oceanic dispersion model

    Energy Technology Data Exchange (ETDEWEB)

    Katata, G.; Chino, M.; Kobayashi, T. [Japan Atomic Energy Agency (JAEA), Ibaraki (Japan); and others

    2015-07-01

    Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Daiichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate the detailed atmospheric releases during the accident using a reverse estimation method which calculates the release rates of radionuclides by comparing measurements of air concentration of a radionuclide or its dose rate in the environment with the ones calculated by atmospheric and oceanic transport, dispersion and deposition models. The atmospheric and oceanic models used are WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information) and SEA-GEARN-FDM (Finite difference oceanic dispersion model), both developed by the authors. A sophisticated deposition scheme, which deals with dry and fog-water depositions, cloud condensation nuclei (CCN) activation, and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I{sub 2} and CH{sub 3}I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The results revealed that the major releases of radionuclides due to the FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, midnight of 14 March when the SRV (safety relief valve) was opened three times at Unit 2, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of release rates. The simulation by WSPEEDI-II using the new source term reproduced the local and regional patterns of

  2. Use of knowledge and experience gained from the Fukushima Daiichi Nuclear Power Station accident to establish the technical basis for strategic off-site response

    International Nuclear Information System (INIS)

    Miyahara, Kaname; Saito, Kimiaki; Iijima, Kazuki; McKinley, Ian; Hardie, Susan

    2015-03-01

    This report provides a concise overview of knowledge and experience gained from the activities for environmental remediation after the Fukushima Daiichi (1F) accident. It is specifically tailored for international use, to establish or refine the technical basis for strategic, off-site response to nuclear incidents. It reflects JAEA's key role in the research associated with both remediation of contaminated areas and also the natural contamination migration processes in non-remediated areas, in collaboration with other Japanese and international organisations and research institutes. Environmental monitoring and mapping to define boundary conditions in terms of the distribution of radioactivity and resultant doses, guides the resultant response. Radiation protection considerations set constraints, with approaches developed to estimate doses to different critical groups and set appropriate dose reduction targets. Decontamination activities, with special emphasis on associated waste management, provide experience in evaluation of the effectiveness of decontamination and the pros and cons of different approaches / technologies. The assessment of the natural behaviour of contaminant radionuclides and their mobility in the environment is now focused almost entirely on radiocaesium. Here, the impact of natural mobility in terms of self-cleaning / re-concentration in cleaned areas is discussed, along with possible actions to modify such transport or manage potential areas of radiocaesium accumulation. Many of the conditions in Fukushima are similar to those following past contamination events in other countries, where natural self-cleaning alone has allowed recovery to such an extent that the original incident is now largely forgotten. Decontamination efforts in Japan will certainly accelerate this process. On-going remediation work is based on a good technical understanding of the movement of radiocaesium in the environment and this understanding is being translated into

  3. Sustainability from the Occurrence of Critical Dynamic Power System Blackout Determined by Using the Stochastic Event Tree Technique

    Directory of Open Access Journals (Sweden)

    Muhammad Murtadha Othman

    2017-06-01

    Full Text Available With the advent of advanced technology in smart grid, the implementation of renewable energy in a stressed and complicated power system operation, aggravated by a competitive electricity market and critical system contingencies, this will inflict higher probabilities of the occurrence of a severe dynamic power system blackout. This paper presents the proposed stochastic event tree technique used to assess the sustainability against the occurrence of dynamic power system blackout emanating from implication of critical system contingencies such as the rapid increase in total loading condition and sensitive initial transmission line tripping. An extensive analysis of dynamic power system blackout has been carried out in a case study of the following power systems: IEEE RTS-79 and IEEE RTS-96. The findings have shown that the total loading conditions and sensitive transmission lines need to be given full attention by the utility to prevent the occurrence of dynamic power system blackout.

  4. Evaluating the effect of a campus-wide social norms marketing intervention on alcohol-use perceptions, consumption, and blackouts.

    Science.gov (United States)

    Su, Jinni; Hancock, Linda; Wattenmaker McGann, Amanda; Alshagra, Mariam; Ericson, Rhianna; Niazi, Zackaria; Dick, Danielle M; Adkins, Amy

    2018-04-01

    To evaluate the effect of a campus-wide social norms marketing intervention on alcohol-use perceptions, consumption, and blackouts at a large, urban, public university. 4,172 college students (1,208 freshmen, 1,159 sophomores, 953 juniors, and 852 seniors) who completed surveys in Spring 2015 for the Spit for Science Study, a longitudinal study of students' substance use and emotional health. Participants were e-mailed an online survey that queried campaign readership, perception of peer alcohol use, alcohol consumption, frequency of consumption, and frequency of blackouts. Associations between variables were evaluated using path analysis. We found that campaign readership was associated with more accurate perceptions of peer alcohol use, which, in turn, was associated with self-reported lower number of drinks per sitting and experiencing fewer blackouts. This evaluation supports the use of social norms marketing as a population-level intervention to correct alcohol-use misperceptions and reduce blackouts.

  5. Internal event analysis for Laguna Verde Unit 1 Nuclear Power Plant. Accident sequence quantification and results; Analisis de eventos internos para la Unidad 1 de la Central Nucleoelectrica de Laguna Verde. Cuantificacion de secuencias de accidente y resultados

    Energy Technology Data Exchange (ETDEWEB)

    Huerta B, A; Aguilar T, O; Nunez C, A; Lopez M, R [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    1994-07-01

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

  6. Cognitive Blackouts in Mild Cognitive Impairment and Alzheimer’s Dementia

    Directory of Open Access Journals (Sweden)

    Georg Adler

    2018-02-01

    Full Text Available Background: Cognitive blackouts, e.g. moments of amnesia, disorientation, or perplexity may be an early sign of incipient Alzheimer’s dementia (AD. A short questionnaire, the checklist for cognitive blackouts (CCB, was evaluated cross-sectionally in users of a memory clinic. Methods: The CCB was performed in 130 subjects, who further underwent a neuropsychological and clinical examination. Subjective memory impairment and depressive symptoms were assessed. Differences in the CCB score between diagnostic groups and relationships with cognitive performance, depression, and subjective memory impairment were analyzed. Results: The CCB score was increased in mild cognitive impairment of the amnestic type or mild AD and correctly predicted 69.2% of the respective subjects. It was negatively correlated with cognitive performance, positively correlated with depressive symptoms, and substantially increased in subjects who estimated their memory poorer than that of other persons of their age. Discussion: The CCB may be a helpful screening tool for the early recognition of AD.

  7. Nuclear accidents

    International Nuclear Information System (INIS)

    1987-01-01

    On 27 May 1986 the Norwegian government appointed an inter-ministerial committee of senior officials to prepare a report on experiences in connection with the Chernobyl accident. The present second part of the committee's report describes proposals for measures to prevent and deal with similar accidents in the future. The committee's evaluations and proposals are grouped into four main sections: Safety and risk at nuclear power plants; the Norwegian contingency organization for dealing with nuclear accidents; compensation issues; and international cooperation

  8. Estimating the Duration of Public Concern After the Fukushima Dai-ichi Nuclear Power Station Accident From the Occurrence of Radiation Exposure-Related Terms on Twitter: A Retrospective Data Analysis.

    Science.gov (United States)

    Nishimoto, Naoki; Ota, Mizuki; Yagahara, Ayako; Ogasawara, Katsuhiko

    2016-11-25

    After the Fukushima Dai-ichi Nuclear Power Station accident in Japan on March 11, 2011, a large number of comments, both positive and negative, were posted on social media. The objective of this study was to clarify the characteristics of the trend in the number of tweets posted on Twitter, and to estimate how long public concern regarding the accident continued. We surveyed the attenuation period of the first term occurrence related to radiation exposure as a surrogate endpoint for the duration of concern. We retrieved 18,891,284 tweets from Twitter data between March 11, 2011 and March 10, 2012, containing 143 variables in Japanese. We selected radiation, radioactive, Sievert (Sv), Becquerel (Bq), and gray (Gy) as keywords to estimate the attenuation period of public concern regarding radiation exposure. These data, formatted as comma-separated values, were transferred into a Statistical Analysis System (SAS) dataset for analysis, and survival analysis methodology was followed using the SAS LIFETEST procedure. This study was approved by the institutional review board of Hokkaido University and informed consent was waived. A Kaplan-Meier curve was used to show the rate of Twitter users posting a message after the accident that included one or more of the keywords. The term Sv occurred in tweets up to one year after the first tweet. Among the Twitter users studied, 75.32% (880,108/1,168,542) tweeted the word radioactive and 9.20% (107,522/1,168,542) tweeted the term Sv. The first reduction was observed within the first 7 days after March 11, 2011. The means and standard errors (SEs) of the duration from the first tweet on March 11, 2011 were 31.9 days (SE 0.096) for radioactive and 300.6 days (SE 0.181) for Sv. These keywords were still being used at the end of the study period. The mean attenuation period for radioactive was one month, and approximately one year for radiation and radiation units. The difference in mean duration between the keywords was attributed

  9. Acute Alcohol Effects on Narrative Recall and Contextual Memory: An Examination of Fragmentary Blackouts

    OpenAIRE

    Wetherill, Reagan R.; Fromme, Kim

    2011-01-01

    The present study examined the effects of alcohol consumption on narrative recall and contextual memory among individuals with and without a history of fragmentary blackouts in an attempt to better understand why some individuals experience alcohol-induced memory impairments whereas others do not, even at comparable blood alcohol concentrations (BACs). Standardized beverage (alcohol, no alcohol) administration procedures and neuropsychological assessments measured narrative recall and context...

