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Sample records for coolant accident results

  1. Results and Observations of the Integral Loss-of-coolant Accident Test with Surface Modified Claddings

    Energy Technology Data Exchange (ETDEWEB)

    Park, Dong Jun; Jung, Yang Il; Park, Jung Hwan; Kim, Hyun Gil; Yang, Jae Ho; Koo, Yang Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this study, integral loss-of-coolant accident (LOCA) test was carried for comprehensive understanding of phenomena such as ballooning, burst failures, and oxidation for the ATF cladding during a LOCA scenario. In this section some of the experimental procedure and technical details of apparatus are described. Highlight data obtained from simulated LOCA test is also presented. Cracks can be initiated at this brittle burst tip and will propagate rapidly though the ballooned region. Therefore, the flexural strength of the ruptured tubes mainly depends on the thickness of the load bearing Zr metal at the opposite side to the rupture opening. To improve the reliability and safety of existing Zr alloy fuel cladding under LWR accident conditions, a high temperature oxidation resistant layer was coated onto the surface of Zr alloy samples using various coating techniques. The rupture temperature of the coated tube was higher than that of the uncoated cladding.

  2. Loss of Coolant Accident Analysis Methodology for SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Bae, K. H.; Lee, G. H.; Yang, S. H.; Yoon, H. Y.; Kim, S. H.; Kim, H. C

    2006-02-15

    The analysis methodology on the Loss-of-coolant accidents (LOCA's) for SMART-P is described in this report. SMART-P is an advanced integral type PWR producing a maximum thermal power of 65.5 MW with metallic fuel. LOCA's are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. Since SMART-P contains the major primary circuit components in a single Reactor Pressure Vessel (RPV), the possibility of a large break LOCA (LBLOCA) is inherently eliminated and only the small break LOCA is postulated. This report describes the outline and acceptance criteria of small break LOCA (SBLOCA) for SMART-P and documents the conservative analytical model and method and the analysis results using the TASS/SMR code. This analysis method is applied in the SBLOCA analysis performed for the ECCS performance evaluation which is described in the section 6.3.3 of the safety analysis report. The prediction results of SBLOCA analysis model of SMART-P for the break flow, system's pressure and temperature distributions, reactor coolant distribution, single and two-phase natural circulation phenomena, and the time of major sequence of events, etc. should be compared and verified with the applicable separate and integral effects test results. Also, it is required to set-up the feasible acceptance criteria applicable to the metallic fueled integral reactor of SMART-P. The analysis methodology for the SBLOCA described in this report will be further developed and validated as the design and licensing status of SMART-P evolves.

  3. Analysis of Loss-of-Coolant Accidents in the NBSR

    Energy Technology Data Exchange (ETDEWEB)

    Baek J. S.; Cheng L.; Diamond, D.

    2014-05-23

    This report documents calculations of the fuel cladding temperature during loss-of-coolant accidents in the NBSR. The probability of a pipe failure is small and procedures exist to minimize the loss of water and assure emergency cooling water flows into the reactor core during such an event. Analysis in the past has shown that the emergency cooling water would provide adequate cooling if the water filled the flow channels within the fuel elements. The present analysis is to determine if there is adequate cooling if the water drains from the flow channels. Based on photographs of how the emergency water flows into the fuel elements from the distribution pan, it can be assumed that this water does not distribute uniformly across the flow channels but rather results in a liquid film flowing downward on the inside of one of the side plates in each fuel element and only wets the edges of the fuel plates. An analysis of guillotine breaks shows the cladding temperature remains below the blister temperature in fuel plates in the upper section of the fuel element. In the lower section, the fuel plates are also cooled by water outside the element that is present due to the hold-up pan and temperatures are lower than in the upper section. For small breaks, the simulation results show that the fuel elements are always cooled on the outside even in the upper section and the cladding temperature cannot be higher than the blister temperature. The above results are predicated on assumptions that are examined in the study to see their influence on fuel temperature.

  4. Cladding embrittlement during postulated loss-of-coolant accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-02-01

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

  6. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  7. Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree.

  8. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    Energy Technology Data Exchange (ETDEWEB)

    Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  9. Analysis of a hypothetical loss of coolant accident in a Konvoi type NPP by GASFLOW and COCOSYS

    Energy Technology Data Exchange (ETDEWEB)

    Benz, Stefan; Royl, Peter [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Band, Sebastian [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

    2013-07-01

    The 3D computational fluid dynamics code GASFLOW and the German containment code system COCOSYS, which is based on a lumped-parameter approach, are used to simulate the hydrogen-air-steam distribution and hydrogen mitigation in a Konvoi type nuclear power plant in a postulated hypothetical core melt accident. A break in a coolant loop and the subsequent loss of the coolant causes a strong heat-up of the core. As a consequence hydrogen is produced by oxidation of cladding tubes. The residual steam and the produced hydrogen are released into the containment through the break in the coolant loop. Without suitable counter measures, sensitive mixtures can build up with a combustion potential which could threaten the integrity of the containment. A model of a Konvoi type nuclear power plant which is equipped with passive autocatalytic recombiners is used to simulate such accident scenario. COCOSYS allows comprehensive simulation of all relevant processes of severe accidents, whereas GASFLOW is primarily designed to simulate the distribution of steam and hydrogen within the containment. This paper presents the comparison of GASFLOW and COCOSYS simulation results for the in-vessel phase of the selected accident. (orig.)

  10. Prototypic Thermal-Hydraulic Experiment in NRU to Simulate Loss-of-Coolant Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Mohr, C. L.; Hesson, G. M.; Russcher, G. E.; Marsh, R. K.; King, L. L.; Wildung, N. J.; Rausch, W. N.; Bennett, W. D.

    1981-04-01

    Quick-look test results are reported for the initial test series of the Loss-of-Coolant Accident (LOCA) Simulation in the National Research Universal {NRU) test program, conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Commission (NRC). This test was devoted to evaluating the thermal-hydraulic characteristics of a full-length light water reactor (LWR) fuel bundle during the heatup, reflood, and quench phases of a LOCA. Experimental results from 28 tests cover reflood rates of 0.74 in./sec to 11 in./sec and delay times to initiate reflood of 3 sec to 66 sec. The results indicate that current analysis methods can predict peak temperatures within 10% and measured quench times for the bundle were significantly less than predicted. For reflood rates of 1 in./sec where long quench times were predicted (>2000 sec}, measured quench times of 200 sec were found.

  11. Prototypic Thermal-Hydraulic Experiment in NRU to Simulate Loss-of-Coolant Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Mohr, C. L.; Hesson, G. M.; Russcher, G. E.; Marsh, R. K.; King, L. L.; Wildung, N. J.; Rausch, W. N.; Bennett, W. D.

    1981-04-01

    Quick-look test results are reported for the initial test series of the Loss-of-Coolant Accident (LOCA) Simulation in the National Research Universal {NRU) test program, conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Commission (NRC). This test was devoted to evaluating the thermal-hydraulic characteristics of a full-length light water reactor (LWR) fuel bundle during the heatup, reflood, and quench phases of a LOCA. Experimental results from 28 tests cover reflood rates of 0.74 in./sec to 11 in./sec and delay times to initiate reflood of 3 sec to 66 sec. The results indicate that current analysis methods can predict peak temperatures within 10% and measured quench times for the bundle were significantly less than predicted. For reflood rates of 1 in./sec where long quench times were predicted (>2000 sec}, measured quench times of 200 sec were found.

  12. Validation of advanced NSSS simulator model for loss-of-coolant accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kao, S.P.; Chang, S.K.; Huang, H.C. [Nuclear Training Branch, Northeast Utilities, Waterford, CT (United States)

    1995-09-01

    The replacement of the NSSS (Nuclear Steam Supply System) model on the Millstone 2 full-scope simulator has significantly increased its fidelity to simulate adverse conditions in the RCS. The new simulator NSSS model is a real-time derivative of the Nuclear Plant Analyzer by ABB. The thermal-hydraulic model is a five-equation, non-homogeneous model for water, steam, and non-condensible gases. The neutronic model is a three-dimensional nodal diffusion model. In order to certify the new NSSS model for operator training, an extensive validation effort has been performed by benchmarking the model performance against RELAP5/MOD2. This paper presents the validation results for the cases of small-and large-break loss-of-coolant accidents (LOCA). Detailed comparisons in the phenomena of reflux-condensation, phase separation, and two-phase natural circulation are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  14. Investigation on two-phase critical flow for loss-of-coolant accident of pressurized water reactor

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    The previous investigations were mainly conducted under the condition of low pressure,however,the steam-water specific volume and the interphase evaporation rate in high pressure are much different from those in low pressure,Therefore,the new experimental and theoretical investigation are performed in Xi'an Jiaotong University.The investigation results could be directly applied to the analysis of loss-of -coolant accident for pressurized water reacor.The system transition characteristics of cold leg and hot leg break loss-of -coolant tests are described for convective circulation test loop.Two types of loss-of-coolant accident are identified for :hot leg” break,while three types for “cold leg”break and the effect parameters on the break geometries.Tests indicate that the mass flow rate with convergent-divergent nozzle reaches the maximum value among the different break sections at the same inlet fluid condition because the fluid separation does not occur.A wall surface cavity nucleation model is developed for prediction of the critical mass flow rate with water flowing in convergentdivergent nozzles.

  15. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  16. Simulating experimental investigation on the safety of nuclear heating reactor in loss-of-coolant accidents

    Science.gov (United States)

    Xu, Zhanjie

    1996-12-01

    The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology (INET) of Tsinghua University of China in 1989. Its main loop is a thermal-hydraulic system with natural circulation. This paper studies the safety of NHR under the condition of loss-of-coolant accidents (LOCAs) by means of simulant experiments. First, the background and necessity of the experiments are presented, then the experimental system, including the thermal-hydraulic system and the data collection system, and similarity criteria are introduced. Up to now, the discharge experiments with the residual heating power (20% rated heating power) have been carried out on the experimental system. The system parameters including circulation flow rate, system pressure, system temperature, void fraction, discharge mass and so on have been recorded and analyzed. Based on the results of the experiments, the conclusions are shown as follos: on the whole, the reactor is safe under the condition of LOCAs, but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

  17. Interfacing systems LOCAs (Loss of Coolant Accidents) at boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency (CDF).

  18. Aging and loss-of-coolant accident (LOCA) testing of electrical connections

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, C.F. [Sandia National Labs., Albuquerque, NM (United States)

    1998-01-01

    This report presents the results of an experimental program to determine the aging and loss-of-coolant accident (LOCA) behavior of electrical connections in order to obtain an initial scoping of their performance. Ten types of connections commonly used in nuclear power plants were tested. These included 3 types of conduit seals, 2 types of cable-to-device connectors, 3 types of cable-to-cable connectors, and 2 types of in-line splices. The connections were aged for 6 months under simultaneous thermal (99 C) and radiation (46 Gy/hr) conditions. A simulated LOCA consisting of sequential high dose-rate irradiation (3 kGy/hr) and high-temperature steam exposures followed the aging. Connection functionality was monitored using insulation resistance measurements during the aging and LOCA exposures. Because only 5 of the 10 connection types passed a post-LOCA, submerged dielectric withstand test, further detailed investigation of electrical connections and the effects of cable jacket integrity on the cable-connection system is warranted.

  19. An advanced method for determination of loss of coolant accident in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoodi, R. [Department of Engineering, Shahid Beheshti University, GC, Evin, Tehran (Iran, Islamic Republic of); Shahriari, M., E-mail: m-shahriari@sbu.ac.ir [Department of Engineering, Shahid Beheshti University, GC, Evin, Tehran (Iran, Islamic Republic of); Zolfaghari, A.; Minuchehr, A. [Department of Engineering, Shahid Beheshti University, GC, Evin, Tehran (Iran, Islamic Republic of)

    2011-06-15

    Highlights: > The considerations of vibration signals are introduced as a new method for determination of accidents directly by detecting of vibration signals without including signals from other components and this is the superiority of the proposed method. > FFT provides an alternate way of representing data. Instead of representing vibration signal amplitude as a function of time, the signal is represented by the amount of information which is contained at different frequencies. > The most of frequencies of structure and fluid coupled are presented in the FFT of structural response and through it the dominant frequency of excitation is obtained. > The Power Spectral Density, a measurement of energy at various frequencies is worked out. MATLAB software is used to convert signals from the time to frequency domain and to obtain PSD of signals. - Abstract: A major objective in reactor design is to provide the capability to withstand a wide range of postulated events without exceeding specified safety limits. Assessment of the consequence of hypothetical loss of coolant accident (LOCA) in primary circuit is an essential element to address fulfilment of acceptance criteria. In addition, finding the position of rupture, one could manage accident in a right direction. In this work, the transient vibration signal from a pipe rupture is used to determine the position of LOCA. A finite element formulation (Galerkin Method) is implemented to include the effect of fluid-structure interaction (FSI). The coupled equations of fluid motion and pipe displacement are solved. The obtained results are in good agreement with published data. Fast Fourier transform (FFT) provides an alternate way of representing data. Instead of representing vibration signal amplitude as a function of time, the signal is represented by the amount of information, which is contained at different frequencies. The most of frequencies of structure and fluid coupled are presented in the FFT of structural

  20. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Heames, T.J. (Science Applications International Corp., Albuquerque, NM (USA)); Williams, D.A.; Johns, N.A.; Chown, N.M. (UKAEA Atomic Energy Establishment, Winfrith (UK)); Bixler, N.E.; Grimley, A.J. (Sandia National Labs., Albuquerque, NM (USA)); Wheatley, C.J. (UKAEA Safety and Reliability Directorate, Culcheth (UK))

    1990-10-01

    This document provides a description of a model of the radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident. This document serves as the user's manual for the computer code called VICTORIA, based upon the model. The VICTORIA code predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating. This document provides a detailed description of each part of the implementation of the model into VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided. The VICTORIA code was developed upon a CRAY-XMP at Sandia National Laboratories in the USA and a CRAY-2 and various SUN workstations at the Winfrith Technology Centre in England. 60 refs.

  1. ROSA-III base test series for a large break loss-of-coolant accident in a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tasaka, K.; Abe, N.; Anoda, Y.; Koizumi, Y.; Shiba, M.

    1982-05-01

    The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. It is confirmed from the experimental results obtained so far that the ROSA-III test facility can simulate major aspects of a BWR LOCA, such as boiling transition by lowering of the mixture level in the core, rewetting by the lower plenum flashing, and final quenching by the ECCS. The overall agreement between the calculated results by the RELAP5/ MOD0 code and the experimental results is good; however, the calculated lower plenum flashing rewetted the whole core and the calculated cladding temperature considerably underpredicts the measured value at the upper part of the core.

  2. Models and numerical methods for the simulation of loss-of-coolant accidents in nuclear reactors

    Science.gov (United States)

    Seguin, Nicolas

    2014-05-01

    In view of the simulation of the water flows in pressurized water reactors (PWR), many models are available in the literature and their complexity deeply depends on the required accuracy, see for instance [1]. The loss-of-coolant accident (LOCA) may appear when a pipe is broken through. The coolant is composed by light water in its liquid form at very high temperature and pressure (around 300 °C and 155 bar), it then flashes and becomes instantaneously vapor in case of LOCA. A front of liquid/vapor phase transition appears in the pipes and may propagate towards the critical parts of the PWR. It is crucial to propose accurate models for the whole phenomenon, but also sufficiently robust to obtain relevant numerical results. Due to the application we have in mind, a complete description of the two-phase flow (with all the bubbles, droplets, interfaces…) is out of reach and irrelevant. We investigate averaged models, based on the use of void fractions for each phase, which represent the probability of presence of a phase at a given position and at a given time. The most accurate averaged model, based on the so-called Baer-Nunziato model, describes separately each phase by its own density, velocity and pressure. The two phases are coupled by non-conservative terms due to gradients of the void fractions and by source terms for mechanical relaxation, drag force and mass transfer. With appropriate closure laws, it has been proved [2] that this model complies with all the expected physical requirements: positivity of densities and temperatures, maximum principle for the void fraction, conservation of the mixture quantities, decrease of the global entropy… On the basis of this model, it is possible to derive simpler models, which can be used where the flow is still, see [3]. From the numerical point of view, we develop new Finite Volume schemes in [4], which also satisfy the requirements mentioned above. Since they are based on a partial linearization of the physical

  3. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  4. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  5. Investigation of loss of coolant accidents in pressurized water reactors using the ''Dynamic Best-Estimate Safety Analysis'' (DYBESA) method for considering of uncertainties in TRACE

    Energy Technology Data Exchange (ETDEWEB)

    Sporn, Michael; Hurtado, Antonio [Technische Univ. Dresden (Germany)

    2016-02-15

    Loss of coolant accident must take uncertainties with potentially strong effects on the accident sequence prediction into account. For example, uncertainties in computational model input parameters resulting from varying geometry and material data due to manufacturing tolerances or unavailable measurements should be considered. The uncertainties of physical models used by the software program are also significant. In this paper, use of the ''Dynamic Best-Estimate Safety Analysis'' (DYBESA) method to quantify the uncertainties in the TRACE thermal-hydraulic program is demonstrated. For demonstration purposes loss of coolant accidents with breaks of various types and sizes in a DN 700 reactor coolant pipe are used as an example Application.

  6. Long-term aging and loss-of-coolant accident (LOCA) testing of electrical cables

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, C.F.; Gauthier, G.; Carlin, F. [and others

    1996-10-01

    Experiments were performed to assess the aging degradation and loss-of-coolant accident (LOCA) behavior of electrical cables subjected to long-term aging exposures. Four different cable types were tested in both the U.S. and France: (1) U.S. 2 conductor with ethylene propylene rubber (EPR) insulation and a Hypalon jacket. (2) U.S. 3 conductor with cross-linked polyethylene (XLPE) insulation and a Hypalon jacket. (3) French 3 conductor with EPR insulation and a Hypalon jacket. (4) French coaxial with polyethylene (PE) insulation and a PE jacket. The data represent up to 5 years of simultaneous aging where the cables were exposed to identical aging radiation doses at either 40{degrees}C or 70{degrees}C; however, the dose rate used for the aging irradiation was varied over a wide range (2-100 Gy/hr). Aging was followed by exposure to simulated French LOCA conditions. Several mechanical, electrical, and physical-chemical condition monitoring techniques were used to investigate the degradation behavior of the cables. All the cables, except for the French PE cable, performed acceptably during the aging and LOCA simulations. In general, cable degradation at a given dose was highest for the lowest dose rate, and the amount of degradation decreased as the dose rate was increased.

  7. Large-break loss-of-coolant accident phenomena identification and ranking table (PIRT) for the advanced Candu reactor

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.; Snell, V.G.; Sills, H.E.; Langman, V.J.; Boyack, B. [Atomic Energy of Canada Ltd (Canada)

    2004-07-01

    The Advanced Candu Reactor (ACR) is an evolutionary advancement of the current Candu-6 reactor, aimed at producing electrical power for a capital cost and unit-energy cost significantly less than that of current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by heavy water moderator, as with all Candu reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper is focused on the double-ended guillotine critical inlet header break (CRIHB) loss-of-coolant accident (LOCA) in an ACR reactor, which is considered as a large break LOCA. Large Break LOCA in water-cooled reactors has been used historically as a design basis event by regulators, and it has attracted a very large share of safety analysis and regulatory review. The LBLOCA event covers a wide range of system behaviours and fundamental phenomena. The Phenomena Identification and Ranking Table (PIRT) for LBLOCA therefore provides a good understanding of many of the safety characteristics of the ACR design. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the LOCA phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the final PIRT summary table. (authors)

  8. International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA

    Directory of Open Access Journals (Sweden)

    N. Aksan

    2008-01-01

    Full Text Available Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs. These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2 and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects

  9. Simulating Experimental Investigation on the Safety of Nuclear Heating Reactor in Loss—of —Coolant Accidents

    Institute of Scientific and Technical Information of China (English)

    ZhanjieXu

    1996-01-01

    The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology(INET) of Tisinghuan University of CHina in 1989,Its main loop is a thermal-hydraulic system with natural circulation.This paper studies the safety of NHR under the condition of loss-of -coolant accidents(LOCAs) by means of simulant experiments.First,the Background and necessity of the experiments are presented.then the experimental system,including the thermal-hydraulic system and the data collection system,and similarity criteria are introduced.Up to now ,the discharge experiments with the residual heating power(20% rated heating power)have been carried out on the experimental system,The system prameters including circulation flow rate,system pressure,system temperature,void fraction,discharge mass and so on have been recorded and analyzed.Based on the results of the experiments,the conclusionas are shown as follos:on the whole,the reactor is safe under the condition of LOCAs,but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

  10. Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

    Energy Technology Data Exchange (ETDEWEB)

    Ruggles, A. E. [Oak Ridge National Lab., TN (United States); Tennessee Univ., Knoxville, TN (United States); Cheng, L. Y. [Brookhaven National Lab., Upton, NY (United States); Dimenna, R. A. [Westinghouse Savannah River Co., Aiken, SC (United States); Griffith, P. [Massachusetts Inst. of Tech., Cambridge, MA (United States); Wilson, G. E. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1994-06-01

    A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis.

  11. Experimental investigation of material chemical effects on emergency core cooling pump suction filter performance after loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Woon, E-mail: parkjw@dongguk.ac.k [Dongguk University, 707 Seokjang-Dong, Gyeongju, 780-714 (Korea, Republic of); Park, Byung Gi [Soonchunhyang University, Asan, Chungnam, 336-745 (Korea, Republic of); Kim, Chang Hyun [Korea Hydro and Nuclear Power Co., Ltd. 25-1, Jang-dong, Yuseong-gu, Daejeon, 305-343 (Korea, Republic of)

    2009-12-15

    Integral tests of head loss through an emergency core cooling filter screen are conducted, simulating reactor building environmental conditions for 30 days after a loss of coolant accident. A test rig with five individual loops each of whose chamber is established to test chemical product formation and measure the head loss through a sample filter. The screen area at each chamber and the amounts of reactor building materials are scaled down according to specific plant condition. A series of tests have been performed to investigate the effects of calcium-silicate, reactor building spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the filter screen is strongly affected by spray duration and the head loss increase is rapid at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKON{sup TM} and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

  12. Permeability and compression of fibrous porous media generated from dilute suspensions of fiberglass debris during a loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Saya, E-mail: sayalee@tamu.edu; Abdulsattar, Suhaeb S.; Vaghetto, Rodolfo; Hassan, Yassin A.

    2015-09-15

    Highlights: • Experimental investigation on fibrous debris buildup was conducted. • Head loss through fibrous media was recorded at different approach velocities. • A head loss model through fibrous media was proposed for high porosity (>0.99). • A compression model of fibrous media was developed. - Abstract: Permeability of fibrous porous media has been studied for decades in various engineering applications, including liquid purifications, air filters, and textiles. In nuclear engineering, fiberglass has been found to be a hazard during a Loss-of-Coolant Accident. The high energy steam jet from a break impinges on surrounding fiberglass insulation materials, producing a large amount of fibrous debris. The fibrous debris is then transported through the reactor containment and reaches the sump strainers. Accumulation of such debris on the surface of the strainers produces a fibrous bed, which is a fibrous porous medium that can undermine reactor core cooling. The present study investigated the buildup of fibrous porous media on two types of perforated plate and the pressure drop through the fibrous porous media without chemical effect. The development of the fibrous bed was visually recorded in order to correlate the pressure drop, the approach velocity, and the thickness of the fibrous porous media. The experimental results were compared to semi-theoretical models and theoretical models proposed by other researchers. Additionally, a compression model was developed to predict the thickness and the local porosity of a fibrous bed as a function of pressure.

  13. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

  14. Effect of Check Valve on Consequences of Coolant Pump Rotor Seizure Accident for EPR Reactor%止回阀对EPR反应堆主泵卡轴事故后果的影响

    Institute of Scientific and Technical Information of China (English)

    陈秋炀; 周拥辉

    2012-01-01

    分析计算欧洲先进压水堆(EPR)反应堆主泵卡轴事故,并对比在主泵出口安装止回阀和没有安装止回阀模型的卡轴事故安全分析.结果表明,在EPR主泵卡轴事故中,止回阀可增加模型堆芯进口流量约4%,有利于堆芯的冷却.止回阀可显著地提高堆芯最小偏离泡核沸腾比(DNBR),降低堆芯偏离泡核沸腾(DNB)份额,降低包壳温度约14℃.模型分析结果表明,在主泵卡轴事故工况下,主泵出口安装止回阀可更好地维持堆芯的完整性.%Counter current flow phenomenon would appear during reactor coolant pump rotor seizure accident. Present work analyzes the coolant pump rotor seizure accident for European Pressurized Reactor (EPR). The accident safety analysis results of model with check valve and without check valve are compared. It can be found that the check valve can increase the core inlet flow rate of model about 4%. The increasing of coolant flow rate is beneficial to the reactor core cooling. Check valve can increase the minimum departure from nucleate boiling ratio (DNBR), reduce the departure from nucleate boiling (DNB) fraction and the fuel rod cladding temperature about 14℃ during coolant pump rotor seizure accident. The analyses results show that the model with check valve can maintain the integrity of nuclear fuel rod effectively during reactor coolant pump rotor seizure accident.

  15. Safety Analysis Results for Cryostat Ingress Accidents in ITER

    Science.gov (United States)

    Merrill, B. J.; Cadwallader, L. C.; Petti, D. A.

    1997-06-01

    Accidents involving the ingress of air, helium, or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.

  16. Development of analytical code for core disruptive accident of metallic fuel core and evaluation of upper limitation of coolant void reactivity

    Energy Technology Data Exchange (ETDEWEB)

    Ueda, Nobuyuki [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab

    2001-06-01

    In the event of HCDA (Hypothetical Core Damage Accident) the safety of the metallic core must be ensured. Coolant void reactivity, which produces a positive reactivity effect, must be restrained to exclude 'energetic sequence' (uncontrollable power generation). As this reactivity and the core performance are in a state of confrontation, an upper limit must be determined through safety analyses. The larger the coolant void reactivity becomes, the more the core performance improves. A core dynamics analysis code named CANIS has been developed to analysis metallic fuel properties, irradiation behavior, and transient behavior in the initiating phase of HCDA. ULOF (Unprotected Loss Of Flow) events, which are considered major HCDA initiators, are analyzed with various parameters of cladding failure conditions and molten fuel dispersion velocities. The results indicate, the cladding failure condition is the most sensitive safety parameter. The coolant void reactivity of 12$ is acceptable in the condition that cladding fails at 1,000 degC. And the upper value is limited to 8$ for the conservative condition of a 1,200 degC failure criteria. (author)

  17. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Heams, T J [Science Applications International Corp., Albuquerque, NM (United States); Williams, D A; Johns, N A; Mason, A [UKAEA, Winfrith, (England); Bixler, N E; Grimley, A J [Sandia National Labs., Albuquerque, NM (United States); Wheatley, C J [UKAEA, Culcheth (England); Dickson, L W [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Osborn-Lee, I [Oak Ridge National Lab., TN (United States); Domagala, P; Zawadzki, S; Rest, J [Argonne National Lab., IL (United States); Alexander, C A [Battelle, Columbus, OH (United States); Lee, R Y [Nuclear Regulatory Commission, Washington, DC (United States)

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided.

  18. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  19. Strategies for reactor safety: Preventing loss of coolant accidents. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, B.O.Y. [RSA Technoligies, Vista (United States)

    1997-12-01

    This final report on the NKS/RAK-1.2 summarizes the main features of the PIFRAP PC-program and its intended implementation. Regardless of the preferred technical approach to LOCA frequency estimation, the analysis approach must include recognition of the following technical issues: (a) Degradation and failure mechanisms potentially affecting piping systems within the reactor coolant pressure boundary (RCPB) and the potential consequences; (b) In-service inspection practices and how they influence piping reliability; and (c) The service experience with piping systems. The report consists of six sections and one appendix. A Nordic perspective on LOCA and nuclear safety is given. It includes summaries of results from research in material sciences and current regulatory philosophies regarding piping reliability. A summary of the LOCA concept is applied in Nordic PSA studies. It includes a discussion on deterministic and probabilistic views on LOCA. The R and D on piping reliability by SKI and the PIFRAP model is summarized. Next, Section 6 presents conclusion and recommendations. Finally, Appendix A contains a list of abbreviations and acronyms, together with a glossary of technical terms. (EG) 16 refs.

  20. Neutron Imaging Investigations of the Secondary Hydriding of Nuclear Fuel Cladding Alloys during Loss of Coolant Accidents

    Science.gov (United States)

    Grosse, M.; Roessger, C.; Stuckert, J.; Steinbrueck, M.; Kaestner, A.; Kardjilov, N.; Schillinger, B.

    The hydrogen concentration and distribution at both sides of the burst opening of cladding tubes used in three QUENCH-LOCA simulation bundle experiments were investigated by means of neutron radiography and tomography. The quantitative correlation between the total macroscopic neutron cross-section and the atomic number density ratio between hydrogen and zirconium was determined by testing calibration specimens with known hydrogen concentrations. Hydrogen enrichments located at the end of the ballooning zone of the tested tubes were detected in the inner rods of the test bundles. Nearly all of the peripheral claddings exposed to lower temperatures do not show such enrichments. This implies that under the conditions investigated a threshold temperature exists below which no hydrogen enrichments can be formed. In order to understand the hydrogen distribution a model was developed describing the processes occurring during loss of coolant accidents after rod burst. The general shape of the hydrogen distributions with a peak each side of the ballooning region is well predicted by this model whereas the absolute concentrations are underestimated compared to the results of the neutron tomography investigations. The model was also used to discuss the influence of the alloy composition on the secondary hydrogenation. Whereas the relations for the maximal hydrogen concentrations agree well for one and the same alloy, the agreement for tests with different alloys is less satisfying, showing that material parameters such as oxidation kinetics, phase transition temperature for the zirconium oxide, and yield strength and ductility at high temperature have to be taken into account to reproduce the results of neutron imaging investigations correctly.

  1. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  2. Fusion Blanket Coolant Section Criteria, Methodology, and Results

    Energy Technology Data Exchange (ETDEWEB)

    DeMuth, J. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meier, W. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jolodosky, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frantoni, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reyes, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-02

    The focus of this LDRD was to explore potential Li alloys that would meet the tritium breeding and blanket cooling requirements but with reduced chemical reactivity, while maintaining the other attractive features of pure Li breeder/coolant. In other fusion approaches (magnetic fusion energy or MFE), 17Li- 83Pb alloy is used leveraging Pb’s ability to maintain high TBR while lowering the levels of lithium in the system. Unfortunately this alloy has a number of potential draw-backs. Due to the high Pb content, this alloy suffers from very high average density, low tritium solubility, low system energy, and produces undesirable activation products in particular polonium. The criteria considered in the selection of a tritium breeding alloy are described in the following section.

  3. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 1 – Aluminum

    Energy Technology Data Exchange (ETDEWEB)

    Howe, Kerry J., E-mail: howe@unm.edu [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Mitchell, Lana, E-mail: lmitchell@alionscience.com [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Kim, Seung-Jun, E-mail: skim@lanl.gov [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Blandford, Edward D., E-mail: edb@unm.edu [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Kee, Ernest J., E-mail: erniekee@gmail.com [South Texas Project Nuclear Operating Company, P.O. Box 270, Wadsworth, TX 77483 (United States)

    2015-10-15

    Highlights: • Trisodium phosphate (TSP) causes aluminum corrosion to cease after 24 h of exposure. • Chloride, iron, and copper have a minimal effect on the rate of aluminum corrosion when TSP is present. • Zinc can reduce the rate of aluminum corrosion when TSP is present. • Aluminum occasionally precipitates at concentrations lower than the calculated solubility for Al(OH){sub 3}. • Corrosion and solubility equations can be used to calculate the solids generated during a LOCA. - Abstract: Bench experiments were conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of aluminum from metallic aluminum surfaces under conditions representative of the containment pool following a postulated loss of coolant accident at a nuclear power generating facility. The experiments showed that TSP is capable of passivating the aluminum surface and preventing continued corrosion after about 24 h at the conditions tested. A correlation that describes the rate of corrosion including the passivation effect was developed from the bench experiments and validated with a separate set of experiments from a different test system. The saturation concentration of aluminum was shown to be well described by the solubility of amorphous aluminum hydroxide for the majority of cases, but instances have been observed when aluminum precipitates at concentrations lower than the calculated aluminum hydroxide solubility. Based on the experimental data and previous literature, an equation was developed to calculate the saturation concentration of aluminum as a function of pH and temperature under conditions representative of a loss of coolant accident (LOCA) in a TSP-buffered pressurized water reactor (PWR) containment. The corrosion equation and precipitation equation can be used in concert with each other to calculate the quantity of solids that would form as a function of time during a LOCA if the temperature and pH profiles were known.