  10. High-Tech, Low-Tech, No-Tech: Communications Strategies During Blackouts

    Science.gov (United States)

    2013-12-01

    straight-line “ derecho ”1 windstorms hit the mid-Atlantic region of the United States. In the National Capital Region (NCR), many residents lost...flooding (Grand Forks, North Dakota, 1997), rolling blackouts (California, 2001), multi-state power outage (Ohio and seven other 1 “ Derecho ” is a...flawlessly throughout the 2012 “super- derecho ”20 storm event.21 During its 2012 typhoon, the government of the Philippines used Twitter to

  11. Review of Leading Approaches for Mitigating Hypersonic Vehicle Communications Blackout and a Method of Ceramic Particulate Injection Via Cathode Spot Arcs for Blackout Mitigation

    Science.gov (United States)

    Gillman, Eric D.; Foster, John E.; Blankson, Isaiah M.

    2010-01-01

    Vehicles flying at hypersonic velocities within the atmosphere become enveloped in a "plasma sheath" that prevents radio communication, telemetry, and most importantly, GPS signal reception for navigation. This radio "blackout" period has been a problem since the dawn of the manned space program and was an especially significant hindrance during the days of the Apollo missions. An appropriate mitigation method must allow for spacecraft to ground control and ground control to spacecraft communications through the reentry plasma sheath. Many mitigation techniques have been proposed, including but not limited to, aerodynamic shaping, magnetic windows, and liquid injection. The research performed on these mitigation techniques over the years will be reviewed and summarized, along with the advantages and obstacles that each technique will need to overcome to be practically implemented. A unique approach for mitigating the blackout communications problem is presented herein along with research results associated with this method. The novel method involves the injection of ceramic metal-oxide particulate into a simulated reentry plasma to quench the reentry plasma. Injection of the solid ceramic particulates is achieved by entrainment within induced, energetic cathode spot flows.

  12. Study of Containment Vent Strategies During Severe Accident Progression for the CANDU6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Youngho; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    In March, 2011, Fukushima daichi nuclear power plants experienced a long term station blackout. Severe core damage occurred and a large amount of radioactive materials are released outside of the plants. After this terrible accident Nuclear Safety and Security Commission (NSSC) enforced to increase nuclear safety for all operating plants in Korea. To increase plant safety, both hardware reinforcement and software improvement are encouraged. Hardware reinforcement includes the preparation of the external water injection paths to the RCS and the spent fuel pool, a filtered containment venting system (CFVS), and AC power generating truck. Software improvement includes the increase of the effectiveness of the severe accident management guidance (SAMG) and plant staff training. To comply with NSSC's request, Wolsong Unit 1 has fulfilled the hardware reinforcement including the installation of a CFVS and started the extension of a SAMG to the low power and shutdown operation mode. Current SAMG deals accident occurred during full power operation only. The CFVS is designed to open and to close isolation valves manually. It does not require AC power. The operation of the CFVS prevents the reactor containment building failure due to the over-pressurization but it may release radioactive materials out of the reactor containment building. This paper discusses the radiological source terms for the containment vent strategy during severe accident progression which occurred during shutdown operation mode. This work is a part of the development of shutdown SAMG.. The CFVS is an effective means to control the containment pressure when the local air coolers are unavailable. Radioactive materials may release through the CFVS, but their amounts are reduced significantly. The alternative means, i.e., containment vent through the ventilation system which does not have an effective filter, is not a good choice to control the containment condition. It can maintain the containment

  13. International collaboration for development of accident-resistant LWR fuel. International Collaboration for Development of Accident Resistant Light Water Reactor Fuel

    International Nuclear Information System (INIS)

    Sowder, Andrew

    2013-01-01

    Following the March 2011 multi-unit accident at the Fukushima Daiichi plant, there has been increased interest in the development of breakthrough nuclear fuel designs that can reduce or eliminate many of the outcomes of a severe accident at a light water reactor (LWR) due to loss of core cooling following an extended station blackout or other initiating event. With this interest and attention comes a unique opportunity for the nuclear industry to fundamentally change the nature and impact of severe accidents. Clearly, this is no small feat. The challenges are many and the technical barriers are high. Early estimates for moving maturing R and D concepts to the threshold of commercialisation exceed one billion USD. Given the anticipated effort and resources required, no single entity or group can succeed alone. Accordingly, the Electric Power Research Institute (EPRI) sees the need for and promise of cooperation among many stakeholders on an international scale to bring about what could be transformation in LWR fuel performance and robustness. An important initial task in any R and D programme is to define the goals and metrics for measuring success. As starting points for accident-tolerant fuel development, the extension of core coolability under loss of coolant conditions and the elimination or reduction of hydrogen generation are widely recognised R and D endpoints for deployment. Furthermore, any new LWR fuel technology will, at a minimum, need to (1) be compatible with the safe, economic operation of existing plants and (2) maintain acceptable or improve nuclear fuel performance under normal operating conditions. While the primary focus of R and D to date has been on cladding and fuel improvements, there are a number of other potential paths to improve outcomes following a severe accident at an LWR that include modifications to other fuel hardware and core internals to fully address core coolability, criticality, and hydrogen generation concerns. The US

  14. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of atmospheric dispersion model with improved deposition scheme and oceanic dispersion model

    Science.gov (United States)

    Katata, G.; Chino, M.; Kobayashi, T.; Terada, H.; Ota, M.; Nagai, H.; Kajino, M.; Draxler, R.; Hort, M. C.; Malo, A.; Torii, T.; Sanada, Y.

    2014-06-01

    Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Dai-ichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate a detailed time trend of atmospheric releases during the accident by combining environmental monitoring data with atmospheric model simulations from WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information), and simulations from the oceanic dispersion model SEA-GEARN-FDM, both developed by the authors. A sophisticated deposition scheme, which deals with dry and fogwater depositions, cloud condensation nuclei (CCN) activation and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I2 and CH3I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The fallout to the ocean surface calculated by WSPEEDI-II was used as input data for the SEA-GEARN-FDM calculations. Reverse and inverse source-term estimation methods based on coupling the simulations from both models was adopted using air dose rates and concentrations, and sea surface concentrations. The results revealed that the major releases of radionuclides due to FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, the morning of 13 March after the venting event at Unit 3, midnight of 14 March when the SRV (Safely Relief Valve) at Unit 2 was opened three times, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of

  15. The analysis of the radiation induced cancer risks of workers of the nuclear industry - Liquidators of the accident on Chernobyl Atomic Station on the basis of modified poisson regressions

    International Nuclear Information System (INIS)

    Shafransky, I.L.; Tukov, A.R.

    2008-01-01

    Full text: The purpose of work consisted in reception of adequate estimations for additional relative risk in recalculation on 1 Sv in two dose diapason: up to 200 mSv and up to 500 mSv on the basis of materials on prevalence of malignant disease of workers of the nuclear industry - liquidators of the accident on Chernobyl Atomic Station. For this purpose methods of cohort analysis were used. This method realized on the basis of Poisson regression has been used. Estimations ERR on 1 Sv have been calculated as under the traditional scheme with use of module AMFIT (software EPICURE), and under the modified formula offered Paretzke. The received results have shown, that in some cases estimations for the risks, received on the modified formula, are more realistic, in other cases both estimations have close values. Also the lead analysis has shown no correct the procedure of carry of estimations of the risk received on one dose interval, on another a dose interval in a kind of nonlinear dependence of function of risk from a doze. As a whole, it is possible to tell, on an interval up to 200 mSv estimations of risk demand use of more complexes, than regression, models. In a range of dozes up to 500 mSv and even up to 1000 mSv the estimation of risk under the modified formula is more adequate. In a range of small doses application of the traditional approach on the basis of Linear non-threshold concept cannot be statistically justified and correct. (author)

  16. Optimization of station battery replacement

    International Nuclear Information System (INIS)

    Jancauskas, J.R.; Shook, D.A.