  4. Investigation of a hydrogen mitigation system during large break loss-of-coolant accident for a two-loop pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dehjourian, Mehdi; Rahgoshay, Mohmmad; Jahanfamia, Gholamreza [Dept. of Nuclear Engineering, Science and Research Branch, Islamic Azad University of Tehran, Tehran (Iran, Islamic Republic of); Sayareh, Reza [Faculty of Electrical and Computer Engineering, Kerman Graduate University of Technology, Kerman (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

  5. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

    Energy Technology Data Exchange (ETDEWEB)

    Seeliger, André, E-mail: a.seeliger@hszg.de [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Alt, Sören; Kästner, Wolfgang; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Kryk, Holger; Harm, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany)

    2016-08-15

    Highlights: • Borated coolant supports corrosion at zinc-coated installations in PWR after LOCA. • Dissolved zinc is injected into core by ECCS during sump recirculation phase. • Corrosion products can reach and settle at further downstream components. • Corrosion products can cause head losses at spacers and influence decay heat removal. • Preventive procedures were tested at semi-technical scale facilities. - Abstract: Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized compounds, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively. First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the Hochschule Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators. As near-plant core components, a 3 × 3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module includes segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid

  6. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2016-09-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

  7. Method for calculating coolant resonance frequencies under normal and accident conditions in nuclear power plants with WWER-type pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N. (Moskovskij Ehnergeticheskij Inst. (USSR))

    1983-03-01

    Mathematical models are proposed for calculating acoustic oscillation resonance frequencies in the coolant in various components of the WWER type primary circuit (core, steam generator, pressurizer, piping). Due to the correspondence between model calculations and experimental results obtained in operating nuclear power plants, the developed models can be used for practical calculations. The possibility of calculating the eigenfrequencies of the coolant oscillation under different operating conditions leads to the interpretation of operational data, to the analysis of operational conditions, to the detection of coolant boiling in the reactor, and to design changes in order to prevent resonance oscillations within the coolant.

  8. Simulation of Containment Pressurization in a Large Break-Loss of Coolant Accident Using Single-Cell and Multicell Models and CONTAIN Code

    Directory of Open Access Journals (Sweden)

    Omid Noori-Kalkhoran

    2016-10-01

    Full Text Available Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model. In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code’s results.

  9. Simulation of containment pressurization in a large break-loss of coolant accident using single-cell and multicell models and CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Kalkahoran, Omid Noori; Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

  10. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

  11. Experimental Investigation on Small Break Loss of Coolant Accident for Direct Vessel Injection Line%DVI管小破口失水事故实验研究

    Institute of Scientific and Technical Information of China (English)

    彭传新; 张妍; 黄志刚; 昝元锋; 卓文彬; 闫晓

    2016-01-01

    在模块化小型反应堆非能动安全系统综合模拟实验装置上进行了压力容器直接注入(DVI)管小破口失水事故实验,研究了DVI管小破口失水事故过程中的热工水力现象和非能动安全系统运行特性。研究结果表明:模块化小型反应堆DVI管小破口失水事故中,非能动安全系统可对堆芯进行注水,有效导出堆芯衰变热量,保护堆芯安全。%T he small break loss of coolant accident (SBLOCA ) experiment for direct vessel injection (DVI ) line , w hich investigated the thermal‐hydraulic phenomena and the performances of passive safety system during the accident ,was performed on the passive safety system test facility for small modular reactor .The experimental results show that the passive safety system of small modular reactor can provide effective cool‐ant injection ,successful removal of core residual heat under the DVI line SBLOCA and protection to reactor core safety .

  12. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 2 – Zinc

    Energy Technology Data Exchange (ETDEWEB)

    Pease, David; LaBrier, Daniel; Ali, Amir [Department of Nuclear Engineering, University of New Mexico (United States); Blandford, Edward D., E-mail: edb@unm.edu [Department of Nuclear Engineering, University of New Mexico (United States); Howe, Kerry J. [Department of Civil Engineering, University of New Mexico (United States)

    2016-04-15

    Highlights: • Zinc release is limited to less than 1 mg/L in TSP-buffered solution under a variety of conditions (pH, temperature, zinc source). • Zinc release in high-temperature non-TSP-buffered environment is approximately 25 mg/L. • Long-term zinc release is controlled by passivation (without TSP) and zinc solubility (with TSP). • Precipitation and solubility of zinc phosphate limit the release of zinc. - Abstract: Bench experiments were conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of zinc from metallic zinc-bearing surfaces under conditions representative of the containment pool following a postulated loss of coolant accident (LOCA) at a nuclear power generating facility. The experiments showed that in non-buffered (acidic) environments, measurable quantities of zinc are released from zinc-bearing surfaces. Precipitation and solubility of phosphate-based corrosion products, such as zinc phosphate, limit the release of zinc from zinc-bearing surfaces. These experiments have found that under a variety of conditions, including variations of temperature, pH, and across different zinc-bearing surfaces, the release of zinc into solution is limited to <1 mg/L when phosphate is present. When phosphate is not present, zinc release is instead bounded by a markedly higher saturation limit which is a strong function of the solution temperature.

  13. Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions

    Science.gov (United States)

    Park, Dong Jun; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

    2016-12-01

    This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility.

  14. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 2: with Beam Shutdown Only

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report. This report documents the results of simulations of a Loss-of-Flow Accident (LOFA) where power is lost to all of the pumps that circulate water in the blanket region, the accelerator beam is shut off and neither the residual heat removal nor cavity flood systems operate.

  15. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  16. ROSA-III double-ended break test series for a loss-of-coolant accident in a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tasaka, K.; Anoda, Y.; Koizumi, Y.; Kumamaru, H.; Nakamura, H.; Shiba, M.; Suzuki, M.; Yonomoto, T.

    1985-01-01

    The Rig of Safety Assessment (ROSA) III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency-core-cooling-system (ECCS) tests. Experimental results obtained so far confirm that the severest single failure assumption in ECCS is the high-pressure core spray system failure even in a large-break LOCA in a BWR. The measured peak cladding temperature was well below the present safety criterion of 1473 K, even with the single failure assumption in ECCS, and the effectiveness of ECCS for core cooling during a double-ended-break LOCA has been confirmed. The overall agreement between the results calculated by the RELAP4/MOD6/U4/J3 computer code and the experimental results is good. The similarity between the ROSA-III test and a BWR LOCA has been confirmed through the comparison of calculated results for the ROSA-III facility and a BWR system.

  17. Analysis of the containment of a compact reactor PWR submitted to loss of coolant accident; Analise da contencao de um reator PWR compacto submetido a acidente de perda de refrigerante

    Energy Technology Data Exchange (ETDEWEB)

    Dutra, Alexandre de Souza; Belchior Junior, Antonio; Guimaraes, Leonam dos Santos [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2000-07-01

    In the present paper analyses were done with the computer code RELAP5/MOD2 for rising the process conditions of the containment of a compact reactor PWR of low potency, submitted to Loss of Coolant Accidents (LOCA). The main results obtained were the behavior of maximum conditions of pressure as a function of the available containment free volume. It was also studied the problem of containment sub-compartmentation, that is to say, the possibility of the rupture to happen in restricted spaces generating high sub-compartment peak pressure and, consequently, high strains on the internal structures. (author)

  18. KSTAR Severe Accident Analysis using MELCOR : Ex-vessel Coolant Pipe Break with Failure of Fusion Power Termination System

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    with the real KSTAR system. But important systems are furnished to study the effect of each system. Accident analysis is carried out after steady state is maintained for 1000 sec. As a result, radioactive material leakage is simulated with aerosol release package to compare to release guideline. In this research, follow-up study of safety analysis and simple safety analysis application in KSTAR was conducted with MELCOR.

  19. Definition of loss-of-coolant accident radiation source: summary and conclusions. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bonzon, L.L.; Lurie, N.A.; Houston, D.H.; Naber, J.A.

    1978-05-01

    The radiation energy release rates and spectra corresponding to those sources specified in USNRC Regulatory Guide 1.89 for the radiation qualification of Class 1E equipment were calculated. The effects of several parameters (some not specific in the Guide), such as reactor fuel composition, operating duration and power level, and treatment of progeny, are evaluated. The results are presented as time-dependent beta and gamma-ray energy release rates and spectra which are fundamental quantities that are not specific to a plant design but are generally applicable to any nuclear power station.

  20. Hydride precipitation, fracture and plasticity mechanisms in pure zirconium and Zircaloy-4 at temperatures typical for the postulated loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Pshenichnikov, Anton, E-mail: anton.pshenichnikov@kit.edu; Stuckert, Juri; Walter, Mario

    2016-05-15

    Highlights: • All δ-hydrides in Zr and Zircaloy-4 have basal or pyramidal types of habit planes. • Seven orientation relationships for δ-hydrides in Zr matrix were detected. • Decohesion fracture mechanism of hydrogenated Zr was investigated by fractography. - Abstract: The results of investigations of samples of zirconium and its alloy Zircaloy-4, hydrogenated at temperatures 900–1200 K (typical temperatures for loss-of-coolant accidents) are presented. The analyses, based on a range of complementary techniques (X-ray diffraction, scanning electron microscopy, electron backscatter diffraction) reveals the direct interrelation of internal structure transformation and hydride distribution with the degradation of mechanical properties. Formation of small-scale zirconium hydrides and their bulk distribution in zirconium and Zircaloy-4 were investigated. Fractographical analysis was performed on the ruptured samples tested in a tensile machine at room temperature. The already-known hydrogen embrittlement mechanisms based on hydride formation and hydrogen-enhanced decohesion and the applicability of them in the case of zirconium and its alloys is discussed.

  1. Large break loss-of-coolant accident analysis for China Qinshan-2 nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yong; Ban, Chang Hwan; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Wang, Rongzhong; Yu, Hongxing [Nuclear Power Institute of China, Chengdu, SC (China)

    1994-12-01

    Large break LOCA analysis for China Qinshan-2 nuclear power plant has been performed using realistic evaluation model which has been being developed by KAERI. RELAP5/MOD3/KAERI code, which is a modified version of RELAP5/MOD3, is coupled with CONTEMPT4/MOD5 and is used as a best estimate code to predict the thermal hydraulic behavior of the system. PCT uncertainty which stems from code uncertainty, plant application uncertainty, scaling uncertainty and PCT bias are discussed. Among them, plant application uncertainty is described in detail. The licensing PCT is calculated by adding all the uncertainties to the best-estimate PCT. The result indicates the Qinshan-2 nuclear power plant has at least 37 deg C safety margin for large break LOCA. (Author) 10 refs., 47 figs., 14 tabs.

  2. A model for the analysis of loss of decay heat removal during loss of coolant accident in MTR pool type research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bousbia-salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2, 56126 Pisa (Italy)]. E-mail: b.salah@ing.unipi.it; Meftah, Brahim [Division Reacteur - Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA - Algiers (Algeria); Hamidouche, Tewfik [Laboratoire des Analyses de Surete, Centre de Recherche Nucleaire d' Alger (CRNA), 02 Boulevard Frantz Fanon, B.P. 399, 16000 Algiers (Algeria)]. E-mail: thamidouche@comena-dz.org; Si-Ahmed, El Khider [Laboratoire des Ecoulements Polyhpasiques, Universite des Sciences et de la Technologie d' Alger, Algiers (Algeria)

    2006-03-15

    During a loss of coolant accident leading to total emptying of the reactor pool, the decay heat could be removed through air natural convection. However, under partial pool emptying the core is partially submerged and the coolant circulation inside the fuel element could no more be possible. Under such conditions, a core overheat takes place, and the thermal energy is essentially diffused from the core to its periphery by combined thermal radiation and conduction. In order to predict fuel element temperature evolution under such conditions a mathematical model is performed. The model is based on a 3D geometry and takes into account a variety of core configurations including fuel elements (standard and control), reflector elements and grid plates. The homogeneous flow model is used and the fluid conservation equations are solved using a semi-implicit finite difference method. Preliminary tests of the developed model were made by considering a series of hypothetical accidents. In the current framework a loss of decay heat removal accidents in the IAEA benchmark open pool MTR-type research reactor is considered. It is shown that in the case of a low core immersion height no water boiling is observed and the fuel surface temperature rise remains below the melting point of the aluminium cladding.

  3. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Directory of Open Access Journals (Sweden)

    Seon Oh Yu

    2017-08-01

    Full Text Available The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

  4. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 4 – Integrated chemical effects testing

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Amir; LaBrier, Daniel [Department of Nuclear Engineering, University of New Mexico (United States); Blandford, Edward, E-mail: edb@unm.edu [Department of Nuclear Engineering, University of New Mexico (United States); Howe, Kerry [Department of Civil Engineering, University of New Mexico (United States)

    2016-04-15

    Highlights: • Integrated test explored the material release of a postulated large break LOCA. • Aluminum concentration was very low (<0.1 mg/L) throughout the test duration. • Zinc concentration was low (<1 mg/L) in TSP-buffered system. • Calcium release showed two distinguished release zones: prompt and meta-stable. • Copper and iron has no distinguishable concentration up to first 24 h of testing. - Abstract: This paper presents the results of an integrated chemical effects experiment executed under conditions representative of the containment pool following a postulated loss of coolant accident (LOCA) at the Vogtle nuclear power plant, operated by the Southern Nuclear Operating Company (SNOC). This test was conducted for closure of a series of bench scale experiments conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of aluminum (Howe et al., 2015) and zinc (Pease et al., 2015) from metallic surfaces, and calcium from NUKON fiberglass insulation (Olson et al., 2015) . The integrated test was performed in the Corrosion/Chemical Head Loss Experimental (CHLE) facility with representative amounts of zinc, aluminum, carbon steel, copper, NUKON fiberglass, and latent debris. The test was conducted using borated TSP-buffered solution under a post-LOCA prototypical temperature profile lasting for 30 days. The results presented in this article demonstrate trends for zinc, aluminum, and calcium release that are consistent with separate bench scale testing and previous integrated tests under TSP conditions. The release rate and maximum concentrations of the released materials were slightly different than the separate effect testing as a result of different experimental conditions (temperature, surface area-to-water volume ratio) and/or the presence of other metals and chemicals in the integrated test. Samples of metal coupons and fiberglass were selected for analysis using Scanning Electron Microscopy

  5. Exact results for car accidents in a traffic model

    Science.gov (United States)

    Huang, Ding-wei

    1998-07-01

    Within the framework of a recent model for car accidents on single-lane highway traffic, we study analytically the probability of the occurrence of car accidents. Exact results are obtained. Various scaling behaviours are observed. The linear dependence of the occurrence of car accidents on density is understood as the dominance of a single velocity in the distribution.

  6. Health effects resulting from the Chernobyl accident.

    Science.gov (United States)

    Sumner, David

    2007-01-01

    This article reviews the health effects of the Chernobyl accident. The clearest effect to be seen to date is the dramatic increase in thyroid cancer in children. The evidence for increased leukaemia is less clear, but there are indications of increased leukaemia incidence in Russian clean-up workers. There is also evidence of increases in breast cancer, cataract and cardiovascular disease. However, to date the largest public health problem caused by the accident is the mental health impact.

  7. An experimental study of the corrosion and precipitation of aluminum in the presence of trisodium phosphate buffer following a loss of coolant accident (LOCA) scenario

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Jun [Department of Nuclear Engineering, University of New Mexico (United States); Howe, Kerry J. [Department of Civil Engineering, University of New Mexico (United States); Leavitt, Janet J. [Department of Civil Engineering, University of New Mexico (United States); Alion Science and Technology (United States); Hammond, Kyle; Mitchell, Lana [Department of Civil Engineering, University of New Mexico (United States); Kee, Ernie [South Texas Project Nuclear Operating Company (STPNOC) (United States); Blandford, Edward D., E-mail: edb@unm.edu [Department of Nuclear Engineering, University of New Mexico (United States)

    2015-02-15

    Highlights: • Experimental head loss testing was conducted by aggressively promoting corrosion in loss of coolant accidents. • Blender-processed debris beds have higher head loss but tend to be less reproducible than NEI-processed debris beds. • Precipitation was observed from aluminum concentration and turbidity measurements. • Precipitation results were compared to predictions from Visual MINTEQ. - Abstract: This paper presents the results of an integrated chemical effects experiment of head loss across the sump pump screen with fibrous debris bed over a non-prototypical 10-day post-LOCA incident window. The corrosion head loss experiments (CHLE) is a reduced scaled integral effects testing facility built at the University of New Mexico (UNM) to investigate potential chemical effects on head loss across prepared fibrous debris beds. The results in this paper come from two integral effect tests performed at UNM in order to determine the chemical effects on head loss induced by a zinc source effect and an aluminum precipitation effect (T3: without Zn source case, T4: with Zn source case in containment). The tests were performed with a large surface area of aluminum coupons in the testing facility for an extended period of elevated temperature to accelerate corrosion above that expected under prototypical conditions. These conditions were sufficient to force aluminum precipitation to occur and induce the onset of chemical effects on debris bed head loss. The head loss behavior on two different types of fiber debris beds (blender-processed and NEI-processed debris bed) was evaluated in this study. It was found that the blender-processed bed is much more sensitive in filtering than the NEI-processed bed and consequently had a much higher head loss value across the beds. Aluminum precipitation was observed, with aluminum concentration and turbidity measurements, to form starting on day 7 in Test T3 and on day 6 in Test T4. The onset of aluminum precipitation

  8. Consideration on hydrogen explosion scenario in APR 1400 containment building during small breakup loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kweonha, E-mail: khpark@kmou.ac.kr [Division of Mechanical & Energy Systems Engineering, Korea Maritime University, Dongsam-dong, Yeongdo-gu, Busan 606-791 (Korea, Republic of); Khor, Chong Lee, E-mail: itachi_829@hotmail.com [Department of Mechanical Engineering, Korea Maritime University, Dongsam-dong, Yeongdo-gu, Busan 606-791 (Korea, Republic of)

    2015-11-15

    Highlights: • Hydrogen behavior in the containment building of APR1400 nuclear plant up to 15 h after the failure happened. • The risk of hydrogen explosion largely depends on the combination of air, hydrogen and steam in the containment. • Hydrogen explosion risk at different locations in the containment was analyzed. - Abstract: This paper describes the analytical result of the potential risk of hydrogen gas up to 15 h after the failure takes place. The major cause of the disaster occurred in Fukushima Daiichi nuclear reactor was the detonation of accumulated hydrogen in the containment by highly increased reactor core temperatures after the failure of the emergency cooling system. The hydrogen risk should be considered in severe accident strategies in current and future NPPs. A hydrogen explosion scenario is proposed. Hydrogen is accumulated on top of the dome during the hydrogen release period. At this point, there are no risk of explosion due to the steam that resides in upper part of the dome. As the hydrogen concentration increase, substantial amount of steams are released. Subsequently, hydrogen is forced into the lower part of the building with high air density—small explosion and dormant steam condensation phase are possible. The light hydrogen rises up slowly with air, gathering on top of the building with high air density. Massive hydrogen explosion is anticipated upon ignition at this stage.