    1994-01-01

    During a loss of ac power at a nuclear generating station (including diesel generators), batteries provide the source of power which is required to operate safety-related components. Because traditional lead-acid batteries have a qualified life of 20 years, the batteries must be replaced a minimum of once during a station's lifetime, twice if license extension is pursued, and more often depending on actual in-service dates and the results of surveillance tests. Replacement of batteries often occurs prior to 20 years as a result of systems changes caused by factors such as Station Blackout Regulations, control system upgrades, incremental load growth, and changes in the operating times of existing equipment. Many of these replacement decisions are based on the predictive capabilities of manual design basis calculations. The inherent conservatism of manual calculations may result in battery replacements occurring before actually required. Computerized analysis of batteries can aid in optimizing the timing of replacements as well as in interpreting service test data. Computerized analysis also provides large benefits in maintaining the as-configured load profile and corresponding design margins, while also providing the capability of quickly analyze proposed modifications and response to internal and external audits

  17. 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

  18. Large scale experiments simulating hydrogen distribution in a spent fuel pool building during a hypothetical fuel uncovery accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Mignot, Guillaume; Paranjape, Sidharth; Paladino, Domenico; Jaeckel, Bernd; Rydl, Adolf [Paul Scherrer Institute, Villigen (Switzerland)

    2016-08-15

    Following the Fukushima accident and its extended station blackout, attention was brought to the importance of the spent fuel pools' (SFPs) behavior in case of a prolonged loss of the cooling system. Since then, many analytical works have been performed to estimate the timing of hypothetical fuel uncovery for various SFP types. Experimentally, however, little was done to investigate issues related to the formation of a flammable gas mixture, distribution, and stratification in the SFP building itself and to some extent assess the capability for the code to correctly predict it. This paper presents the main outcomes of the Experiments on Spent Fuel Pool (ESFP) project carried out under the auspices of Swissnuclear (Framework 2012–2013) in the PANDA facility at the Paul Scherrer Institut in Switzerland. It consists of an experimental investigation focused on hydrogen concentration build-up into a SFP building during a predefined scaled scenario for different venting positions. Tests follow a two-phase scenario. Initially steam is released to mimic the boiling of the pool followed by a helium/steam mixture release to simulate the deterioration of the oxidizing spent fuel. Results shows that while the SFP building would mainly be inerted by the presence of a high concentration of steam, the volume located below the level of the pool in adjacent rooms would maintain a high air content. The interface of the two-gas mixture presents the highest risk of flammability. Additionally, it was observed that the gas mixture could become stagnant leading locally to high hydrogen concentration while steam condenses. Overall, the experiments provide relevant information for the potentially hazardous gas distribution formed in the SFP building and hints on accident management and on eventual retrofitting measures to be implemented in the SFP building.

  19. 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)

  20. Accident Statistics

    Data.gov (United States)

    Department of Homeland Security — Accident statistics available on the Coast Guard’s website by state, year, and one variable to obtain tables and/or graphs. Data from reports has been loaded for...

  1. Radiation accidents

    International Nuclear Information System (INIS)

    Nenot, J.C.

    1996-01-01

    Analysis of radiation accidents over a 50 year period shows that simple cases, where the initiating events were immediately recognised, the source identified and under control, the medical input confined to current handling, were exceptional. In many cases, the accidents were only diagnosed when some injuries presented by the victims suggested the radiological nature of the cause. After large-scale accidents, the situation becomes more complicated, either because of management or medical problems, or both. The review of selected accidents which resulted in severe consequences shows that most of them could have been avoided; lack of regulations, contempt for rules, human failure and insufficient training have been identified as frequent initiating parameters. In addition, the situation was worsened because of unpreparedness, insufficient planning, unadapted resources, and underestimation of psychosociological aspects. (author)

  2. Sports Accidents

    CERN Multimedia

    Kiebel

    1972-01-01

    Le Docteur Kiebel, chirurgien à Genève, est aussi un grand ami de sport et de temps en temps médecin des classes genevoises de ski et également médecin de l'équipe de hockey sur glace de Genève Servette. Il est bien qualifié pour nous parler d'accidents de sport et surtout d'accidents de ski.

  3. Radiation accidents

    International Nuclear Information System (INIS)

    Poplavskij, K.K.; Smorodintseva, G.I.

    1978-01-01

    On the basis of a critical analysis of the available data on causes and consequences of radiation accidents (RA), a classification of RA by severity (five groups of accidents) according to biomedical consequences and categories of exposed personnel is proposed. A RA is defined and its main characteristics are described. Methods of RA prevention are proposed, as is a plan of specific measures to deal with RA in accordance with the proposed classification

  4. 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.)

  5. Real-time stability in power systems techniques for early detection of the risk of blackout

    CERN Document Server

    Savulescu, Savu

    2014-01-01

    This pioneering volume has been updated and enriched to reflect the state-of-the-art in blackout prediction and prevention. It documents and explains background and algorithmic aspects of the most successful steady-state, transient and voltage stability solutions available today in real-time. It also describes new, cutting-edge stability applications of synchrophasor technology, and captures industry acceptance of metrics and visualization tools that quantify and monitor the distance to instability. Expert contributors review a broad spectrum of additionally available techniques, such as traje

  6. Characterization of weakly ionized argon flows for radio blackout mitigation experiments

    Science.gov (United States)

    Steffens, L.; Koch, U.; Esser, B.; Gülhan, A.

    2017-06-01

    For reproducing the so-called E × B communication blackout mitigation scheme inside the L2K arc heated facility of the DLR in weakly ionized argon §ows, a §at plate model has been equipped with a superconducting magnet, electrodes, and a setup comprising microwave plasma transmission spectroscopy (MPTS). A thorough characterization of the weakly ionized argon §ow has been performed including the use of microwave interferometry (MWI), Langmuir probe measurements, Pitot probe pro¦les, and spectroscopic methods like diode laser absorption spectroscopy (DLAS) and emission spectroscopy.

  7. Resonant interaction of electromagnetic wave with plasma layer and overcoming the radiocommunication blackout problem

    Science.gov (United States)

    Bogatskaya, A. V.; Klenov, N. V.; Tereshonok, M. V.; Adjemov, S. S.; Popov, A. M.

    2018-05-01

    We present an analysis of the possibility of penetrating electromagnetic waves through opaque media using an optical-mechanical analogy. As an example, we consider the plasma sheath surrounding the vehicle as a potential barrier and analyze the overcoming of radiocommunication blackout problem. The idea is to embed a «resonator» between the surface on the vehicle and plasma sheath which is supposed to provide an effective tunneling of the signal to the receiving antenna. We discuss the peculiarities of optical mechanical analogy applicability and analyze the radio frequency wave tunneling regime in detail. The cases of normal and oblique incidence of radiofrequency waves on the vehicle surface are studied.

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

    International Nuclear Information System (INIS)

    De Rosa, Felice

    2006-01-01

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

  9. Advanced Approach to Consider Aleatory and Epistemic Uncertainties for Integral Accident Simulations

    International Nuclear Information System (INIS)

    Peschke, Joerg; Kloos, Martina

    2013-01-01

    , and two application examples are briefly presented. The first application refers to a station black-out scenario. The other application is an analysis of the emergency operating procedure 'Secondary Side Bleed and Feed' which has to be applied after the loss of steam generator feed-water supply. (authors)

  10. Preliminary thermal design of a pressurized water reactor containment for handling severe accident consequences

    International Nuclear Information System (INIS)

    Abdullah, A.M.; Karameldin, A.

    1998-01-01

    A one-dimensional mathematical model has been developed for a 4250 MW(th) Advanced Pressurized Water Reactor containment analysis following a severe accident. The cooling process of the composite containment-steel shell and concrete shield- is achievable by natural circulation of atmospheric air. However, for purpose of gettering higher degrees of safety margin, the present study undertakes two objectives: (1) Installment of a diesel engine-driven air blower to force air through the annular space between the steel shell and concrete shield. The engine can be remotely operated to be effective in case of station blackout. (ii) Fixing longitudinally plate fins on the circumference of the inside and outside containment steel shell. These fins increase the heat transfer areas and hence the rate of heat removal from the containment atmosphere. In view of its importance - from the safety viewpoint - the long term behaviour of the containment which is a quasi-steady state problem, is formulated through a system of coupled nonlinear algebraic equations which describe the thermal-hydraulic and thermodynamic behaviour of the double shell containment. The calculated results revealed the following: (i) the passively air cooled containment can remove maximum heat load of 11.5 MW without failure, (ii) the effect of finned surface in the air passage tends to decrease the containment pressure by 20 to 30%, depending on the heat load, (iii) the effect of condensing fins is negligible for the proposed fin dimensions and material. However, by reducing the fin width, increasing their thickness, doubling their number, and using a higher conductive metal than the steel, it is expected that the containment pressure can be further reduced by 10% or more, (iv) the fins' dimensions and their number must be optimized via maximizing the difference or the ratio between the heat removed and pressure drop to get maximum heat flow rate

  11. Ecosystem monitoring on dispersion and export dynamics of ¹³⁷Cs deposited on the forested area in Fukushima after the nuclear power stations accident in March 2011

    Energy Technology Data Exchange (ETDEWEB)

    Ohte, Nobuhito; Endo, Izuki; Iseda, Kohei; Tanoi, Keitaro [University of Tokyo, Bunkyo-ku, Tokyo, 113-8657 (Japan); Murakami, Masashi [Chiba University, Inage-ku, Chiba, 263-8522 (Japan); Ishii, Nobuyoshi [National Institute of Radiological Sciences, Inage-ku, Chiba, 263-8555 (Japan)

    2014-07-01

    The accident of the Fukushima Daiichi Nuclear Power Stations in March 2011 emitted 1.5x10¹⁷ Bq of ¹³¹I, 1.2x10¹⁶ Bq of ¹³⁷Cs into the surrounding environment. Those radioactive substances including ¹³⁷Cs deposited onto the forest areas of the northeastern region of the Honshu Island, Japan. Forests of these regions are especially important for the people not only for their forest production industry, but also for the source areas of drinking water. The first phase of the governmental survey and investigations showed that major portion of deposited ¹³⁷Cs is trapped on the canopies and litter layers of soil surface. As ¹³⁷Cs is easily adsorbed onto clay minerals in soil, major portion of those can be transported by eroded soils and particulate organic matters through the hydrological pathways. Dissolved ¹³⁷Cs which is relatively free from soil adsorption can be taken up by microbes, algae and plants in soil and aquatic systems. Then, those ¹³⁷Cs are eventually introduced into soil insects, worms, aquatic insects, fishes and birds through the food web. To clarify the mechanisms of dispersion and export of ¹³⁷Cs within and from a forested ecosystem, we started intensive field observations and samplings in a small catchment including forests and farmlands in the area ~50 km distant from the Nuclear Power Station. The study is conducted at the Kami-Oguni River catchment in the northern part of Fukushima Prefecture. Major land uses of the catchment are forests and paddy fields. Expected two major pathways of ¹³⁷Cs diffusion and export are investigated; 1) transportation by water movement with dissolved and particulate or colloidal forms through the hydrological processes, 2) dispersion through the food web including both detritus food chains based upon plant litters and grazing food chains based upon living plant leaves in the forest-stream ecological continuum. The ¹³⁷Cs concentrations of stream waters at several sampling

  12. Criticality accident:

    International Nuclear Information System (INIS)

    Canavese, Susana I.