  9. The noncondensable gas effects on loss-of-coolant accident steam condensation loads in boiling water reactor pressure suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    Kukita, Y.; Namatame, K.; Shiba, M.; Takeshita, I.

    1983-11-01

    The noncondensable gas effects on the loss-ofcoolant-accident-induced steam condensation loads in the boiling water reactor pressure suppression pool have been investigated with regard to experimental data obtained from a large-scale multivent test program. Previous studies have noted that the presence of the noncondensable gas (air), which initially fills the containment drywell space, stabilizes the direct-contact condensation in the pressure suppression pool and hampers onset of the chugging phenomenon, which induces most significant steam condensation load onto the pool boundary. This was found to be true for the tests with relatively small-break diameters, where the maximum steam mass fluxes in the vent pipe were lower than the upper threshold value for the onset of chugging. However, in the tests with the maximum vent steam mass fluxes moderately higher than the chugging upper threshold value, early depletion of the noncondensable gas tended to result in significant stabilization of steam condensation accompanied by an excursion of temperature of pool water surrounding the vent pipe outlets, which led to a delayed onset of chugging. Due to this combined influence of the noncondensable gas and nonuniform pool temperature, and due to dependence of magnitude of chugging load on the vent steam mass flux, the peak magnitude of the steam condensation load appearing in a blowdown can be very sensitive to the initial and break conditions.

  10. Severe accident progression perspectives for Mark I containments based on the IPE results

    Energy Technology Data Exchange (ETDEWEB)

    Lin, C.C.; Lehner, J.R.; Pratt, W.T. [Brookhaven National Lab., Upton, NY (United States); Drouin, M. [Nuclear Regulatory Commission, N. Bethesda, MD (United States)

    1995-12-31

    Based on level 2 analyses in IPE (Individual Plant Examination) submittals accident progression, perspectives were obtained for all containment types. These perspectives consisted of insights on containment failure modes, releases therein, and factors responsible for the results. To illustrate the types of perspectives acquired on severe accident progresssion, insights obtained for (BWR) Mark I containments are discussed here. Mark I containments have high strength but small volumes and rely on pressure suppression pools to condense steam released from the reactor coolant system during an accident. Accidents causing structural failure of the drywell shortly after the core debris melts through the reactor vessel were found to be dominant contributors to risk. Importance of individual containment failure mechanisms depends on plant features and in some cases on modeling assumptions; however the following mechanisms were found important: drywell shell melt-through caused by direct contact with core debris and drywell failure caused by rapid pressure/temperature pulses at time of vessel melt-through. Drywell failure caused by gradual pressure/temperature buildup due to gases and steam released during core/concrete interactions is important in some IPEs. In other IPEs vent was an important contributor. However, accidents that bypass containment (eg interfacing systems LOCA)or involve containment isolation failure were not important contributors to the CDF in any of the IPEs for Mark I plants. These accidents are also not important to risk (even though they can involve large fission product release) because their frequencies of occurrence are so much lower than frequencies of early structural failure caused by other accidents that dominate the CDF.

  11. LWR fuel rod behavior during reactor tests under loss-of-coolant conditions: Results of the FR2 in-pile tests

    Energy Technology Data Exchange (ETDEWEB)

    Karb, E.H.; Sepold, L.; Hofmann, P.; Petersen, C.; Schanz, G.; Zimmermann, H. (Kernforschungszentrum Karlsruhe G.m.b.H. (Germany, F.R.))

    1982-05-01

    Results of the FR2 in-pile tests on fuel rod behavior under loss-of-coolant accident (LOCA) conditions are presented. To investigate the possible influence of a nuclear environment on fuel rod failure mechanisms, unirradiated as well as irradiated (2500 to 35,000 MWd/tsub(U)) PWR-type test fuel rods were exposed to temperature transients simulating the second heatup phase of a LOCA. Loaded by internal overpressure, the cladding ballooned and ruptured. The burst data do not indicate major differences from results obtained out-of-pile with electrically heated fuel rod simulators, and do not show an influence of burnup. The fuel pellets in previously irradiated rods, already cracked during normal operation, crumbled completely in the regions with large cladding deformation. Post-test examinations revealed cladding mechanical behavior and oxidation to be comparable to out-of-pile results, with relatively little fission gas release during the transient.

  12. Safety analysis of a loss-of-coolant accident in a breeding blanket for experimental fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rocco, P.; Casini, G.; Djerassi, H.; Papa, L.; Pautasso, G.; Renda, V.; Rouyer, J.L.

    1985-07-01

    A LOCA in a blanket design proposed for NET (Next European Torus) is investigated. The structural analysis of a damaged breeder unit shows that this first containment barrier has a high probability of survival to this accident. The radioactive sources involved are evaluated and an assessment is made of all containment barriers and associated protection systems.

  13. Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

    Science.gov (United States)

    Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

    1972-01-01

    A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

  14. New procedure for declaring accidents resulting in bodily injuries

    CERN Multimedia

    2014-01-01

    The HR Department would like to remind members of personnel that, according to Administrative Circular No. 14 (Rev. 3), entitled “Protection of members of the personnel against the financial consequences of illness, accident and incapacity for work”, accidents resulting in bodily injuries and presumed to be of an occupational nature should, under normal circumstances, be declared within 10 working days of the accident having occurred, accompanied by a medical certificate. In an effort to streamline procedures, occupational accident declarations should be made via EDH using the “declaration of occupational accident” electronic form. For the declaration of non-occupational accidents resulting in bodily injuries of members of the CERN Health Insurance Scheme (CHIS), a new paper form has been elaborated that can be downloaded from the CHIS website and is also available from the UNIQA Helpdesk in the Main Building. If you encounter technical difficulties with these new ...

  15. Evaluation of a postulated loss of coolant accident (LOCA) due to a 160 cm{sup 2} break in a cold leg of Angra 2 nuclear power plant[LOCA; RELAP-5/MOD3.2.2g; nodalization]; Calculo do acidente postulado de perda de refrigerante por uma ruptura de 160 cm{sup 2} na perna fria da central nuclear Angra 2

    Energy Technology Data Exchange (ETDEWEB)

    Azevedo, Carlos Vicente Goulart de; Palmieri, Elcio Tadeu; Aronne, Ivan Dionysio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)

    2002-07-01

    The development of a qualified full nodalization of Angra2 NPP for RELAP5/Mod 3.2.2 gamma, aiming at the evaluation of a comprehensive number of accidents and transients, thus providing suitable safety analysis support for licensing purposes, is being carried out within the framework of CNEN internal technical cooperation, involving some of its institutes (CDTN, IPEN and IEN) and the Reactors Coordination (CODRE). This work presents a simulation of a postulated Angra2 small cold leg break loss of coolant accident (SBLOCA). A 160 cm{sup 2} break is supposed to occur at one cold leg between the main coolant pump and the reactor vessel and is described in the Angra2 Final Safety Analysis Report, section 15.6.4.1.3.4. The simulation of several types of transients and accidents is necessary to verify the adequate performance of the modeled logic and systems. In general, the analysis of such and accident allows to demonstrate the safety Injection System performance and the reliable transition between the high pressure safety injection, the accumulator injection and the residual heat removal phases. Furthermore, it is assumed that some components are out of service due to fail or repair in order to make a conservative analysis. The results showed a compatible behavior of the molded systems and that the simulated Emergency Core Cooling System was able to provide sufficient cooling to avoid any damage to the core. (author)

  16. Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 3—Calcium

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Sterling; Ali, Amir; LaBrier, Daniel [Department of Nuclear Engineering, University of New Mexico (United States); Blandford, Edward D, E-mail: edb@unm.edu [Department of Nuclear Engineering, University of New Mexico (United States); Howe, Kerry [Department of Civil Engineering, University of New Mexico (United States)

    2016-04-15

    Highlights: • Calcium leaching from NUKON fiberglass in borated TSP-buffered solution is independent of the level of fiberglass destruction. • The initial calcium release rate and the maximum calcium concentration increases with increased fiber concentration. • The calcium release in solution has a repeatable pattern of four distinct regions (prompt release, metastable, autocatalytic drop, and stable region) for all experiments. • Magnesium plays a significant role in initiating calcium precipitation in TSP-buffered environment. • Head loss through multi-constituents debris beds was found to increase progressively in all calcium concentration regions. - Abstract: Calcium that leaches from damaged or destroyed NUKON fiberglass in containment post a loss of coolant accident (LOCA) could lead to the formation of chemical precipitates. These precipitates could be filtered through the accumulated fibrous debris on the sump screen and compromising the emergency core cooling system (ECCS) sump pump performance. Reduced-scale leaching experiments were conducted on three solution inventory scales—bench (0.5 L), vertical column (31.5 L), and tank (1136 L) using three different flow conditions, and fiberglass concentrations (1.18–8 g/L) to investigate calcium release from NUKON fiber. All experiments were conducted in simulated post-LOCA water chemistry. (∼220 mM boric acid with ∼5.8 mM trisodium phosphate (TSP) buffer). Prior to the leaching tests, a preliminary experiment was carried out on the bench scale to determine the effect of the fiber preparation (unaltered and blended) method on calcium leaching. Results indicate that the extent of fiberglass destruction does not affect the amount of calcium released from fiberglass. Long-term calcium leach testing at constant temperature (80 °C) in borated TSP-buffered solution had repeatable behavior on all solution scales for different fiberglass concentrations. The calcium-leaching pattern can be divided into

  17. Risk-informed analysis of the large break loss of coolant accident and PCT margin evaluation with the RISMC methodology

    Energy Technology Data Exchange (ETDEWEB)

    Liang, T.H. [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Liang, K.S., E-mail: ksliang@alum.mit.edu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Cheng, C.K.; Pei, B.S. [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Patelli, E. [Institute of Risk and Uncertainty, University of Liverpool, Room 610, Brodie Tower, L69 3GQ (United Kingdom)

    2016-11-15

    Highlights: • With RISMC methodology, both aleatory and epistemic uncertainties have been considered. • 14 probabilistically significant sequences have been identified and quantified. • A load spectrum for LBLOCA has been conducted with CPCT and SP of each dominant sequence. • Comparing to deterministic methodologies, the risk-informed PCT margin can be greater by 44–62 K. • The SP of the referred sequence to cover 99% in the load spectrum is only 5.07 * 10{sup −3}. • The occurrence probability of the deterministic licensing sequence is 5.46 * 10{sup −5}. - Abstract: For general design basis accidents, such as SBLOCA and LBLOCA, the traditional deterministic safety analysis methodologies are always applied to analyze events based on a so called surrogate or licensing sequence, without considering how low this sequence occurrence probability is. In the to-be-issued 10 CFR 50.46a, the LBLOCA will be categorized as accidents beyond design basis and the PCT margin shall be evaluated in a risk-informed manner. According to the risk-informed safety margin characterization (RISMC) methodology, a process has been suggested to evaluate the risk-informed PCT margin. Following the RISMC methodology, a load spectrum of PCT for LBLOCA has been generated for the Taiwan’s Maanshan Nuclear Power plant and 14 probabilistic significant sequences have been identified. It was observed in the load spectrum that the conditional PCT generally ascends with the descending sequence occurrence probability. With the load spectrum covering both aleatory and epistemic uncertainties, the risk-informed PCT margin can be evaluated by either expecting value estimation method or sequence probability coverage method. It was found that by comparing with the traditional deterministic methodology, the PCT margin evaluated by the RISMC methodology can be greater by 44–62 K. Besides, to have a cumulated occurrence probability over 99% in the load spectrum, the occurrence probability

  18. Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

  19. A study of different cases of VVER reactor core flooding in a large break loss of coolant accident

    Directory of Open Access Journals (Sweden)

    Bezrukov Yury Alekseevich

    2016-01-01

    Full Text Available The paper covers the results of VVER core reflooding studies in fuel assembly (FA mockup of 126 fuel rod simulators with axial power peaking. The experiments were performed for two types of flooding. The first type is top flooding of the empty (steamed FA mockup. The second type is bottom flooding of the FA mockup with level of boiling water. The test parameters are as follows: the range of the supplied power to the bundle is from 40 to 320 kW, the cooling water flow rate is from 0.04 to 1.1 kg/s, the maximum temperature of the fuel rod simulator is 800 °C and the linear heat flux is from 0.1 to 1.0 kW/m. The test results were used for computer code validation.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hill, Robin L.; Conrady, Matthew M.

    2011-10-28

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

  1. Experimental Results on Pouring and Underwater Liquid Melt Spreading and Energetic Melt-coolant Interaction

    OpenAIRE

    Konovalenko, Alexander; Karbojian, Aram; Kudinov, Pavel

    2012-01-01

    In a hypothetical light water reactor (LWR) core-melt accident with corium release from the reactor  vessel,  the  ultimate  containment  integrity  is  contingent  on  coolability  of  the decay-heated core debris. Pouring of melt into a pool of water located in the reactor cavity is considered in several designs of existing and new LWRs  as a part of severe accident (SA) management strategies. At certain conditions of melt release into the pool (e.g. large ratio of the  vessel  breach  size...

  2. Statistical analysis of the blowdown phase of a loss-of-coolant accident in a pressurized water reactor as calculated by RELAP4/MOD6

    Energy Technology Data Exchange (ETDEWEB)

    Berman, M.; Byers, R.K.; Steck, G.P.

    1979-01-01

    A statistical study is presented of the blowdown phase of a design basis accident (double-ended cold leg guillotine break) in the Zion pressurized water reactor. The response surface method was employed to generate a polynomial approximation of the peak clad temperatures calculated by RELAP4/MOD6. The nodalization was a modification of the RELAP model of Zion developed in the BE/EM study. Twenty one variables were initially selected for the study. These variables, their ranges and distributions resulted from the best engineering judgement of NRC, Sandia, INFL, and other interested and knowledgeable investigators.

  3. Analysis of swimming pool accidents resulting in spinal cord injury.

    Science.gov (United States)

    Green, B A; Gabrielsen, M A; Hall, W J; O'Heir, J

    1980-04-01

    This paper is a summary of a study of 72 cases of swimming-pool accidents resulting in serious injuries with the potential of permanent disability. Sixty-four of the 72 cases resulted in spinal cord injuries, 57 of which involved quadriplegic lesions. The authors observed that the majority of these injuries resulted from a lack of good judgement and common sense rather than from intoxication or pool structural deficiencies. Also of note was the lack of appropriate first-aid and extrication rendered, as well as the absence of uniform treatment and care received by the majority of the patients.

  4. Westinghouse accident tolerant fuel program. Current results and future plans

    Energy Technology Data Exchange (ETDEWEB)

    Ray, Sumit; Xu, Peng; Lahoda, Edward; Hallstadius, Lars; Boylan, Frank [Westinghouse Electric Company LLC, Hopkins, SC (United States)

    2016-07-15

    This paper discusses the current status, results from initial tests, as well as the future direction of the Westinghouse's Accident Tolerant Fuel (ATF) program. The current preliminary testing is addressed that is being performed on these samples at the Massachusetts Institute of Technology (MIT) test reactor, initial results from these tests, as well as the technical learning from these test results. In the Westinghouse ATF approach, higher density pellets play a significant role in the development of an integrated fuel system.

  5. Sharp reduction in maximum fuel temperatures during loss of coolant accidents in a PBMR DPP-400 core, by means of optimised placement of neutron poisons

    Energy Technology Data Exchange (ETDEWEB)

    Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za

    2014-05-01

    In a preceding study, coupled neutronics and thermo-hydraulic simulations were performed with the VSOP-A diffusion code for the standard 9.6 wt% enriched 9 g uranium fuel spheres in the 400 MWth Pebble Bed Modular Reactor Demonstration Power Plant. The axial power profile peaked at about a third from the top of the fuel core and the radial profile peaked directly adjacent to the central graphite reflector. The maximum temperature during a Depressurised Loss of Coolant (DLOFC) incident was 1581.0 °C, which is close to the limit of 1600 °C above which the leakage of radioactive fission products through the TRISO coatings around the fuel kernels may become unacceptable. This may present licensing challenges and also limits the total power output of the reactor. In this article the results of an optimisation study of the axial and radial power profiles for this reactor are reported. The main aim was to minimise the maximum DLOFC temperature. Reducing the maximum equilibrium temperature during normal operation was a lesser aim. Minimising the maximum DLOFC temperature was achieved by placing an optimised distribution of {sup 10}B neutron poison in the central reflector. The standard power profiles are sub-optimal with respect to the passive leakage of decay heat during a DLOFC. Since the radial power profile peaks directly adjacent to the central reflector, the distance that the decay heat needs to be conducted toward the outside of the reactor and the ultimate heat sink is at a maximum. The sharp axial power profile peak means that most of the decay power is concentrated in a small part of the core volume, thereby sharply increasing the required outward heat flux in this hotspot region. Both these features sharply increase the maximum DLOFC temperatures in this hotspot. Therefore the axial distribution of the neutron poisons in the central reflector was optimised so as to push the equilibrium power density profile radially outward and to suppress the axial power peak

  6. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 4: External Pressurizer Surge Line Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report (PSAR) for the APT.

  7. APT Blanket System Loss-of-Coolant Accident Based on Initial Conceptual Design - Case 5: External RHR Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report.

  8. APT Blanket System Loss-of-Coolant Accident (LOCA) Analysis Based on Initial Conceptual Design - Case 3: External HR Break at Pump Outlet without Pump Trip

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal (HR) system. These simulations were performed for the Preliminary Safety Analysis Report.

  9. Environmentally Friendly Coolant System

    Energy Technology Data Exchange (ETDEWEB)

    David Jackson Principal Investigator

    2011-11-08

    Energy reduction through the use of the EFCS is most improved by increasing machining productivity. Throughout testing, nearly all machining operations demonstrated less land wear on the tooling when using the EFCS which results in increased tool life. These increases in tool life advance into increased productivity. Increasing productivity reduces cycle times and therefore reduces energy consumption. The average energy savings by using the EFCS in these machining operations with these materials is 9%. The advantage for end milling stays with flood coolant by about 6.6% due to its use of a low pressure pump. Face milling and drilling are both about 17.5% less energy consumption with the EFCS than flood coolant. One additional result of using the EFCS is improved surface finish. Certain machining operations using the EFCS result in a smoother surface finish. Applications where finishing operations are required will be able to take advantage of the improved finish by reducing the time or possibly eliminating completely one or more finishing steps and thereby reduce their energy consumption. Some machining operations on specific materials do not show advantages for the EFCS when compared to flood coolants. More information about these processes will be presented later in the report. A key point to remember though, is that even with equivalent results, the EFCS is replacing petroleum based coolants whose production produces GHG emissions and create unsafe work environments.

  10. Impact evaluation of the accident with release of a PWR coolant. Case study: Angra 3; Avaliacao do impacto de acidente com liberacao do refrigerante de reator PWR. Estudo de caso: Angra 3

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, Andre Silva de; Simoes Filho, Francisco Fernando Lamego; Soares, Abner Duarte; Lapa, Celso Marcelo Franklin, E-mail: flamego@ien.gov.b, E-mail: asoares@cnen.gov.b, E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-10-26

    It was postulated in the cooling system, a LOCA where was lost 431 m{sup 3} of coolant. The inventory was 1.87 x 10{sup 10} Bq/m{sup 3} of tritium, 2.22 x 10{sup 7} Bp/m{sup 3} of cobalt and 3.48 x 10{sup 8} Bq/m{sup 3} of cesium and was launched near tue Itaorna beach, Angra dos Reis, RJ, Brazil. By applying the model in the proposed scenery (Angra 1 and 2 functioning and Angra 3 with variation of water taking and discharge with a progressive reduction after the accident), the dilution of specific activity of the radionuclides reached inferior values after 22 hours, to the reference values. After 54 hours, the levels of radionuclides, in the indirect influence are already below the minimum values of activity detected by the laboratory of environmental monitoring of the CNAAA

  11. Source Term Analysis in Severe Accident Induced by Large Break Loss of Coolant Accident Coincident With Ship Blackout for Ship Reactor%船用堆大破口失水叠加全船断电严重事故源项分析

    Institute of Scientific and Technical Information of China (English)

    张彦招; 张帆; 赵新文; 郑映峰

    2013-01-01

    以某船用压水堆为研究对象,采用M ELCOR程序建立事故分析模型,研究大破口失水事故叠加全船断电严重事故下放射性裂变产物的行为,着重分析了惰性气体和CsI的释放、迁移、滞留特点及在堆舱内的分布。结果表明,83.12%惰性气体从堆芯释放出来,并主要存在于堆舱的气空间;83.08%的CsI从堆芯释放出来,其中,72.66%滞留在堆坑熔融物与一回路内,27.34%释放到堆舱内,并主要溶解于舱底水池中。本文分析结果可为舱室剂量评估、核应急管理提供依据。%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 . T he 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 w hich 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 .

  12. Simulation of a postulated loss of coolant accident due to a break in the pressurizer surge line of Angra 2 Nuclear Power Plan; Calculo do acidente postulado de perda de refrigerante por uma ruptura na linha de surto do pressurizador da central nuclear Angra 2

    Energy Technology Data Exchange (ETDEWEB)

    Azevedo, Carlos V. Goulart de; Palmieri, Elcio T.; Aronne, Ivan D. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)]. E-mail: cvga@cdtn.br; etp@cdtn.br; aroneid@cdtn.br

    2005-07-01

    This work presents the simulation of a postulated loss of coolant accident due to a 437 cm{sup 2} break in the pressurizer surge line of Angra 2 Nuclear Power Plant, as described in its Final Safety Analysis Report, section 15.6.4.1.3.11. This accident is characterized by a fast depressurization of the reactor coolant system followed by the actuation of the safety injection system. This work, which aims to develop and qualify a basic Angra 2 nodalization for RELAP5, was done in the framework of a CNEN internal technical cooperation involving three of its research centers (CDTN, IPEN and IEN) and its Reactors Division. This simulation is part of a comprehensive number of accidents and transients necessary to verify the adequacy of the modeled logic of the control and protection systems as well as the performance of the modeled thermal-hydraulic systems. Therefore this work contributes to the qualification process of the developed nodalization. (author)

  13. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 1: External HR Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    The APT blanket system has about 57 MW of thermal energy deposited within the blanket region under normal operating conditions from the release of neutrons and the interaction of the High energy particles with the blanket materials. This corresponds to about 48 percent of total thermal energy deposited in the APT target/blanket system. The deposited thermal energy under normal operation conditions is an important input parameter used in the thermal-hydraulic design and accident analysis.

  14. Traffic accident or dumping? - Striking results of a traffic accident reconstruction.

    Science.gov (United States)

    Nerger, Eric; Bayer, Ronny; Gärtner, Tobias; Dreßler, Jan; Ondruschka, Benjamin

    2017-01-01

    An atypical traffic accident scenario should be investigated directly at the crash site from all concerned professions, especially police men, forensic pathologists and technical experts, to get a personal overview and impression of the situation and the opportunity for interdisciplinary discussion. We present the rare case of a fatal traffic accident on a German motorway which was initially thought to be an accidental discovery of dumping a corpse. Based on autopsy findings, the technical investigation and the accident reconstruction, this case was solved as a spectacular form of a collision between a pedestrian and a bonnet-front car, which was not described elsewhere in scientific literature to the best of our knowledge. The pedestrian was hit in an upright body position, was lifted up by the car, smashed the windscreen and flew over the car with several body rotations. His flight curve ends directly at the roof of the car during brake processing, where the body touched the roof, smashed the rear-window and landed in the trunk. Based on the technical investigation, the driver of the car was not able to hide the accident. However, the pedestrian could have avoided the collision if he did not cross the motorway on foot.

  15. INTERCOMPARISON OF RESULTS FOR A PWR ROD EJECTION ACCIDENT

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-10-01

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

  16. [E-bikers are more often seriously injured in bicycle accidents: results from the Groningen bicycle accident database].

    Science.gov (United States)

    Poos, H P A M; Lefarth, T L; Harbers, J S; Wendt, K W; El Moumni, M; Reininga, I H F

    2017-01-01

    Analysing injury types, injury severity and mortality in victims of accidents with electric bicycles in comparison with conventional bicycles. Prospective cohort study. Data of patients treated at the Accident & Emergency Department of the University Medical Center Groningen after a bicycle accident are being entered in a database since 2014. We have analysed this database for accidents with electric bicycles (e-bikes) and conventional bicycles occurring among adult patients for the period of July 2014 to May 2016. 'Propensity score matching' was used to match e-bikers to conventional cyclists, based on age, gender and the presence of comorbidities. 107 of the 475 included victims were riding an e-bike. Average age of e-bikers and conventional cyclists was 65 years and 39 years respectively. Comorbidity was more common in e-bikers. E-bikers were injured significantly more severely than conventional cyclists. They had more severe injuries of the head and face, and upper and lower extremities. E-bikers were also admitted to the hospital more often, and for longer periods, and they underwent surgery more often. Mortality was the same. Propensity score matching revealed that e-bikers had multiple severe injuries (ISS > 15) twice as often as conventional cyclists, that they had more severe head injuries and were admitted for longer periods than conventional cyclists. E-bikers who had a bicycle accident had more severe injuries, more frequently had multiple injuries and had more severe head injuries than conventional cyclists. This resulted in a greater need for care. Preventive measures such as riding lessons and helmet use should be encouraged. Care providers should pay extra attention to the possibility of severe injuries when a patient had a bicycle accident with an e-bike.

  17. Spinal injury resulting from car accident: Focus to prevention.

    Science.gov (United States)

    Fakharian, Esmaeil; Mohammadzadeh, Mahdi; Saberi, Hamid Reza; Fazel, Mohammad Reza; Rejali, Mohsen; Akbari, Hossein; Mirzadeh, Azadeh Sadat; Mohammadzadeh, Javad

    2017-01-01

    To determine and compare the patterns of spinal injury in car occupants. Retrospective cross-sectional study enrolling all patients with spinal fracture after car accidents, who were admitted to hospital more than 24 h during 2004-2009. The lumbosacral spine was the most commonly involved region (64.8). Six patients had spinal cord injury (6.6%). The majority of the victims were drivers of the vehicle (86.8%) and remaining were passengers. There was a significant difference in lumbar anatomic region (P = 0.05) and place of accident (P = 0.05) in car occupants' position (P = 0.05). Car rollover was the most common mechanism of spinal fractures. There was a significant difference in lumbar anatomic region (P = 0.05), and two or more associated organ injuries (P ≤ 0.05) in car accident mechanism (P = 0.05). The chance of sustaining serious spine and associated multiple injuries in car accidents is quite high in our today's society. This may be due to the low level of standards for car manufacturing, absence or inadequacy of appropriate safety measures in cars, and poorly designed roads and problems in quality of driving to mention some reasons. Therefore, these victims are prone to significant morbidity and even mortality and need more specific prehospital supportive interventions.

  18. Extended Life Coolant Testing

    Science.gov (United States)

    2016-06-06

    ELC – Extended Life Coolant SCA – Supplemental Coolant Additive SOW – Scope of Work SwRI – Southwest Research Institute TARDEC – Tank Automotive...ethylene or propylene glycol and 35% extended life coolant #1 (ELC1) with a balance of water. At a higher ELC1 content of 45% or 50%, the mass loss...UNCLASSIFIED TABLE OF CONTENTS EXTENDED LIFE COOLANT TESTING INTERIM REPORT TFLRF No. 478 by Gregory A. T. Hansen Edwin A

  19. Simulation of a transient with loss of primary coolant due to a small rupture in Angra 2 nuclear power plant with RELAP5/MOD3.2.2G code; Simulacao de um acidente postulado de perda de refrigerante primario por pequena ruptura na usina de Angra 2 com o codigo RELAP5/MOD3.2.2G

    Energy Technology Data Exchange (ETDEWEB)

    Sabundjian, Gaiane; Andrade, Delvonei Alves de [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    2002-07-01

    This paper presents a nodalization for Angra 2 Nuclear Power Plant, as well as the results obtained for a Small Break Loss of Coolant Accident (SBLOCA), simulated with RELAP5/MOD3.2G code. This accident consists in a small break (380 m{sup 2}) in the line of the Emergency Core Coolant System (ECCS) in loop 20 of Angra 2. Results are not as expected, however they are satisfactory regarding the nodalization used. (author)

  20. Effects of molten material temperatures and coolant temperatures on vapor explosion

    Institute of Scientific and Technical Information of China (English)

    LI Tianshu; YANG Yanhua; YUAN Minghao; HU Zhihua

    2007-01-01

    An observable experiment facility for low-temperature molten materials to be dropped into water was set up in this study to investigate the mechanism of the vapor explosion. The effect of the fuel and coolant interaction(FCI) on the vapor explosion during the severe accidents of a fission nuclear reactor has been studied. The experiment results showed that the molten material temperature has an important effect on the vapor explosion behavior and pressure. The increase of the coolant temperature would decrease the pressure of the vapor explosion.

  1. CHEMICAL EFFECTS ON PWR SUMP STRAINER BLOCKAGE AFTER A LOSS-OF-COOLANT ACCIDENT: REVIEW ON U.S. RESEARCH EFFORTS

    Directory of Open Access Journals (Sweden)

    CHI BUM BAHN

    2013-06-01

    Full Text Available Industry- or regulatory-sponsored research activities on the resolution of Generic Safety Issue (GSI-191 were reviewed, especially on the chemical effects. Potential chemical effects on the head loss across the debris-loaded sump strainer under a post-accident condition were experimentally evidenced by small-scale bench tests, integrated chemical effects test (ICET, and vertical loop head loss tests. Three main chemical precipitates were identified by WCAP-16530-NP: calcium phosphate, aluminum oxyhydroxide, and sodium aluminum silicate. The former two precipitates were also identified as major chemical precipitates by the ICETs. The assumption that all released calcium would form precipitates is reasonable. CalSil insulation needs to be minimized especially in a plant using trisodium phosphate buffer. The assumption that all released aluminum would form precipitates appears highly conservative because ICETs and other studies suggest substantial solubility of aluminum at high temperature and inhibition of aluminum corrosion by silicate or phosphate. The industry-proposed chemical surrogates are quite effective in increasing the head loss across the debris-loaded bed and more effective than the prototypical aluminum hydroxide precipitates generated by in-situ aluminum corrosion. There appears to be some unresolved potential issues related to GSI-191 chemical effects as identified in NUREG/CR-6988. The United States Nuclear Regulatory Commission, however, concluded that the implications of these issues are either not generically significant or are appropriately addressed, although several issues associated with downstream in-vessel effects remain.

  2. Experimental Study on the Effect of Late-Phase Coolant Injection on the Metallic Layer

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik; Hong, Seong Wan; Kim, Hee Dong

    2007-04-15

    Sustained heating experiments, named ELIAS (Experiments on Late-phase coolant Injection to ASsess the mitigation of focusing effect of metallic layer), were performed to quantify the boiling heat removal rate at the upper surface of a metallic layer for precise evaluations on the effect of a late in-vessel coolant injection. Heat fluxes from the melt layer to the water pool varied from 250 to 550kW/m2 depending on the experimental conditions. Comparison of boiling heat fluxes between the ELIAS experiments and the calculation using the Berenson's film boiling correlation shows that effective heat removal was accomplished via late-phase coolant injection in the ELIAS experiments. In this study, simple model was developed to evaluate the mitigation of focusing effect in the metallic layer via late-phase coolant injection. The ELIAS experimental data on the heat transfer rate at the upper surface of the metallic layer were used as input data in the simple model. The calculation results for the large break loss of coolant accident in the APR1400 show that the risk induced by the focusing effect is highly dependent on the metallic layer thickness and the integrity of the reactor pressure vessel can be enhanced via late-phase coolant injection.

  3. Severe accident analysis of a station blackout accident using MAAP-CANDU for the Point Lepreau station refurbishment project level 2 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Petoukhov, S.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station, using the MAAP-CANDU code to simulate the progression of severe core damage accidents and fission product releases. Five representative severe accidents were selected: Station Blackout, Small Loss-of-Coolant, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State. Analysis results for the reference station blackout accident are discussed in this paper. (author)

  4. Machine coolant waste reduction by optimizing coolant life. Project summary

    Energy Technology Data Exchange (ETDEWEB)

    Pallansch, J.