    2000-01-01

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

  13. Advanced validation of CFD-FDTD combined method using highly applicable solver for reentry blackout prediction

    International Nuclear Information System (INIS)

    Takahashi, Yusuke

    2016-01-01

    An analysis model of plasma flow and electromagnetic waves around a reentry vehicle for radio frequency blackout prediction during aerodynamic heating was developed in this study. The model was validated based on experimental results from the radio attenuation measurement program. The plasma flow properties, such as electron number density, in the shock layer and wake region were obtained using a newly developed unstructured grid solver that incorporated real gas effect models and could treat thermochemically non-equilibrium flow. To predict the electromagnetic waves in plasma, a frequency-dependent finite-difference time-domain method was used. Moreover, the complicated behaviour of electromagnetic waves in the plasma layer during atmospheric reentry was clarified at several altitudes. The prediction performance of the combined model was evaluated with profiles and peak values of the electron number density in the plasma layer. In addition, to validate the models, the signal losses measured during communication with the reentry vehicle were directly compared with the predicted results. Based on the study, it was suggested that the present analysis model accurately predicts the radio frequency blackout and plasma attenuation of electromagnetic waves in plasma in communication. (paper)

  14. Interpreting transnational infrastructure vulnerability: European blackout and the historical dynamics of transnational electricity governance

    International Nuclear Information System (INIS)

    Vleuten, Erik van der; Lagendijk, Vincent

    2010-01-01

    Recent transnational blackouts exposed two radically opposed interpretations of Europe's electricity infrastructure, which inform recent and ongoing negotiations on transnational electricity governance. To EU policy makers such blackouts revealed the fragility of Europe's power grids and the need of a more centralized form of governance, thus legitimizing recent EU interventions. Yet to power sector spokespersons, these events confirmed the reliability of transnational power grids and the traditional decentralized governance model: the disturbances were quickly contained and repaired. This paper inquires the historic legacies at work in these conflicting interpretations and associated transnational governance preferences. It traces the power sector's interpretation to its building of a secure transnational power grid from the 1950s through the era of neoliberalization. Next it places the EU interpretation and associated policy measures against the historical record of EU attempts at transnational infrastructure governance. Uncovering the historical roots and embedding of both interpretations, we conclude that their divergence is of a surprisingly recent date and relates to the current era of security thinking. Finally we recommend transnational, interpretative, and historical analysis to the field of critical infrastructure studies.

  15. Analysis of Radio Frequency Blackout for a Blunt-Body Capsule in Atmospheric Reentry Missions

    Directory of Open Access Journals (Sweden)

    Yusuke Takahashi

    2016-01-01

    Full Text Available A numerical analysis of electromagnetic waves around the atmospheric reentry demonstrator (ARD of the European Space Agency (ESA in an atmospheric reentry mission was conducted. During the ARD mission, which involves a 70% scaled-down configuration capsule of the Apollo command module, radio frequency blackout and strong plasma attenuation of radio waves in communications with data relay satellites and air planes were observed. The electromagnetic interference was caused by highly dense plasma derived from a strong shock wave generated in front of the capsule because of orbital speed during reentry. In this study, the physical properties of the plasma flow in the shock layer and wake region of the ESA ARD were obtained using a computational fluid dynamics technique. Then, electromagnetic waves were expressed using a frequency-dependent finite-difference time-domain method using the plasma properties. The analysis model was validated based on experimental flight data. A comparison of the measured and predicted results showed good agreement. The distribution of charged particles around the ESA ARD and the complicated behavior of electromagnetic waves, with attenuation and reflection, are clarified in detail. It is suggested that the analysis model could be an effective tool for investigating radio frequency blackout and plasma attenuation in radio wave communication.

  16. An adaptive reentry guidance method considering the influence of blackout zone

    Science.gov (United States)

    Wu, Yu; Yao, Jianyao; Qu, Xiangju

    2018-01-01

    Reentry guidance has been researched as a popular topic because it is critical for a successful flight. In view that the existing guidance methods do not take into account the accumulated navigation error of Inertial Navigation System (INS) in the blackout zone, in this paper, an adaptive reentry guidance method is proposed to obtain the optimal reentry trajectory quickly with the target of minimum aerodynamic heating rate. The terminal error in position and attitude can be also reduced with the proposed method. In this method, the whole reentry guidance task is divided into two phases, i.e., the trajectory updating phase and the trajectory planning phase. In the first phase, the idea of model predictive control (MPC) is used, and the receding optimization procedure ensures the optimal trajectory in the next few seconds. In the trajectory planning phase, after the vehicle has flown out of the blackout zone, the optimal reentry trajectory is obtained by online planning to adapt to the navigation information. An effective swarm intelligence algorithm, i.e. pigeon inspired optimization (PIO) algorithm, is applied to obtain the optimal reentry trajectory in both of the two phases. Compared to the trajectory updating method, the proposed method can reduce the terminal error by about 30% considering both the position and attitude, especially, the terminal error of height has almost been eliminated. Besides, the PIO algorithm performs better than the particle swarm optimization (PSO) algorithm both in the trajectory updating phase and the trajectory planning phases.

  17. Silencing Boko Haram: Mobile Phone Blackout and Counterinsurgency in Nigeria’s Northeast region

    Directory of Open Access Journals (Sweden)

    Jacob Udo-Udo Jacob

    2015-03-01

    Full Text Available In the summer of 2013, the Nigerian military, as part of its counterinsurgency operations against Boko Haram insurgents, shut down GSM mobile telephony in three northeast states – Adamawa, Borno and Yobe. This article explores the rationale, impact and citizens’ opinion of the mobile phone blackout. It draws on focus group discussions with local opinion leaders and in-depth personal interviews with military and security insiders, as well as data of Boko Haram incidences before, during and after the blackout from military sources and conflict databases. It argues that, although the mobile phone shutdown was ‘successful’ from a military- tactical point of view, it angered citizens and engendered negative opinions toward the state and new emergency policies. While citizens developed various coping and circumventing strategies, Boko Haram evolved from an open network model of insurgency to a closed centralized system, shifting the center of its operations to the Sambisa Forest. This fundamentally changed the dynamics of the conflict. The shutdown demonstrated, among others, that while ICTs serve various desirable purposes for developing states, they will be jettisoned when their use challenges the state’s legitimacy and raison d'être, but not without consequences.

  18. Assessment of impact of a severe accident at nuclear power plant of Angra dos Reis with release of radionuclides to the atmosphere; Avaliacao do impacto de um acidente severo na usina de Angra dos Reis com liberacao de radionuclideos para a atmosfera

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, Andre Silva de

    2015-07-01

    This study had as purpose the assess the impact of a severe accident, and also analyze the dispersion of {sup 131}I in the atmosphere, so that, through concentrating and inhaling dose of the plume, were possible to verify if the results are in accordance with the indicated data by the Plan of Emergency of the CNAAA regarding the Impact Zone and Control. This exercise was performed with the aid of an atmospheric model and a dispersion where to atmospheric modeling we used the data coupling WRF / CALMET and of dispersion, CALPUFF. The suggested accident consists of a Station Blackout at Nuclear Power of Angra (Unit 1), where through the total core involvement, will release 100% of the {sup 131}I to the atmosphere. The value of the total activity in the nucleus to this radionuclide is 7.44 x 1017 Bq, that is relative on the sixth day of burning. This activity will be released through the chimney at a rate in Bq/s in the scenario of 12, 24, 48 and 72 hours of release. Applying the model in the proposed scenario, it is verified that the plume has concentrations of the order of 1020 Bq/m³ and dose of about 108 Sv whose value is beyond of the presented by Eletronuclear in your current emergency plan. (author)

  19. Development of filtered containment venting system and application for Kashiwazaki-Kariwa Nuclear Power Station Unit 6, 7

    International Nuclear Information System (INIS)

    Murai, Soutarou; Hiranuma, Naoki; Kimura, Takeo; Omori, Shuichi; Watanabe, Fumitoshi; Sasa, Daisuke