    1995-08-01

    The project was designed to study the following: A specific water-soluble coolant (Blasocut 2000 Universal) in use with a variety of machines, tools, and materials; Coolant maintenance practices associated with three types of machines; Health effects of use and handling of recycled coolant; Handling practices for chips and waste coolant; Chip/coolant separation; and Oil/water separation.

  5. Severe accident analysis of a small LOCA accident using MAAP-CANDU support level 2 PSA for the Point Lepreau station refurbishment project

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Brown, M.J.; Mathew, P.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station. The MAAP4-CANDU code was used to calculate the progression of postulated severe core damage accidents and fission product releases. Five representative severe core damage accidents were selected: Station Blackout, Small Loss-of-Coolant Accident, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State Accident. Analysis results for only the reference Small LOCA Accident scenario (which is a very low probability event) are discussed in this paper. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  7. Comparison of THALES and VIPRE-01 Subchannel Codes for Loss of Flow and Single Reactor Coolant Pump Rotor Seizure Accidents using Lumped Channel APR1400 Geometry

    Energy Technology Data Exchange (ETDEWEB)

    Oezdemir, Erdal; Moon, Kang Hoon; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Kim, Yongdeog [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    Subchannel analysis plays important role to evaluate safety critical parameters like minimum departure from nucleate boiling ratio (MDNBR), peak clad temperature and fuel centerline temperature. In this study, two different subchannel codes, VIPRE-01 (Versatile Internals and Component Program for Reactors: EPRI) and THALES (Thermal Hydraulic AnaLyzer for Enhanced Simulation of core) are examined. In this study, two different transient cases for which MDNBR result play important role are selected to conduct analysis with THALES and VIPRE-01 subchannel codes. In order to get comparable results same core geometry, fuel parameters, correlations and models are selected for each code. MDNBR results from simulations by both code are agree with each other with negligible difference. Whereas, simulations conducted by enabling conduction model in VIPRE-01 shows significant difference from the results of THALES.

  8. Flow boiling test of GDP replacement coolants

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.H. [comp.

    1995-08-01

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C{sub 4}F{sub 10} and C{sub 4}F{sub 8}, were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C{sub 4}F{sub 10} mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C{sub 4}F{sub 10} weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd.

  9. A numerical study on a lumped-parameter model and a CFD code coupling for the analysis of the loss of coolant accident in a reactor containment; Etude numerique 0D-multiD pour l'analyse de perte de refrigerant dans une enceinte de confinement d'un reacteur nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Y.J.

    2005-12-15

    In the case of PWR severe accident (Loss of Coolant Accident, LOCA), the inner containment ambient properties such as temperature, pressure and gas species concentrations due to the released steam condensation are the main factors that determine the risk. For this reason, their distributions should be known accurately, but the complexity of the geometry and the computational costs are strong limitations to conduct full three-dimensional numerical simulations. An alternative approach is presented in this thesis, namely, the coupling between a lumped-parameter model and a CFD. The coupling is based on the introduction of a 'heat transfer function' between both models and it is expected that large decreases in the CPU-costs may be achieved. First of all, wall condensation models, such as the Uchida or the Chilton-Colburn models which are implemented in the code CAST3M/TONUS, are investigated. They are examined through steady-state calculations by using the code TONUS-0D, based on lumped parameter models. The temperature and the pressure within the inner containment are compared with those reported in the archival literature. In order to build the 'heat transfer function', natural convection heat transfer is then studied by using the code CAST3M for a partitioned cavity which represents a simplified geometry of the reactor containment. At a first step, two-dimensional natural convection heat transfer without condensation is investigated only. Either the incompressible-Boussinesq fluid flow model or the asymptotic low Mach model are considered for solving the time dependent conservation equations. The SUPG finite element method and the implicit scheme are applied for the numerical discretization. The computed results are qualified by the second-order Richardson extrapolation method which allows obtaining the so-called 'Exact values', i.e. grid size independent values. The computations are also validated through non-partitioned cavity case

  10. Development and test results of the Realtime Severe Accident Model 5 (RSAM5) based on the MAAP5 For the Kori 1 simulator

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jin Hyuk; Lee, Myeong Soo [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The Real Time Severe Accident Model (RSAM) in the Kori simulator employs the standard MAAP 5.01.1101 code (which is defined as MAAP 5.01) plus several statically linked libraries that interface with the simulator environment. The physical phenomena that can be envisioned inside the reactor vessel, the reactor coolant system (RCS), and the containment during severe accidents are comprehensively modeled by the MAAP5 code. The MAAP5 code has been known to be a reliable tool for understanding the sequence of events that occur during severe LWR accidents, evaluating the consequences of the failure of emergency systems, assessing the effects of operator interventions, and investigating the influence of design features of the RCS, containment, and safety systems on the accident consequences. The purpose of this paper is to describe the modeling of the Kori Unit 1 nuclear plant with the MAAP5 code and major outputs in the event of the SBO, SBO + SGTR, SBO + LBLOCA.

  11. Development of the Severe Accident Analysis DB for the Severe Accident Management Expert System (I)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Ahn, Kwang Il [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    This report contains analysis methodologies and calculation results of 5 initiating events of the severe accident analysis database system. The Ulchin 3,4 NPP has been selected as reference plants. Based on the probabilistic safety analysis of the corresponding plant, 54 accident scenarios, which was predicted to have more than 10-10 /ry occurrence frequency, have been analyzed as base cases for the Large loss of Coolant sequence database. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results in this report will be utilized as input data to develop the database system

  12. Corrosion of magnesium alloys in commercial engine coolants

    Energy Technology Data Exchange (ETDEWEB)

    Song, G.; StJohn, D.H. [CRC for Cast Metals Manufacturing (CAST), Division of Materials, School of Engineering, The University of Queensland, Brisbane, QLD 4072 (Australia)

    2005-01-01

    A number of magnesium alloys show promise as engine block materials. However, a critical issue for the automotive industry is corrosion of the engine block by the coolant and this could limit the use of magnesium engine blocks. This work assesses the corrosion performance of conventional magnesium alloy AZ91D and a recently developed engine block magnesium alloy AM-SC1 in several commercial coolants. Immersion testing, hydrogen evolution measurement, galvanic current monitoring and the standard ASTM D1384 test were employed to reveal the corrosion performance of the magnesium alloys subjected to the coolants. The results show that the tested commercial coolants are corrosive to the magnesium alloys in terms of general and galvanic corrosion. The two magnesium alloys exhibited slightly different corrosion resistance to the coolants with AZ91D being more corrosion resistant than AM-SC1. The corrosivity varied from coolant to coolant. Generally speaking, an organic-acid based long life coolant was less corrosive to the magnesium alloys than a traditional coolant. Among the studied commercial coolants, Toyota long life coolant appeared to be the most promising one. In addition, it was found that potassium fluoride effectively inhibited corrosion of the magnesium alloys in the studied commercial coolants. Both general and galvanic corrosion rates were significantly decreased by addition of KF, and there were no evident side effects on the other engine block materials, such as copper, solder, brass, steel and aluminium alloys, in terms of their corrosion performance. The ASTM D 1384 test further confirmed these results and suggested that Toyota long life coolant with 1%wt KF addition is a promising coolant for magnesium engine blocks. (Abstract Copyright [2005], Wiley Periodicals, Inc.)

  13. An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

    Science.gov (United States)

    Gamble, K. A.; Barani, T.; Pizzocri, D.; Hales, J. D.; Terrani, K. A.; Pastore, G.

    2017-08-01

    Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterion is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Further experiments are required to confirm these observations.

  14. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  15. Video analysis of the biomechanics of a bicycle accident resulting in significant facial fractures.

    Science.gov (United States)

    Syed, Shameer H; Willing, Ryan; Jenkyn, Thomas R; Yazdani, Arjang

    2013-11-01

    This study aimed to use video analysis techniques to determine the velocity, impact force, angle of impact, and impulse to fracture involved in a video-recorded bicycle accident resulting in facial fractures. Computed tomographic images of the resulting facial injury are presented for correlation with data and calculations. To our knowledge, such an analysis of an actual recorded trauma has not been reported in the literature. A video recording of the accident was split into frames and analyzed using an image editing program. Measurements of velocity and angle of impact were obtained from this analysis, and the force of impact and impulse were calculated using the inverse dynamic method with connected rigid body segments. These results were then correlated with the actual fracture pattern found on computed tomographic imaging of the subject's face. There was an impact velocity of 6.25 m/s, impact angles of 14 and 6.3 degrees of neck extension and axial rotation, respectively, an impact force of 1910.4 N, and an impulse to fracture of 47.8 Ns. These physical parameters resulted in clinically significant bilateral mid-facial Le Fort II and III pattern fractures. These data confer further understanding of the biomechanics of bicycle-related accidents by correlating an actual clinical outcome with the kinematic and dynamic parameters involved in the accident itself and yielding a concrete evidence of the velocity, force, and impulse necessary to cause clinically significant facial trauma. These findings can aid in the design of protective equipment for bicycle riders to help avoid this type of injury.

  16. Evaluating the consequences of loss of flow accident for a typical VVER-1000 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mirvakili, S.M.; Safaei, S. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering, School of Mechanical Engineering; Faghihi, F. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Safety Research Center

    2010-07-01

    The loss of coolant flow in a nuclear reactor can result from a mechanical or electrical failure of the coolant pump. If the reactor is not tripped promptly, the immediate effect is a rapid increase in coolant temperature, decrease in minimum departure from nucleate boiling ratio (DNBR) and fuel damage. This study evaluated the shaft seizure of a reactor coolant pump in a VVER-1000 nuclear reactor. The locked rotor results in rapid reduction of flow through the affected reactor coolant loop and in turn leads to an increase in the primary coolant temperature and pressure. The analysis was conducted with regard for superimposing loss of power to the power plant at the initial accident moment. The required transient functions of flow, pressure and power were obtained using system transient calculations applied in COBRA-EN computer code in order to calculate the overall core thermal-hydraulic parameters such as temperature, critical heat flux and DNBR. The study showed that the critical period for the locked rotor accident is the first few seconds during which the maximum values of pressure and temperature are reached. 10 refs., 1 tab., 3 figs.

  17. Radiological protection from radioactive waste management in existing exposure situations resulting from a nuclear accident.

    Science.gov (United States)

    Sugiyama, Daisuke; Hattori, Takatoshi

    2013-01-01

    In environmental remediation after nuclear accidents, radioactive wastes have to be appropriately managed in existing exposure situations with contamination resulting from the emission of radionuclides by such accidents. In this paper, a framework of radiation protection from radioactive waste management in existing exposure situations for application to the practical and reasonable waste management in contaminated areas, referring to related ICRP recommendations was proposed. In the proposed concept, intermediate reference levels for waste management are adopted gradually according to the progress of the reduction in the existing ambient dose in the environment on the basis of the principles of justification and optimisation by taking into account the practicability of the management of radioactive waste and environmental remediation. It is essential to include the participation of relevant stakeholders living in existing exposure situations in the selection of reference levels for the existing ambient dose and waste management.

  18. Blade-to-coolant heat-transfer results and operating data from a natural-convection water-cooled single-stage turbine

    Science.gov (United States)

    Diaguila, Anthony J; Freche, John C

    1951-01-01

    Blade-to-coolant heat-transfer data and operating data were obtained with a natural-convection water-cooled turbine over range of turbine speeds and inlet-gas temperatures. The convective coefficients were correlated by the general relation for natural-convection heat transfer. The turbine data were displaced from a theoretical equation for natural convection heat transfer in the turbulent region and from natural-convection data obtained with vertical cylinders and plates; possible disruption of natural convection circulation within the blade coolant passages was thus indicated. Comparison of non dimensional temperature-ratio parameters for the blade leading edge, midchord, and trailing edge indicated that the blade cooling effectiveness is greatest at the midchord and least at the trailing edge.

  19. 二代改进型核电厂严重事故下一回路卸压时机敏感性研究%Sensitivity Analysis on Time of Reactor Coolant System Depressurization under Severe Accident for Generation II+ Nuclear Power Plants

    Institute of Scientific and Technical Information of China (English)

    种毅敏; 杨志义; 石雪垚; 张佳佳; 李春; 倪曼; 徐雨婷

    2015-01-01

    Reactor coolant system (RCS)depressurization is necessary measure for nuclear power plant to mitigate the severe accident,as well as a significant part of the severe accident management guidelines (SAMG).Difference may exist on the time of RCS depressurization in different NPPs.In this paper,based on MAAP4,the sensitivity analysis of time to implement RCS depressurization is performed.A typical integrated computer program,and different effects on mitigation of severe accident are compared.The simulation scenario is typical drill situation of generation II + NPPs,and conclusions can be reference for similar NPPs to implement severe accident management strategy.%一回路卸压是核电厂缓解严重事故的必要手段,也是严重事故管理导则(SAMG)的重要内容,国内核电厂严重事故管理中对一回路卸压的要求并不相同,本文基于典型二代改进型核电厂 SAMG 演练的场景,使用一体化计算程序 MAAP4,对一回路卸压时机进行敏感性分析,比较不同卸压时机对缓解严重事故效果的影响,所给出的结论可为相同类型核电厂制定严重事故管理策略时提供参考。

  20. SIMMER-III Analyses of Local Fuel-Coolant Interactions in a Simulated Molten Fuel Pool: Effect of Coolant Quantity

    Directory of Open Access Journals (Sweden)

    Songbai Cheng

    2015-01-01

    Full Text Available Studies on local fuel-coolant interactions (FCI in a molten pool are important for the analyses of severe accidents that could occur for sodium-cooled fast reactors (SFRs. To clarify the mechanisms underlying this interaction, in recent years, several experimental tests, with comparatively larger difference in coolant volumes, were conducted at the Japan Atomic Energy Agency by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this study, to further understand this interaction, interaction characteristics including the pressure buildup as well as mechanical energy release and its conversion efficiency are investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It is found that the SIMMER-III code not only reasonably simulates the transient pressure and temperature variations during local FCIs, but also supports the limited tendency of pressurization and resultant mechanical energy release as observed from experiments when the volume of water delivered into the pool increases. The performed analyses also suggest that the most probable reason leading to such limited tendency should be primarily due to an isolation effect of vapor bubbles generated at the water-melt interface.

  1. Modeling of melt-coolant mixing by bottom injection

    Energy Technology Data Exchange (ETDEWEB)

    Kazachkov, I.V.; Paladino, D.; Sehgal, B.R. [Royal Inst. of Tech., Div. of Nuclear Power Safety, Stockholm (Sweden)

    2001-07-01

    In this paper, the flow characteristics during the coolant injection, with submerged nozzles, at the bottom of a molten pool are studied. The flow pattern developed by the rising coolant is considered for the case of complete coolant vaporization, and the pool-coolant phase distributions are assessed by a modeling approach delivered from literature for a heterogeneous turbulent jet. To calculate the basic characteristics of such flow, integral relationships are proposed for the two-phase boundary layer. The results of numerical computations and approximate solution are compared with the experimental data obtained in the low temperature experiments, conducted in the DECOBI (debris coolability by bottom injection) facility. (authors)

  2. Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ, Seoul (Korea, Republic of)

    2015-10-15

    In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

  3. Resolve! Version 2.5: Flammable Gas Accident Analysis Tool Acceptance Test Plan and Test Results

    Energy Technology Data Exchange (ETDEWEB)

    LAVENDER, J.C.

    2000-10-17

    RESOLVE! Version 2 .5 is designed to quantify the risk and uncertainty of combustion accidents in double-shell tanks (DSTs) and single-shell tanks (SSTs). The purpose of the acceptance testing is to ensure that all of the options and features of the computer code run; to verify that the calculated results are consistent with each other; and to evaluate the effects of the changes to the parameter values on the frequency and consequence trends associated with flammable gas deflagrations or detonations.

  4. Lithium as a blanket coolant

    Energy Technology Data Exchange (ETDEWEB)

    Wells, W.M.

    1977-01-01

    Recent re-assessment of tokamak reactors which move towards smaller size and lower required field strength (higher beta)/sup 2/ change the picture as regards the magnitude of MHD effects on flow resistance for lithium coolant. Perhaps the most important consequence of this as regards use of this coolant is that of clear acceptability of such effects when the flow is predominantly transverse to the magnetic field. This permits defining a blanket that consists entirely of round tubes containing the circulated lithium with voids between the tubes. Required thermal-hydraulic calculations are then on bases which are well established, especially in view of recent results dealing with perturbations of ducts and magnetic fields. Mitigation of MHD effects is feasible through tapering of tube wall thickness or use of insulated layers, but their use was not mandatory for the assumed conditions. Blanket configurations utilizing flowing lithium in round tubes immersed in static lithium may be suitable, but calculational methods do not now exist for this situation. Use of boiling potassium or cesium appears to be prohibitive in terms of vapor flow area when temperature levels are consistent with stainless steel. Liquid sodium, in addition to not being a breeding material, requires higher velocity than lithium for the same heat removal.