    2014-01-01

    The Fukushima Dai-ichi Nuclear Power Station (1F) of Tokyo Electric Power Company (TEPCO) had experienced severe radio-active release to the environment in the Tohoku Region Pacific Coast Earthquake (alias: the Great East Japan Earthquake) in 2011. Under the Station Black-Out (SBO) conditions caused by tsunami with the earthquake, the 1F operators had tried to vent the gasses in the Primary Containment Vessels (PCVs) of the unit 1, 2 and 3 to the environment through the water pools in the suppression chambers of the PCVs. Its venting, however, was imperfect and, as a result, major direct radio-active release to the environment was caused. After this disaster, TEPCO launched a project to develop the Filtered Containment Venting System (FCVS), in which our very bitter experiences in the 1F accident as described above are reflected. One of the main purposes of the development of the FCVS is to enhance operability of venting under the severe plant conditions such as the SBO during progressing of severe core damage, and another is to enhance removal performance of radio-nuclides with the newly added filtering equipment, which is installed in the venting line from the PCV to the outer. The Kashiwazaki-Kariwa NPS unit 6 and 7 will be the first reactors applied the FCVSs. In this paper, we show the design concept of the TEPCO's FCVS, the brief overview of the system design and the summary of experiment which has been performed for getting the performance data of the FCVS such as decontamination factor in various conditions. (author)

  20. Evaluating the Effect of a Campus-Wide Social Norms Marketing Intervention on Alcohol-Use Perceptions, Consumption, and Blackouts

    Science.gov (United States)

    Su, Jinni; Hancock, Linda; Wattenmaker McGann, Amanda; Alshagra, Mariam; Ericson, Rhianna; Niazi, Zackaria; Dick, Danielle M.; Adkins, Amy

    2018-01-01

    Objective: To evaluate the effect of a campus-wide social norms marketing intervention on alcohol-use perceptions, consumption, and blackouts at a large, urban, public university. Participants: 4,172 college students (1,208 freshmen, 1,159 sophomores, 953 juniors, and 852 seniors) who completed surveys in Spring 2015 for the Spit for Science…

  1. Quality function deployment applied to local traffic accident reduction.

    Science.gov (United States)

    Sohn, S Y

    1999-11-01

    One of the major tasks of police stations is the management of local road traffic accidents. Proper prevention policy which reflects the local accident characteristics could immensely help individual police stations in decreasing various severity levels of road traffic accidents. In order to relate accident variation to local driving environmental characteristics, we use both cluster analysis and Poisson regression. The fitted result at the level of each cluster for each type of accident severity is utilized as an input to quality function deployment. Quality function deployment (QFD) has been applied to customer satisfaction in various industrial quality improvement settings, where several types of customer requirements are related to various control factors. We show how QFD enables one to set priorities on various road accident control policies to which each police station has to pay particular attention.

  2. Fire Stations

    Data.gov (United States)

    Department of Homeland Security — Fire Stations in the United States Any location where fire fighters are stationed or based out of, or where equipment that such personnel use in carrying out their...

  3. Calculations of a station blackout transient in a RBMK type nuclear power plant with the CATHARE code

    International Nuclear Information System (INIS)

    Niklaus, F.; Korteniemi, V.

    1996-01-01

    At the Department of Energy Technology at Lappeenranta University of Technology a CATHARE model of one unit of the St. Petersburg (RBMK) nuclear power plant has been generated. The investigations have been done in order to understand better the thermal-hydraulic behaviour of RBMK type reactors and in order to see how far the French thermal-hydraulic safety code CATHARE can predict the physical phenomena during various RBMK transients. (12 refs.)

  4. Torness: proposed nuclear power station

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The need for and desirability of nuclear power, and in particular the proposed nuclear power station at Torness in Scotland, are questioned. Questions are asked, and answered, on the following topics: position, appearance and cost of the proposed Torness plant, and whether necessary; present availability of electricity, and forecast of future needs, in Scotland; energy conservation and alternative energy sources; radiation hazards from nuclear power stations (outside, inside, and in case of an accident); transport of spent fuel from Torness to Windscale; radioactive waste management; possibility of terrorists making a bomb with radioactive fuel from a nuclear power station; cost of electricity from nuclear power; how to stop Torness. (U.K.)

  5. Evaluation of severe accident risks and the potential for risk reduction: Peach Bottom, Unit 2. Main report. Draft for comment, February 1987

    Energy Technology Data Exchange (ETDEWEB)

    Amos, C N [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States); Benjamin, A S; Griesmeyer, J M; Haskin, F E; Kunsman, D M [Sandia National Laboratories, Albuquerque, NM (United States); Boyd, G J; Lewis, S R [Safety and Reliability Optimization Services, Inc., Knoxville, TN (United States); Helton, J C [Arizona State University, Tempe, AZ (United States); Smith, L N [Science Applications International Corporation, Albuquerque, NM (United States)

    1987-04-01

    The Severe Accident Risk Reduction Program (SARRP) has completed a rebaselining of the risks to the public from a boiling water reactor with a Mark I containment (Peach Bottom, Unit 2). Emphasis was placed on determining the magnitude and character of the uncertainties, rather than focusing on a point estimate. The risk-reduction potential of a set of proposed safety option backfits was also studied, and their costs and benefits were also evaluated. It was found that the risks from internal events are generally low relative to previous studies; for example, most of the uncertainty range is lower than the point estimate of risk for the Peach Bottom plant in the Reactor Safety Study (RSS). However, certain unresolved issues cause the top of the uncertainty band to appear at a level that is comparable with the RSS point estimate. These issues include the modeling of the common-mode failures for the dc power system, the likelihood of offsite power recovery versus time during a station blackout, the probability of drywell failure resulting from meltthrough of the drywell shell, the magnitude of the fission product releases during core-concrete interactions, and the decontamination effectiveness of the reactor enclosure building. Most of the postulated safety options do not appear to be cost effective, although some based on changes to procedures or inexpensive hardware additions may be marginally cost effective. This draft for comment of the SARRP report for Peach Bottom does not include detailed technical appendices, which are still in preparation. The appendices will be issued under separate cover when completed. This work supports the Nuclear Regulatory Commission's assessment of severe accidents in NUREG-1150. (author)

  6. Tchernobyl accident

    International Nuclear Information System (INIS)

    1986-06-01

    First, R.M.B.K type reactors are described. Then, safety problems are dealt with reactor control, behavior during transients, normal loss of power and behavior of the reactor in case of leak. A possible scenario of the accident of Tchernobyl is proposed: events before the explosion, possible initiators, possible scenario and events subsequent to the core meltdown (corium-concrete interaction, interaction with the groundwater table). An estimation of the source term is proposed first from the installation characteristics and the supposed scenario of the accident, and from the measurements in Europe; radiological consequences are also estimated. Radioactivity measurements (Europe, Scandinavia, Western Europe, France) are given in tables (meteorological maps and fallouts in Europe). Finally, a description of the site is given [fr

  7. Accident: Reminder

    CERN Multimedia

    2003-01-01

    There is no left turn to Point 1 from the customs, direction CERN. A terrible accident happened last week on the Route de Meyrin just outside Entrance B because traffic regulations were not respected. You are reminded that when travelling from the customs, direction CERN, turning left to Point 1 is forbidden. Access to Point 1 from the customs is only via entering CERN, going down to the roundabout and coming back up to the traffic lights at Entrance B

  8. Out of the Blackout and into the Light: How the Arts Survived Pinochet’s Dictatorship

    Directory of Open Access Journals (Sweden)

    Paula Thorrington Cronovich

    2014-06-01

    This article shows how various artists segued out of the cultural blackout of the late seventies and into a phase of surprising artistic production during the military regime in Chile. At a time when political parties were banned and public gatherings considered illegal, Chileans found alternative ways to oppose the military government. In this climate, I argue that artistic expression took on political meaning. The fact that the “No” Campaign of 1988 was able to oust the dictator with an optimistic message of joy and hope, attests to the point that Chileans were able to shed their fears and change their outlook. Throughout the decade, the arts—innovations in poetry, music, theater, narrative and the audiovisual media—had offered people a much-needed forum for expression.

  9. Contributions of the restructuring of the electric power industry to the August 14, 2003 blackout

    International Nuclear Information System (INIS)

    Casazza, J.; Delea, F.; Loehr, G.

    2005-01-01

    A review of the roles of industry and government in the 2003 blackout was presented. This white paper was prepared by a group of engineers with high level experience in the electric power industry who are concerned that deregulation of the industry has led to a significant decrease in reliability. It was noted that post-blackout reviews have focused on technical failures instead of examining the responsibilities and failures of the National Electric Reliability Council (NERC). Deficiencies in the analytical capabilities of control centres were discussed, as well as issues concerning communication protocols and training. Deregulation and the concomitant restructuring of the electric power industry has led to a shift from long term optimization, inter-system coordination and reliability towards dependence on immediate profits. In addition, there have been significant reductions in personnel at electric power organizations and companies, as well as increasing complexity in operations. Increased complexity has resulted in a dilution of management responsibility, as well as over-reliance on markets to solve scientifically complex problems. There have also been cutbacks in training and research. The functional separation of generation and transmission within companies has contributed to the diffusion of best technical knowledge. Many private utilities have divested their generation resources in response to regulatory pressures. The entrance of merchant power plants in the power system has led to the establishment of new market areas that are inconsistent with the boundaries of responsible operating entities. It was concluded that all these changes have created a more complicated and compartmentalized industry structure. Decisions are now made by a large number of entities, most of which are competitors and each of which has more interest in profit than in bulk power system reliability. Procedural rules established between and among the various parties are no longer

  10. Monitoring of nuclear power stations

    International Nuclear Information System (INIS)

    Ull, E.; Labudda, H.J.

    1987-01-01

    The purpose of the invention is to create a process for undelayed automated detection and monitoring of accidents in the operation of nuclear power stations. According to the invention, this problem is solved by the relevant local measurements, such as radiation dose, components and type of radiation and additional relevant meteorological parameters being collected by means of wellknown data collection platforms, these being transmitted via transmission channels by means of satellites to suitable worldwide situated receiving stations on the ground, being processed there and being evaluated to recognise accidents. The local data collection platforms are used in the immediate vicinity of the nuclear power station. The use of aircraft, ships and balloons as data collection systems is also intended. (HWJ)

  11. Exchange of pressurizer safeguarding system at Biblis nuclear power station

    International Nuclear Information System (INIS)

    Weber, D.; Hofbeck, W.

    1991-01-01

    Valves and piping of the pressurizer safeguarding system are exchanged and reset in such a way that they are suitable not only for discharging steam, but also for discharging a water-steam mixture and hot pressurized water; for the emergency measure of primary depressurization by hand (bleed) in the event of failure of the entire feedwater supply and station black-out, and in the event of operational transients with supposed failure of the reactor scram (ATWS). To achieve this, in addition to the requirements of the pressurizer discharging station, changes have to be made to the valve drive to dominate the water loads. During the 1990 inspection this exchange of the pressurizer discharging station was performed at the Biblis A unit as the first German plant. (orig.) [de

  12. Simulations of the design basis accident at conditions of power increase and the o transient of MSIV at overpressure conditions of the Laguna Verde Power Station; Simulaciones del accidente base de diseno a condiciones de aumento de potencia y del transitorio de cierre de MSIV a condiciones de sobrepresion de la Central Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Araiza M, E.; Nunez C, A. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    2001-07-01

    This document presents the analysis of the simulation of the loss of coolant accident at uprate power conditions, that is 2027 MWt (105% of the current rated power of 1931MWt). This power was reached allowing an increase in the turbine steam flow rate without changing the steam dome pressure value at its rated conditions (1020 psiaJ. There are also presented the results of the simulation of the main steam isolation va/ve transient at overpressure conditions 1065 psia and 1067 MWt), for Laguna Verde Nuclear Power Station. Both simulations were performed with the best estimate computer code TRA C BF1. The results obtained in the loss of coolant accident show that the emergency core coolant systems can recover the water level in the core before fuel temperature increases excessively, and that the peak pressure reached in the drywell is always below its design pressure. Therefore it is concluded that the integrity of the containment is not challenged during a loss of coolant accident at uprate power conditions.The analysis of the main steam isolation valve transients at overpressure conditions, and the analysis of the particular cases of the failure of one to six safety relief valves to open, show that the vessel peak pressures are below the design pressure and have no significant effect on vessel integrity. (Author)

  13. Prevention of pedestrian accidents.

    OpenAIRE

    Kendrick, D

    1993-01-01

    Child pedestrian accidents are the most common road traffic accident resulting in injury. Much of the existing work on road traffic accidents is based on analysing clusters of accidents despite evidence that child pedestrian accidents tend to be more dispersed than this. This paper analyses pedestrian accidents in 573 children aged 0-11 years by a locally derived deprivation score for the years 1988-90. The analysis shows a significantly higher accident rate in deprived areas and a dose respo...

  14. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    Castillo G, F.

    2015-01-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

  15. Amtrak Stations

    Data.gov (United States)

    Department of Homeland Security — Updated database of the Federal Railroad Administration's (FRA) Amtrak Station database. This database is a geographic data set containing Amtrak intercity railroad...

  16. Radiation accidents

    International Nuclear Information System (INIS)

    Saenger, E.L.

    1986-01-01

    It is essential that emergency physicians understand ways to manage patients contaminated by radioactive materials and/or exposed to external radiation sources. Contamination accidents require careful surveys to identify the metabolic pathway of the radionuclides to guide prognosis and treatment. The level of treatment required will depend on careful surveys and meticulous decontamination. There is no specific therapy for the acute radiation syndrome. Prophylactic antibodies are desirable. For severely exposed patients treatment is similar to the supportive care given to patients undergoing organ transplantation. For high-dose extremity injury, no methods have been developed to reverse the fibrosing endarteritis that eventually leads to tissue death so frequently found with this type of injury. Although the Three Mile Island episode of March 1979 created tremendous public concern, there were no radiation injuries. The contamination outside the reactor building and the release of radioiodine were negligible. The accidental fuel element meltdown at Chernobyl, USSR, resulted in many cases of acute radiation syndrome. More than 100,000 people were exposed to high levels of radioactive fallout. The general principles outlined here are applicable to accidents of that degree of severity

  17. Chernobyl accident

    International Nuclear Information System (INIS)

    Bar'yakhtar, V.G.

    1995-01-01

    The monograph contains the catastrophe's events chronology, the efficiency assessed of those measures assumed for their localization as well as their environmental and socio-economic impact. Among materials of the monograph the results are presented of research on the radioactive contamination field forming as well as those concerning the investigation of biogeochemical properties of Chernobyl radionuclides and their migration process in the environment of the Ukraine. The data dealing with biological effects of the continued combined internal and external radioactive influence on plants, animals and human health under the circumstances of Chernobyl accident are of the special interest. In order to provide the scientific generalizing information on the medical aspects of Chernobyl catastrophe, the great part of the monograph is allotted to appraise those factors affecting the health of different population groups as well as to depict clinic aspects of Chernobyl events and medico-sanitarian help system. The National Programme of Ukraine for the accident consequences elimination and population social protection assuring for the years 1986-1993 and this Programme concept for the period up to the year 2000 with a special regard of the world community participation there

  18. [The results of a cytogenetic examination of children who were victims of the accident at the Chernobyl Atomic Electric Power Station and who reside in an ecologically unfavorable region].

    Science.gov (United States)

    Frolov, V M; Baryliak, I R; Peresadin, M O; Khrypko, S V; Kuziuberdina, K M

    1994-01-01

    It has been detected that children evacuated from the zone with the elevated radioactivity level have the increased number of chromosomal and chromatid aberrations. The frequency of chromatid aberrations is 2.1-fold lower in country-side children than in children from industrial towns. Cytogenetic disturbances in immunocompetent cells correlate with the level of immune imbalance. Children born in families of liquidators of the ChNPP accident in the first year after their return from the accident zone have substantial cytogenetic and immune disturbances. Children living near sources of atmospheric pollution are characterized by the 3-4-fold increase in frequencies of chromatid anomalies and by the development of a pronounced immunodeficiency.

  19. Chernobyl accident

    International Nuclear Information System (INIS)

    Capra, D.; Facchini, U.; Gianelle, V.; Ravasini, G.; Bacci, P.

    1988-01-01

    The radioactive cloud released during the Chernobyl accident reached the Padana plain and Lombardy in the night of April 30th 1986; the cloud remained in the northern Italian skies for a few days and then disappeared either dispersed by winds and washed by rains. The evidence in atmosphere of radionuclides as Tellurium, Iodine, Cesium, was promptly observed. The intense rain, in first week of may, washed the radioactivity and fall-out contamined the land, soil, grass. The present work concerns the overall contamination of the Northern Italy territory and in particular the radioactive fall-out in the Lakes region. Samples of soil have been measured at the gamma spectroscope; a correlation is found between the radionuclides concentration in soil samples and the rain intensity, when appropriate deposition models are considered. A number of measurements has been done on the Como'lake ecosystem: sediments, plankton, fishes and the overall fall-out in the area has been investigated

  20. Performance Evaluation of Target Detection with a Near-Space Vehicle-Borne Radar in Blackout Condition.

    Science.gov (United States)

    Li, Yanpeng; Li, Xiang; Wang, Hongqiang; Deng, Bin; Qin, Yuliang

    2016-01-06

    Radar is a very important sensor in surveillance applications. Near-space vehicle-borne radar (NSVBR) is a novel installation of a radar system, which offers many benefits, like being highly suited to the remote sensing of extremely large areas, having a rapidly deployable capability and having low vulnerability to electronic countermeasures. Unfortunately, a target detection challenge arises because of complicated scenarios, such as nuclear blackout, rain attenuation, etc. In these cases, extra care is needed to evaluate the detection performance in blackout situations, since this a classical problem along with the application of an NSVBR. However, the existing evaluation measures are the probability of detection and the receiver operating curve (ROC), which cannot offer detailed information in such a complicated application. This work focuses on such requirements. We first investigate the effect of blackout on an electromagnetic wave. Performance evaluation indexes are then built: three evaluation indexes on the detection capability and two evaluation indexes on the robustness of the detection process. Simulation results show that the proposed measure will offer information on the detailed performance of detection. These measures are therefore very useful in detecting the target of interest in a remote sensing system and are helpful for both the NSVBR designers and users.