  5. A Taylor-Affine Arithmetic for analyzing the calculation result uncertainty in accident reconstruction.

    Science.gov (United States)

    Zou, Tiefang; Peng, Haitao; Cai, Ming; Wu, Hequan; Hu, Lin

    2016-09-01

    In order to analyze the uncertainty of a reconstructed result, the Interval Algorithm (IA), the Affine Arithmetic (AA) and the Modified Affine Arithmetic (MAA) were introduced firstly, and then a Taylor-Affine Arithmetic (TAA) was proposed based on the MAA and Taylor series. Steps of the TAA, especially in analyzing uncertainty of a simulation result were given. Through the preceding five numerical cases, its application was demonstrated and its feasibility was validated. Results showed that no matter other methods (The IA, AA, the Upper and Lower bound Method, the Finite Difference Method) work well or bad, the TAA work well, even under the condition that the MAA cannot work in some cases because of the division/root operation in these models. Furthermore, in order to make sure that the result obtained from the TAA can be very close to the accurate interval, a simple algorithm was proposed based on the sub-interval technique, its feasibility was validated by two other numerical cases. Finally, a vehicle-pedestrian test was given to demonstrate the application of the TAA in practice. In the vehicle-pedestrian test, the interval [35.5, 39.1]km/h of the impact velocity can be calculated according to steps of the TAA, such interval information will be more useful in accident responsibility identification than a single number. This study will provide a new alternative method for uncertainty analysis in accident reconstruction.

  6. Progress in accident analysis of the HYLIFE-II inertial fusion energy power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, S; Latkowski, J F; Gomez del Rio, J; Sanz, J

    2000-10-11

    The present work continues our effort to perform an integrated safety analysis for the HYLIFE-II inertial fusion energy (IFE) power plant design. Recently we developed a base case for a severe accident scenario in order to calculate accident doses for HYLIFE-II. It consisted of a total loss of coolant accident (LOCA) in which all the liquid flibe (Li{sub 2}BeF{sub 4}) was lost at the beginning of the accident. Results showed that the off-site dose was below the limit given by the DOE Fusion Safety Standards for public protection in case of accident, and that his dose was dominated by the tritium released during the accident.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  8. LLNL Results from CALIBAN-PROSPERO Nuclear Accident Dosimetry Experiments in September 2014

    Energy Technology Data Exchange (ETDEWEB)

    Lobaugh, M. L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hickman, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wong, C. W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wysong, A. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Merritt, M. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinrichs, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Topper, J. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-05-21

    Lawrence Livermore National Laboratory (LLNL) uses thin neutron activation foils, sulfur, and threshold energy shielding to determine neutron component doses and the total dose from neutrons in the event of a nuclear criticality accident. The dosimeter also uses a DOELAP accredited Panasonic UD-810 (Panasonic Industrial Devices Sales Company of America, 2 Riverfront Plaza, Newark, NJ 07102, U.S.A.) thermoluminescent dosimetery system (TLD) for determining the gamma component of the total dose. LLNL has participated in three international intercomparisons of nuclear accident dosimeters. In October 2009, LLNL participated in an exercise at the French Commissariat à l’énergie atomique et aux énergies alternatives (Alternative Energies and Atomic Energy Commission- CEA) Research Center at Valduc utilizing the SILENE reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison at CEA Valduc, this time with exposures at the CALIBAN reactor (Hickman et al. 2011). This paper discusses LLNL’s results of a third intercomparison hosted by the French Institut de Radioprotection et de Sûreté Nucléaire (Institute for Radiation Protection and Nuclear Safety- IRSN) with exposures at two CEA Valduc reactors (CALIBAN and PROSPERO) in September 2014. Comparison results between the three participating facilities is presented elsewhere (Chevallier 2015; Duluc 2015).

  9. A study on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tai; Choi, Y. D.; Choi, J. H.; Kim, T. J.; Jeong, K. C.; Kwon, S. W.; Kim, B. H.; Jeong, J. Y.; Park, J. H.; Kim, K. R.; Jo, B. R.

    1997-08-01

    A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H{sub 2} gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs.

  10. RESULTS OF IMPLEMENTATION OF TARGET PROGRAMS ON RADIATION ACCIDENT REMEDIATION FOR THE PERIOD TO 2010 AND PROSPECTS OF FURTHER ACTIVITIES

    Directory of Open Access Journals (Sweden)

    T. A. Marchenko

    2008-01-01

    Full Text Available The report contains information about measures undertaken by the Ministry of Civil Defense and Emergencies of the Russian Federation in the framework of implementation of the state policy in the field of radiation accidents remediation. Results of works realized in the framework of target programs on remediation of radiation accidents at Chernobyl NPP and Production Association MAYAK, and on problems caused by nuclear weapon tests at Seminalatinsk test site are presented.

  11. Transgenerational genomic instability in children of irradiated parents as a result of the Chernobyl Nuclear Accident

    Energy Technology Data Exchange (ETDEWEB)

    Aghajanyan, Anna, E-mail: ann-aghajanyan@yandex.ru [Cytogenetics Laboratory, FSI Russian Scientific Center of Roentgenology and Radiology, Profsoyuznaya 86, GSP-7, Moscow, 117997 (Russian Federation); Suskov, Igor [Laboratory of Ecological Genetics, N.I. Vavilov Institute of General Genetics Russian Academy of Sciences, Gybkin st. 3, Moscow 119991 (Russian Federation)

    2009-12-01

    The study of families irradiated as a result of the accident at the Chernobyl Nuclear Power Plant revealed significantly increased aberrant genomes frequencies (AGFs) not only in irradiated parents (n = 106, p < 0.01), but also in their children born after the accident (n = 159, p < 0.05). This is an indicative of the phenomenon of transgenerational genomic instability. To elucidate this phenomenon, experiments were undertaken to model genomic instability by using single and fractional in vitro {gamma}-irradiation ({sup 137}Cs) of peripheral blood samples from the children and their parents at doses of 0.1, 0.2 and 0.3 Gy. The spectrum and frequency of chromosome aberrations were studied in the 1st and 2nd cell generations. The average AGF was significantly increased at all doses (except 0.1 Gy) in children of irradiated parents, as compared to children born from non-irradiated parents. Amplification of cells with single-break chromosome aberrations in mitosis 2, as compared to mitosis 1, suggests the replication mechanism of realization of potential damage in DNA and the occurrence of genomic instability in succeeding cell generations.

  12. Unusually extensive head trauma in a hydraulic elevator accident: post-mortem MSCT findings, autopsy results and scene reconstruction.

    Science.gov (United States)

    Jacobsen, Christina; Schön, Corinna A; Kneubuehl, Beat; Thali, Michael J; Aghayev, Emin

    2008-10-01

    Accidental or intentional falls from a height are a form of blunt trauma and occur frequently in forensic medicine. Reports describing elevator accidents as a small subcategory of falls from heights are rare in the medical literature and no report on injury patterns or scene reconstruction of such an accident was found. A case of an accident in a hydraulic elevator with a man falling 3m was examined using post-mortem multi-slice computed tomography (MSCT) and autopsy. The man suffered an unusually extensive trauma and died at the scene. Post-mortem MSCT examination showed a comminute fracture of the skull, the right femur and the first lumbar vertebra. Severe lacerations of the brain with epidural, subdural and subarachnoidal haemorrhages over both hemispheres were diagnosed. Autopsy confirmed these findings. To reconstruct the accident we used radiological and autopsy results as well as findings at the scene.

  13. The estimation of economic impacts resulting from the severe accidents of a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Jung, Won dea

    2001-03-01

    The economic impacts resulting from the severe accidents of a nuclear power plant were estimated for the different combinations of a release parameters and metrorological data. According to the cost estimation for the basic scenarios, the population dependent cost is dominant. The cost for the protective actions such as evacuation and relocation have a small portion in the total cost and show little variation from scenario to scenario. The economic cost estimation for the seasonal scenarios show very similar trend as that for the basic scenarios. There are little or small variation in the economic cost for the different scenarios for each season except for the season-5 scenario. The health effect value shows maximum in Summer and minimum in Fall. On the contrast, the economic cost shows maximum in Fall and minimum in Summer. The result will be used as basic data in the establishment of effective emergency response and in the cost/benefit analysis in developing optimum risk reduction strategies.

  14. The estimation of economic impacts resulting from the severe accidents of a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Jung, Won dea

    2001-03-01

    The economic impacts resulting from the severe accidents of a nuclear power plant were estimated for the different combinations of a release parameters and metrorological data. According to the cost estimation for the basic scenarios, the population dependent cost is dominant. The cost for the protective actions such as evacuation and relocation have a small portion in the total cost and show little variation from scenario to scenario. The economic cost estimation for the seasonal scenarios show very similar trend as that for the basic scenarios. There are little or small variation in the economic cost for the different scenarios for each season except for the season-5 scenario. The health effect value shows maximum in Summer and minimum in Fall. On the contrast, the economic cost shows maximum in Fall and minimum in Summer. The result will be used as basic data in the establishment of effective emergency response and in the cost/benefit analysis in developing optimum risk reduction strategies.

  15. Experimental study on thermal-hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Takashi; Watanabe, Hironori; Araya, Fumimasa; Nakajima, Katsutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwamura, Takamichi; Murao, Yoshio

    1998-02-01

    A conceptual design study of a passive safety light water reactor JPSR has been performed at Japan Atomic Energy Research Institute JAERI. A pressure balanced coolant injection experiment has been carried out, with an objective to understand thermal-hydraulic characteristics of a passive coolant injection system which has been considered to be adopted to JPSR. This report summarizes experimental results and data recorded in experiment run performed in FY. 1993 and 1994. Preliminary experiments previously performed are also briefly described. As the results of the experiment, it was found that an initiation of coolant injection was delayed with increase in a subcooling in the pressure balance line. By inserting a separation device which divides the inside of core make-up tank (CMT) into several small compartments, a diffusion of a high temperature region formed just under the water surface was restrained and then a steam condensation was suppressed. A time interval from an uncovery of the pressure balance line to the initiation of the coolant injection was not related by a linear function with a discharge flow rate simulating a loss-of-coolant accident (LOCA) condition. The coolant was injected intermittently by actuation of a trial fabricated passive valve actuated by pressure difference for the present experiment. It was also found that the trial passive valve had difficulties in setting an actuation set point and vibrations noises and some fraction of the coolant was remained in CMT without effective use. A modification was proposed for resolving these problems by introducing an anti-closing mechanism. (author)

  16. MELCOR simulation of postulated severe accidents in OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seongn Yeon; Kim Sung Joong [Hanyang Univ., Seoul (Korea, Republic of); Kim, Hwan Yeol; Park, Jong Hwa [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Since the Fukushima accident in 2011, severe accidents of a nuclear power plant have been a target of big debate whether the defense in depth philosophy applied to current nuclear system is still vigorous enough to ensure the protection of the operators and the public. Thus an accurate prediction of severe accident has become a critical task for the nuclear engineers with reliable employment of Probabilistic Risk Analysis (PRA). According to a recent PRA result, Small Break Loss Of Coolant Accident (SBLOCA) without safety injection and Station Black Out (SBO) show high probability of proceeding to severe accidents. Thus, these accident scenarios need to be evaluated properly with reliable prediction tools. Song and Ahn analyzed SBO sequences in KSNP using MELCOR 1.8.5. Park and Song examined SBLOCA scenarios based on the PSA of KNSP using MAAP 4.06. Their studies utilized severe accident database. In continuation of the further analysis, several scenarios of postulated SBO and SBLOCA in OPR1000 are investigated using the severe accident database and MELCOR 1.8.6.

  17. Heat transfer and fluid flow aspects of fuel--coolant interactions. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M L

    1978-09-01

    A major portion of the safety analysis effort for the LMFBR is involved in assessing the consequences of a Hypothetical Core Disruptive Accident (HCDA). The thermal interaction of the hot fuel and the sodium coolant during the HCDA is investigated in two areas. A postulated loss of flow transient may produce a two-phase fuel at high pressures. The thermal interaction phenomena between fuel and coolant as the fuel is ejected into the upper plenum are investigated. A postulated transient overpower accident may produce molten fuel being released into sodium coolant in the core region. An energetic coolant vapor explosion for these reactor materials does not seem likely. However, experiments using other materials (e.g., Freon/water, tin/water) have demonstrated the possibility of this phenomenon.

  18. Severe accident testing of electrical penetration assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))

    1989-11-01

    This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs.

  19. Advances in Forecasting and Prevention of Resonances Between Coolant Acoustical Oscillations and Fuel Rod Vibrations

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, Konstantin Nicolaevich [NPP, NPEI, 14, Krasnokazarmennaya str. Moscow, 111250 (Russian Federation)

    2009-06-15

    the two-phase EFCPO calculations. Paper presents the results of EFCPO calculation. The result analysis shows that coolant acoustic characteristics depends from the aggregative state of coolant and also depends from the time interval after emergency situation with loss-of-coolant beginning. The essential differences of EFCPO in similar acoustic elements were revealed. As the coolant acoustic oscillations are one of the main reasons which control the vibration level in structure it is possible to provide conclusion that vibration characteristics of PSB equipment and real object also must be different. The calculation shows that Q-factor of VVER-1000 is much more than Q-factor of coolant system in PSB. Taking into consideration this difference it would be possible to forecast that intensity of coolant pressure oscillation in VVER-1000 will be several times more correspondingly data received by measurements on PSB. These phenomena could initiate appearance of burnout in reactor core of VVER-1000 while it is absent in PSB experiments. The same conclusion could be truth for any similar experimental installation, which are used for investigation of transient processes in passive cooling systems. As it follow from the analysis carried out conditions of geometric similarity of the PSB VVER are fulfilled only for the core. Lack of geometric similarity for the other elements leads to different EFCPO values. Taking into account the greater length of the real object tube system EFCPO of the real object are found to be lesser than EFCPO of the test facility and it results in larger coolant oscillation period under real object conditions. It leads to non-identical conditions, which define critical heat fluxes for a model and an object in pulsation conditions. On this account oscillation processes similarity criterion fulfillment becomes one of the most important condition when modeling emergency systems of the nuclear power plants and loss-of-coolant accidents analysis. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Rathbone, Bruce A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Morley, Shannon M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Stephens, John A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-05-01

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

  1. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  2. 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.)

  3. SUBSTATIONS OF DISTRICT HEATING SYSTEMS WITH PULSE COOLANT CIRCULATION

    Directory of Open Access Journals (Sweden)

    Andrey N. Makeev

    2017-01-01

    Full Text Available Abstract. Objectives The aim of the study is to generalise the results of the application of technologies and means for organising pulse coolant flow within a district heating system in order to increase its energy efficiency based on the organisation of local hydraulic shocks and the subsequent use of their energy to ensure the purification of heat energy equipment, intensify the heat transfer process and realise the possibility of transforming the available head from one hydraulic circuit to another. Methods Substations connecting the thermal power installations of consumers with heat networks via dependent and independent schemes are analytically generalised. The use of pulse coolant circulation is proposed as a means of overcoming identified shortcomings. Results Principal schemes of substations with pulse coolant circulation for dependent and independent connection of thermal power installations are detailed. A detailed description of their operation is given. The advantages of using pulse coolant circulation in substations are shown. The materials reflecting the results of the technical implementation and practical introduction of this technology are presented. Conclusion Theoretical analysis of the operation of the basic schemes of substations with pulse coolant circulation and the results of their practical application, as well as the materials of scientific works devoted to the use of the energy of a hydraulic impact and the study of the effect of pulse coolant flow on thermal and hydrodynamic processes, have yielded a combination of factors reflecting technical and economic rationality of application of pulse coolant circulation. 