  1. Self-reported accidents

    DEFF Research Database (Denmark)

    Møller, Katrine Meltofte; Andersen, Camilla Sloth

    2016-01-01

    The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals.......The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals....

  2. Accident Assessment

    International Nuclear Information System (INIS)

    Tripputi, Ivo; Lund, Ingemar

    2002-01-01

    There is a general feeling that decommissioning is an activity involving limited risks, compared to NPP operation, and in particular risks involving the general public. This is technically confirmed by licensing analysis and evaluations, where, once the spent fuel has been removed from the plant, the radioactivity inventory available to be released to the environment is very limited. Decommissioning activities performed so far in the world have also confirmed the first assumptions and no specific issue has been identified, in this field, to justify a completely new approach. Commercial interests in international harmonization, which could drive an in-depth discussion about the bases of this approach, are weak at the moment. However, there are several reasons why a discussion in an international framework about the Safety Case for decommissioning (and, in particular, about Accident Assessment) may be considered necessary and important, and why it may show some specific and peculiar aspects. An effort for a comprehensive and systematic D and D accident safety assessment of the decommissioning process is justified. It is necessary also to explore in a holistic way the aspects of industrial safety, and develop tools for the decision-making process optimization. The expected results are the implementation of appropriate and optimized protective measures in any event and of adequate on/off-site emergency plans for optimal public and workers protection. The experience from other decommissioning projects and large-scale industrial activities is essential to balance provisions and an Operating Experience review process (specific for decommissioning) should help to focus on real issues

  3. Accident Precursor Analysis and Management: Reducing Technological Risk Through Diligence

    Science.gov (United States)

    Phimister, James R. (Editor); Bier, Vicki M. (Editor); Kunreuther, Howard C. (Editor)

    2004-01-01

    Almost every year there is at least one technological disaster that highlights the challenge of managing technological risk. On February 1, 2003, the space shuttle Columbia and her crew were lost during reentry into the atmosphere. In the summer of 2003, there was a blackout that left millions of people in the northeast United States without electricity. Forensic analyses, congressional hearings, investigations by scientific boards and panels, and journalistic and academic research have yielded a wealth of information about the events that led up to each disaster, and questions have arisen. Why were the events that led to the accident not recognized as harbingers? Why were risk-reducing steps not taken? This line of questioning is based on the assumption that signals before an accident can and should be recognized. To examine the validity of this assumption, the National Academy of Engineering (NAE) undertook the Accident Precursors Project in February 2003. The project was overseen by a committee of experts from the safety and risk-sciences communities. Rather than examining a single accident or incident, the committee decided to investigate how different organizations anticipate and assess the likelihood of accidents from accident precursors. The project culminated in a workshop held in Washington, D.C., in July 2003. This report includes the papers presented at the workshop, as well as findings and recommendations based on the workshop results and committee discussions. The papers describe precursor strategies in aviation, the chemical industry, health care, nuclear power and security operations. In addition to current practices, they also address some areas for future research.

  4. Utilities must leverage existing resources and upgrade technology to avoid future blackouts

    International Nuclear Information System (INIS)

    Masiello, R.

    2004-01-01

    The blackout of August 14, 2003 is used as the incentive to examine transmission grid reliability, expose its deficiencies in terms of technology, standards and processes used to manage reliability, and to make recommendations for leveraging existing resources and upgrading both the technology and procedures to avoid similar breakdowns in the future. It is recommended that in the area of monitoring transmission grid reliability utilities should borrow a page from the Enterprise Risk Management Practices of the financial industry by adopting a system which looks beyond the first 'credible' contingency (the current system) and examine many more 'incredible' contingencies, and underlying events that can trigger multiple contingencies, and plan for them in their operations. Utility companies are also urged to upgrade their energy management systems (EMS) technology to be able to deal with the kinds of severely depressed voltages and overloaded circuits that many grids experience today. Investment in new capabilities in control rooms, more and better communications will be essential. EMS algorithms and models must be upgraded to operate under a broader spectrum of grid conditions and to simulate the once-in-a-lifetime outage scenarios that most operators believe could never strike their utility. Expert opinion strongly suggests that as part of this process of upgrading, contingency analysis should shift from the 'N-1' model of the present to a stochastic model that considers a wider range of possible events in a probabilistic framework

  5. Utilities must leverage existing resources and upgrade technology to avoid future blackouts

    Energy Technology Data Exchange (ETDEWEB)

    Masiello, R. [KEMA Inc. Burlington, MA (United States)

    2004-06-01

    The blackout of August 14, 2003 is used as the incentive to examine transmission grid reliability, expose its deficiencies in terms of technology, standards and processes used to manage reliability, and to make recommendations for leveraging existing resources and upgrading both the technology and procedures to avoid similar breakdowns in the future. It is recommended that in the area of monitoring transmission grid reliability utilities should borrow a page from the Enterprise Risk Management Practices of the financial industry by adopting a system which looks beyond the first 'credible' contingency (the current system) and examine many more 'incredible' contingencies, and underlying events that can trigger multiple contingencies, and plan for them in their operations. Utility companies are also urged to upgrade their energy management systems (EMS) technology to be able to deal with the kinds of severely depressed voltages and overloaded circuits that many grids experience today. Investment in new capabilities in control rooms, more and better communications will be essential. EMS algorithms and models must be upgraded to operate under a broader spectrum of grid conditions and to simulate the once-in-a-lifetime outage scenarios that most operators believe could never strike their utility. Expert opinion strongly suggests that as part of this process of upgrading, contingency analysis should shift from the 'N-1' model of the present to a stochastic model that considers a wider range of possible events in a probabilistic framework.

  6. Final programmatic environmental impact statement related to decontamination and disposal of radioactive wastes resulting from March 28, 1979 accident, Three Mile Island Nuclear Station, Unit 2, Docket No. 50-320

    International Nuclear Information System (INIS)

    1981-03-01

    The appendices included in this report include the following: Comments on the Draft Programmatic Environmental Impact Statement (A-1); Commission's Statement of Policy and Notice of Intent to Prepare a Programmatic Environmental Impact Statement (B-1); 'Final Environmental Assessment for Decontamination of the Three Mile Island Unit 2 Reactor Building Atmosphere, Final NRC Staff Report,' US Nuclear Regulatory Commission, NUREG-0662, May 1980 (C-1); 'Environmental Assessment for Use of EPICOR-Il at Three Mile Island Unit 2,' US Nuclear Regulatory Commission, NUREG-0591, October 3, 1979 (D-1); Fish and Fisheries of York Haven Pond and Conowingo Pond of the Susquehanna River and Upper Chesapeake Bay (E1); Reuse of Accident Water (F-1); Engineering Considerations for Treatment of TMI-2 Accident-Generated Liquid Waste G-1); Engineering Considerations Related to Immobilization of Radioactive Wastes (H-1); Justification for Radiation Fields Used in Section 6 I-1); Economic Cost Basis (K-1); Average Individual Quarterly Dose Limits Used in Determinations of Work Force Estimates (L-1); 'Long-Term Environmental Radiation Surveillance Plan for Three Mile Island,' US Environmental Protection Agency, 1981 (M-1); Occupational Radiation Exposure during Onsite Waste Handling (N-1); Decontamination Status of Auxiliary and Fuel Handling Buildings (0-1); Chemical Systems for Decontamination of Primary System Components (P-1); Onsite Storage Facility (Q-1); Proposed Additions to Technical Specifications for TMI-2 Cleanup Program (R-1); Calculations of Discharge of Processed Accident Water to the Atmosphere (S-1); The Behavior of Sorbable Radionuclides in the Susquehanna River and Chesapeake Bay (T-1); Decommissioning of TMI-2 (U-1); Assessment of Groundwater Liquid Pathway from Leakage of Containment Water at Three Mile Island, Unit 2 (V-1); Calculation Models and Parameters Used in Estimating Doses, and Interpretation of Model Results (W-1); Contributors to the PEIS X-1); Scheduled

  7. Programmatic environmental impact statement related to decontamination and disposal of radioactive wastes resulting from March 28, 1979 accident, Three Mile Island Nuclear Station, Unit 2 (Docket No. 50-320): Draft supplement dealing with post-defueling monitored storage and subsequent cleanup

    International Nuclear Information System (INIS)

    1988-04-01

    In accordance with the National Environmental Policy Act, the Commission's implementing regulations, and its April 27, 1981,Statement of Policy, the Programmatic environmental impact statement related to decontamination and disposal of radioactive wastes resulting from March 28, 1979 accident Three Mile Island Nuclear Station, Unit 2, NUREG-0683 (PEIS) is being supplemented. This draft supplement updates the enviromental evaluation of cleanup alternatives published in the PEIS, utilizing more complete and current information. Also, the draft supplement includes a specific environmental evaluation of the licensee's recently submitted proposal for post-defueling monitored storage. The NRC staff has concluded that the licensee's proposal to place the facility in a monitored storage configuration will not significantly affect the quality of the human environment. Further, any impacts from the long-term storage of the facility are outweighed by its benefits. 63 refs., 23 figs., 65 tabs

  8. The nuclear accident risk: a territorial approach

    International Nuclear Information System (INIS)

    Ambroise, Pascal

    2011-01-01

    How many people live in the vicinity of French nuclear power stations? Recent events - notably in Japan, but also in France - highlight the urgent need to be able to predict the possible effects of a nuclear accident on surrounding territories. Here, Ambroise Pascal identifies two key criteria for such an estimation: residential density and land use. (author)

  9. Electrical Systems at Laguna Verde Nuclear Power Plant (LVNPP) after the Fukushima accident

    International Nuclear Information System (INIS)

    Lopez Jimenez, Jose Francisco

    2015-01-01

    the accident in Fukushima Daiichi, for this reason the main electrical equipment belonging to the offsite power system was changed and the electrical analysis was reviewed (such as: short-circuit, load flow, electrical stability analysis, etc.), 2) Generic Letter 2006-02 'Grid Reliability and the Impact on Plant Risk and the Operability of Offsite Power' is in process of implementation, this aims to verify that it maintains compliance with regulatory requirements which govern electrical systems and 3) the USNRC is in the process of reviewing the 10 CFR 50.63 and Regulatory Guide 1.155 'Station Blackout', once issued, CNSNS will require its implementation at Laguna Verde NPP. Based on the above, CNSNS concludes that all actions are being taken to enhance the robustness of Laguna Verde NPP's electrical systems, in order to increase their reliability, safety and operation as required in order to cope with events beyond design basis as that occurred at Fukushima Daiichi and avoid as far as possible damage to the reactor core. (authors)

  10. Accident information needs

    International Nuclear Information System (INIS)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information

  11. Accident information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-12-31

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

  12. Accident information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

  13. Camelias: a tool for the management of a complex set of γ spectrometry stations for the on-line monitoring of a reactor severe accident simulation in the framework of the Phebus FP programme

    International Nuclear Information System (INIS)

    Pantera, L.; Cornu, B.

    2003-01-01

    During a PHEBUS PF experiment, the on-line fission product release measurement is carried out by a set of seven γ stations supervised by a computer network. These stations generate about 20000 spectra for each experiment. These spectra are obtained under complex experimental conditions (high count rate, constrained nuclear environment and large diversity of configuration). The objective is to perform a quantitative γ-spectrometry analysis. The possibly complex use of self-attenuation models as well as the large volume of data to be treated led its to consider a flexible and generic method to manage the whole set of stations. The γ station calibration is carried out on-site. For each case, the developed methodology strictly uses the same approach for collecting data during the calibration phase, for calculations and for the software environment: that is the CAMELIAS application. The on-site calibration methodology is based on the use of a linear reference source allowing to reduce the 3D calculations to 2D calculations to obtain the reference efficiency function and the collimation function. Additional calculations then allow computing the efficiency function associated with the actual source. To centralize and to guarantee the coherence of the large volume of data collected during the calibration phase, we use a relational database management system. The computation of the self-attenuation coefficients associated with the actual source is performed by considering a simple model of self-attenuation along a straight line. The method, which can be generalized, is further applied to the delayed analysis of samplings and experimental devices of all the programs managed by the laboratory. (authors)

  14. Severe accident phenomena

    International Nuclear Information System (INIS)

    Jokiniemi, J.; Kilpi, K.; Lindholm, I.; Maekynen, J.; Pekkarinen, E.; Sairanen, R.; Silde, A.

    1995-02-01

    Severe accidents are nuclear reactor accidents in which the reactor core is substantially damaged. The report describes severe reactor accident phenomena and their significance for the safety of nuclear power plants. A comprehensive set of phenomena ranging from accident initiation to containment behaviour and containment integrity questions are covered. The report is based on expertise gained in the severe accident assessment projects conducted at the Technical Research Centre of Finland (VTT). (49 refs., 32 figs., 12 tabs.)

  15. Station Capacity

    DEFF Research Database (Denmark)

    Landex, Alex

    2011-01-01

    the probability of conflicts and the minimum headway times into account. The last method analyzes how optimal platform tracks are used by examining the arrival and departure pattern of the trains. The developed methods can either be used separately to analyze specific characteristics of the capacity of a station......Stations are often limiting the capacity of railway networks. This is due to extra need of tracks when trains stand still, trains turning around, and conflicting train routes. Although stations are often the capacity bottlenecks, most capacity analysis methods focus on open line capacity. Therefore...... for platform tracks and the probability that arriving trains will not get a platform track immediately at arrival. The third method is a scalable method that analyzes the conflicts in the switch zone(s). In its simplest stage, the method just analyzes the track layout while the more advanced stages also take...

  16. Use of simulators in severe accident management

    International Nuclear Information System (INIS)

    Evans, R.C.

    1994-01-01

    The U.S. nuclear utility industry is moving in a deliberate fashion through a coordinated industry severe accident working group to study and augment, where appropriate, the existing utility organizational and emergency planning structure to address accident and severe accident management. Full-scope simulators are used extensively to train licensed operators for their initial license examinations and continually thereafter in licensed operator requalification training and yearly examinations. The goal of the training (both initial and requalification) is to ensure that operators possess adequate knowledge, skills and abilities to prevent an event from progressing to core damage. The use of full-scope simulators in severe accident management training is in large part viewed by the industry as being premature. The working group study has not progressed to the point where the decision to employ full-scope simulators can be logically considered. It is not however premature to consider part-task or work station simulators as invaluable research tools to support the industry's study. These simulators could be employed, subject to limitations in the current state of knowledge regarding severe accident progression and phenomenological responses, in the validation and verification (V and V) of severe accident models or codes as they are developed. The U.S. nuclear utility industry has made substantial strides in the past 12 years in the accident prevention, mitigation and management arena. These strides are a product of the industry's preference for a logical and systematic approach to change. (orig.)

  17. Binge Drinking and the Young Brain: A Mini Review of the Neurobiological Underpinnings of Alcohol-Induced Blackout

    Directory of Open Access Journals (Sweden)

    Daniel F. Hermens

    2018-01-01

    Full Text Available Binge drinking has significant effects on memory, particularly with regards to the transfer of information to long-term storage. Partial or complete blocking of memory formation is known as blackout. Youth represents a critical period in brain development that is particularly vulnerable to alcohol misuse. Animal models show that the adolescent brain is more vulnerable to the acute and chronic effects of alcohol compared with the adult brain. This mini-review addresses the neurobiological underpinnings of binge drinking and associated memory loss (blackout in the adolescent and young adult period. Although the extent to which there are pre-existing versus alcohol-induced neurobiological changes remains unclear, it is likely that repetitive binge drinking in youth has detrimental effects on cognitive and social functioning. Given its role in learning and memory, the hippocampus is a critical region with neuroimaging research showing notable changes in this structure associated with alcohol misuse in young people. There is a great need for earlier identification of biological markers associated with alcohol-related brain damage. As a means to assess in vivo neurochemistry, magnetic resonance spectroscopy (MRS has emerged as a particularly promising technique since changes in neurometabolites often precede gross structural changes. Thus, the current paper addresses how MRS biomarkers of neurotransmission (glutamate, GABA and oxidative stress (indexed by depleted glutathione in the hippocampal region of young binge drinkers may underlie propensity for blackouts and other memory impairments. MRS biomarkers may have particular utility in determining the acute versus longer-term effects of binge drinking in young people.

  18. Blackout risk prevention in a smart grid based flexible optimal strategy using Grey Wolf-pattern search algorithms

    International Nuclear Information System (INIS)

    Mahdad, Belkacem; Srairi, K.

    2015-01-01

    Highlights: • A generalized optimal security power system planning strategy for blackout risk prevention is proposed. • A Grey Wolf Optimizer dynamically coordinated with Pattern Search algorithm is proposed. • A useful optimized database dynamically generated considering margin loading stability under severe faults. • The robustness and feasibility of the proposed strategy is validated in the standard IEEE 30 Bus system. • The proposed planning strategy will be useful for power system protection coordination and control. - Abstract: Developing a flexible and reliable power system planning strategy under critical situations is of great importance to experts and industrials to minimize the probability of blackouts occurrence. This paper introduces the first stage of this practical strategy by the application of Grey Wolf Optimizer coordinated with pattern search algorithm for solving the security smart grid power system management under critical situations. The main objective of this proposed planning strategy is to prevent the practical power system against blackout due to the apparition of faults in generating units or important transmission lines. At the first stage the system is pushed to its margin stability limit, the critical loads shedding are selected using voltage stability index. In the second stage the generator control variables, the reactive power of shunt and dynamic compensators are adjusted in coordination with minimization the active and reactive power at critical loads to maintain the system at security state to ensure service continuity. The feasibility and efficiency of the proposed strategy is applied to IEEE 30-Bus test system. Results are promising and prove the practical efficiency of the proposed strategy to ensure system security under critical situations

  19. Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Fynan, Douglas A.; Ahn, Kwang-Il, E-mail: kiahn@kaeri.re.kr

    2016-12-15

    Highlights: • Pressure drop-flow rate curves for superheated steam in U-tubes were generated. • Forward flow of hot steam is favored in the longer and taller U-tubes. • Reverse flow of cold steam is favored in short U-tubes. • Steam generator U-tube bundle geometry and tube diameter are important. • Need for correlation development for natural convention heat transfer coefficient. - Abstract: Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the s