  4. CONSEQUENCES FOR HEALTH AFTER THE CHERNOBYL ACCIDENT: MAIN RESULTS AND UNSOLVED PROBLEMS

    Directory of Open Access Journals (Sweden)

    A. K. Gus’kova

    2011-01-01

    Full Text Available Data of researches of the consequences for health after the Chernobyl accident of 1986 are generalized. All these years all over the world and especially in our country the basic parameters were studied describing type and the reason of the accident, doses levels for various groups of persons and a condition of their physical and sincere health. Accumulation of the extensive information allows returning to the initial concepts which have arisen directly after the accident, to estimate critically reliability accepted at that time criteria and adequacy of measures for overcoming and minimization of the consequences of the accident for health. In a basis of an assessment of the exposure levels and possible consequences for health in early timeframes have been put the information on the capacity of doses scale-radiation on various distances from the damaged reactor both total activity and structure of emission of radioactive substances. Three basic groups of the persons involved in the emergency with a various combination of risk factors for their health are allocated: the personnel of emergency changes, participants of liquidation of the accident consequences, the population of emergency emission zones. Consequences for health for these groups and principles of the further supervision over them are estimated. The increase of leukemia among the reasons for death (5 of 21 attracts attention. The group of patients transferred acute radiation syndrome in connection with the Chernobyl accident differs for the reasons for death in the remote timeframes from participants of other radiating accidents. By retrospective consideration there is a question on a possibility of the insufficient account of toxic influences accompanying the exposure. Comparison of urgent decisions and the retrospective analysis of assessments during the early period of accident allow considering these early decisions adequate to volume of the information available during this

  5. Analysis of simulation results of damaged nuclear fuel accidents at NPPs with shell-type nuclear reactors

    Directory of Open Access Journals (Sweden)

    Igor L. Kozlov

    2015-03-01

    Full Text Available Lessons from the accident at the Fukushima Daiichi NPP made it necessary to reevaluate and intensificate the work on modeling and analyzing various scenarios of severe accidents with damage to the nuclear fuel in the reactor, containment and spent nuclear fuel storage pool with the expansion of the primary initiating event causes group listing. Further development of computational tools for modeling the explosion prevention criteria as to steam and gas mixtures, considering the specific thermal-hydrodynamic conditions and mechanisms of explosive situations arrival at different stages of a severe accident development, is substantiated. Based on the analysis of the known shell-type nuclear reactors accidents results the explosion safety thermodynamic criteria are presented, the parameters defining the steam and gas explosions conditions are found, the need to perform the further verification and validation of deterministic codes serving to simulate general accident processes behavior as well as phase-to-phase interaction calculated dependencies is established. The main parameters controlling and defining the criteria explosion safety effective regulation areas and their optimization conditions are found.

  6. Proceedings of the workshop on severe accident research, Japan (SARJ-99)

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, Kazuichiro [ed.

    2000-11-01

    The Workshop on Severe Accident Research, Japan (SARJ-99) was taken place at Hotel Lungwood on November 8-10, 1999, and attended by 156 participants from 12 countries. A total of 46 papers, which covered wide areas of severe accident research both in experiments and analyses, such as fuel/coolant interaction, accident analysis and modeling, in-vessel phenomena, accident management, fission product behavior, research reactors, ex-vessel phenomena, and structural integrity, were presented. The panel discussion titled 'Link of Severe Accident Research Results to Regulation: Current Status and Future Perspective' was successfully conducted, and the wide variety of opinions and views were exchanged among panelists and experts. (J.P.N.)

  7. [Biliary tract diseases in persons suffering as a result of the accident at the Chernobyl Atomic Electric Power Station].

    Science.gov (United States)

    Komarenko, D I; Soboleva, L P; Kadiuk, E N; Glukhen'kiĭ, E V; Nosach, E V

    1999-07-01

    A retrospective analysis was performed of case histories and of results of sonographic investigations in liquidators of the Chernobyl accident suffering from chronic abnormalities of the biliary ducts. Patients with cholecystitis were studied for the biochemical composition of their bile. The incidence of the gallbladder disorders (chronic cholecystitis, angiocholitis, dyskinesias of the biliary ducts) has not changed much over the last 10 years having elapsed since the accident. The biochemical composition of bile was found to have been changed to a greater extent in the liquidators of the Chernobyl accident than it was in those having avoided danger of exposure to ionizing radiation. Mechanisms of origination of cholelithiasis are discussed on the basis of investigations designed to study biochemical properties of bile and findings secured with the aid of the ultrasound techniques.

  8. 2011 Fukushima Dai-ichi nuclear power plant accident: summary of regional radioactive deposition monitoring results.

    Science.gov (United States)

    Hirose, Katsumi

    2012-09-01

    After the Great East Japan Earthquake and resulting Tsunami on March 11, 2011, serious accident of the Fukushima Dai-ichi Nuclear Power Plant has been occurred. Huge amounts of radionuclides were released in atmosphere and ocean. Japanese prefectural governments have carried out environmental radioactivity monitoring; external dose rate, radioactivity measurements in environmental samples and others. Since March 18, 2011, daily and monthly deposition samples were collected in 45 stations covering Japanese Islands and radionuclides in the deposition samples were determined. We summarize radioactive deposition data reported by Japanese Government and study the depositional behaviors of the Fukushima-derived radionuclides. The results revealed that Fukushima-derived radioactive cloud dominantly affected in the central and eastern part of Honshu-Island, although it affected all of Japanese land area and also western North Pacific. The temporal change of the Fukushima-derived (137)Cs revealed that the apparent atmospheric residence time of the Fukushima-derived (137)Cs in sites within 300 km from the Fukushima Dai-ichi NPPis about 10 d.

  9. Competing Sentence Test (CST Results in Patients With Cerebrovascular Accident (CVA

    Directory of Open Access Journals (Sweden)

    Shoreh Jalaei

    2003-09-01

    Full Text Available Objective: As the cerebrovascular diseases (CVD are among the most common anomalies that are directly or indirectly affecting the auditory cortex, studying in this area is important. We tried to evaluate the function of CANS in a group of 50-70 years old cerebrovascular accident (CVA patients with no hearing complaint by competing sentences test (CST as a dichotic speech test. Method and Material: This cross-sectional analytical study was conducted at Loghman-e-Hakim hospital between 22, august 1999 and 20, July 2000 in a group of 30 normal 50-70 years old persons and a group of 40 patients with CVA including 30 patients with abnormal CT scan and 10 patients with normal CT scan. Results: The results point out the mean scores of CST in the normal group were at normal range (80-100% in both ears. And mean scores of CST decreased in the patients groups. There were significant difference between mean scores of CST in the patients with CVA and control groups. Conclusion: CST seems to be a valuable test in the CANS test batteries for the cerebrovascular diseases.

  10. Sandia National Laboratories results for the 2010 criticality accident dosimetry exercise, at the CALIBAN reactor, CEA Valduc France.

    Energy Technology Data Exchange (ETDEWEB)

    Ward, Dann C.

    2011-09-01

    This document describes the personal nuclear accident dosimeter (PNAD) used by Sandia National Laboratories (SNL) and presents PNAD dosimetry results obtained during the Nuclear Accident Dosimeter Intercomparison Study held 20-23 September, 2010, at CEA Valduc, France. SNL PNADs were exposed in two separate irradiations from the CALIBAN reactor. Biases for reported neutron doses ranged from -15% to +0.4% with an average bias of -7.7%. PNADs were also exposed on the back side of phantoms to assess orientation effects.

  11. ACCIDENT AT «FUKUSHIMA-I» NPP: FIRST RESULTS OF EMERGENCY RESPONSE. REPORT 1: GENERAL INFORMATION ABOUT THE ACCIDENT AND RADIATION SITUATION

    Directory of Open Access Journals (Sweden)

    G. G. Onischenko

    2011-01-01

    Full Text Available The article presents preliminary assessment of the accident scale, level of its radiation effect on the population and response of national authorities of various countries for the population radiation protection provisions. High levels of the environmental radioactive contamination demanded the application of complex measures for the radiation protection of the population from adjacent territories of Japan. Radionuclides from the region of damaged NPP can get to the other countries by means of the long-distant air and sea water mass transfer. Specific activity of cesium radionuclides in the sea biota can reach the level recommended by the Codex Alimentarius Commission as the safe level for the international trade or exceed it. Significant radioactive contamination of the territories of other countries resulting from the «Fukushima-I» NPP accident did not occur. Many world countries applied measures of the Japanese food products import control on the base of the radiation monitoring data. These prohibitions are consequently remitted following the radiation situation improvement.

  12. The severe accident research programme PHEBUS F.P.: First results and future tests

    Energy Technology Data Exchange (ETDEWEB)

    Schwarz, M. [Institut de Protection et de Surete Nucleaire IPSN, Saint Paul Lez Durance (France); Hardt, P. von der [Joint Research Centre - Safety Technology Institute, Saint Paul Lez Durance (France)

    1996-03-01

    PHEBUS FP is an international programme, managed by the French Institut de Protection et de Surete Nucleaire, Electricite de France and the European Commission in close collaboration with the USNRC (US), COG (Canada), NUPEC and JAERI (Japan) and KAERI (South Korea). Its objective is to investigate through a series of in-pile integral experiments, key phenomena involved in LWR severe accident such as the degradation of core materials up to molten pool, the subsequent release of fission products and of structural materials, their transport in the cooling system and their deposition in the containment with a special emphasis on the volatility of iodine. After a general programme description, the paper focuses on the status of analysis of the first test FPT-0, which involved trace irradiated fuel and which has shown some quite unexpected results regarding fuel degradation and iodine behaviour, and on the upcoming test FPT-1 which will use irradiated fuel. The status of the preparation of the remaining tests of the programme is also presented.

  13. RADIATION-HYGIENIC AND MEDICAL CONSEQUENCES OF THE СHERNOBYL ACCIDENT: RESULTS AND PROGNOSIS

    Directory of Open Access Journals (Sweden)

    G. G. Onischenko

    2011-01-01

    Full Text Available An article is devoted to the analysis of the radiation situation in the dynamics during the years since the accident at the Chernobyl nuclear power plant in 1986. Data on the scope of activities fulfilled for the assessment of the territories radioactive contamination levels and foodstuffs contamination levels, on the values of the exposure doses for the population living on the contaminated territories, on the medical and socio-psychological consequences of the Chernobyl accident is presented. Basic norms and principles, used during the protective measures development and introduction, are considered, their effectiveness is demonstrated. Mistakes emerged during protective measures implementation are analyzed, the prognosis of the population exposure dose values for the 70-year period since the accident and main directions of activities for the contaminated territories remediation and normal life conditions restoration for the population at these territories are presented.

  14. Enhancing resistance to burnout via coolant chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Tu, J. P.; Dinh, T. N.; Theofanous, T. G. [Univ. of California, Santa Barbara (United States)

    2003-07-01

    Boiling Crisis (BC) on horizontal, upwards-facing copper and steel surfaces under the influence of various coolant chemistries relevant to reactor containment waters is considered. In addition to Boric Acid (BA) and TriSodium Phosphate (TSP), pure De-Ionized Water (DIW) and Tap Water (TW) are included in experiments carried out in the BETA facility. The results are related to a companion paper on the large scale ULPU facility.

  15. Analysis on the severe accidents in KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myoung Jae; Cheong, Y. H.; Choi, Y. S.; Cheon, E. J. [PlaGen, Seoul (Korea, Republic of)

    2003-11-15

    The establishment of regulatory and approval systems for KSTAR (Korea Superconducting Tokamak Advanced Research) has been demanded as the facility is targeted to be completed in the year of 2005. Such establishment can be achieved by performing adequate and in-depth analyses on safety issues covering radiological and chemical hazard materials, radiation protection, high vacuum, very low temperature, etc. The loss of coolant accidents and the loss of vacuum accident in fusion facilities have been introduced with summary of simulation results that were previously reported for ITER and JET. Computer codes that are actively used for accident simulation research are examined and their main features are briefly described. It can be stated that the safety analysis is indispensable to secure the safety of workers and individual members of the public as well as to establish the regulatory and approval systems for KSTAR tokamak.

  16. Preliminary Modeling of Accident Tolerant Fuel Concepts under Accident Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle A.; Hales, Jason D.

    2016-12-01

    The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. Thus, the United States Department of Energy through its NEAMS (Nuclear Energy Advanced Modeling and Simulation) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is funded for a three-year period. The purpose of the HIP is to perform research into two potential accident tolerant concepts and provide an in-depth report to the Advanced Fuels Campaign (AFC) describing the behavior of the concepts, both of which are being considered for inclusion in a lead test assembly scheduled for placement into a commercial reactor in 2022. The initial focus of the HIP is on uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (INL, LANL, and ANL) a comprehensive mulitscale approach to modeling is being used including atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. In this paper, we present simulations of two proposed accident tolerant fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. The simulations investigate the fuel performance response of the proposed ATF systems under Loss of Coolant and Station Blackout conditions using the BISON code. Sensitivity analyses are completed using Sandia National Laboratories’ DAKOTA software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). Early results indicate that each concept has significant advantages as well as areas of concern. Further work is required prior to formulating the proposition report for the Advanced Fuels Campaign.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2013-12-15

    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.)

  18. A Statistical Description of the Types and Severities of Accidents Involving Tractor Semi-Trailers, Updated Results for 1992-1996

    Energy Technology Data Exchange (ETDEWEB)

    BLOWER,DANIEL F.; CLAUSS,DAVID B.

    1999-10-01

    This report provides a statistical description of the types and severities of tractor semi-trailer accidents involving at least one fatality. The data were developed for use in risk assessments of hazardous materials transportation. A previous study (SAND93-2580) reviewed the availability of accident data, identified the TIFA (Trucks Involved in Fatal Accidents) as the best source of accident data for accidents involving heavy trucks, and provided statistics on accident data collected between 1980 and 1990. The current study is an extension of the previous work and describes data collected for heavy truck accidents occurring between 1992 and 1996. The TIFA database created at the University of Michigan Transportation Research Institute was extensively utilized. Supplementary data on collision and fire severity, which was not available in the TIFA database, were obtained by reviewing police reports and interviewing responders and witnesses for selected TEA accidents. The results are described in terms of frequencies of different accident types and cumulative distribution functions for the peak contact velocity, rollover skid distance, effective fire temperature, fire size, fire separation, and fire duration.

  19. A Statistical Description of the Types and Severities of Accidents Involving Tractor Semi-Trailers, Updated Results for 1992-1996

    Energy Technology Data Exchange (ETDEWEB)

    BLOWER,DANIEL F.; CLAUSS,DAVID B.

    1999-10-01

    This report provides a statistical description of the types and severities of tractor semi-trailer accidents involving at least one fatality. The data were developed for use in risk assessments of hazardous materials transportation. A previous study (SAND93-2580) reviewed the availability of accident data, identified the TIFA (Trucks Involved in Fatal Accidents) as the best source of accident data for accidents involving heavy trucks, and provided statistics on accident data collected between 1980 and 1990. The current study is an extension of the previous work and describes data collected for heavy truck accidents occurring between 1992 and 1996. The TIFA database created at the University of Michigan Transportation Research Institute was extensively utilized. Supplementary data on collision and fire severity, which was not available in the TIFA database, were obtained by reviewing police reports and interviewing responders and witnesses for selected TEA accidents. The results are described in terms of frequencies of different accident types and cumulative distribution functions for the peak contact velocity, rollover skid distance, effective fire temperature, fire size, fire separation, and fire duration.

  20. NONUNIFORMITIES OF TWO-PHASE COOLANT DISTRIBUTION IN A HEAT GENERATING PARTICLES BED

    Directory of Open Access Journals (Sweden)

    V. V. Sorokin

    2014-01-01

    Full Text Available Sufficient atomic power generation safety increase may be done with microfuel adapting to reactor plants with water coolant. Microfuel particle is a millimeter size grain containing fission material core in a protecting coverage. The coverage protects fuel contact with coolant and provides isolation of fission products inside. Well thermophysical properties of microfuel bed in a direct contact with water coolant excludes fuel overheating when accidents. Microfuel use was suggested for a VVER, а direct flow reactor for superheat steam generation, a reactor with neutron spectra adjustment by the steam partial content varying in the coolant.Nonuniformities of two-phase coolant distribution in a heat generating particles bed are predicted by calculations in this text. The one is due to multiple-valuedness of pressure drop across the bed on the steam quality dependency. The nonuniformity decreases with flow rate and particle size growths absolute pressure diminishing while porosity effect is weak. The worse case is for pressure quality of order of one. Some pure steam filled pores appears parallel to steam water mixture filled pores, latter steam quality is less than the mean of the bed. Considering this regime for the direct flow reactor for superheat steam generation we predict some water drops at the exit flow. The two-phase coolant filtration with subcooled water feed is unstable to strong disturbance effects are found. Uniformity of two-phase coolant distribution is worse than for one-phase in the same radial type reactor.

  1. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A., E-mail: anlafuente@etsii.upm.e [ETSII-UPM, c/Jose Gutierrez Abascal, 2, 28006 Madrid (Spain); Piera, M. [ETSII:UNED, c/Juan del Rosal, 12, 28040 Madrid (Spain)

    2010-06-15

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F{sub 2}Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F{sub 2}Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void

  2. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition

  3. Development and Validation of Accident Models for FeCrAl Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, Jason Dean [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    The purpose of this milestone report is to present the work completed in regards to material model development for FeCrAl cladding and highlight the results of applying these models to Loss of Coolant Accidents (LOCA) and Station Blackouts (SBO). With the limited experimental data available (essentially only the data used to create the models) true validation is not possible. In the absence of another alternative, qualitative comparisons during postulated accident scenarios between FeCrAl and Zircaloy-4 cladded rods have been completed demonstrating the superior performance of FeCrAl.

  4. Microstructural analysis of MTR fuel plates damaged by a coolant flow blockage

    Science.gov (United States)

    Leenaers, A.; Joppen, F.; Van den Berghe, S.

    2009-10-01

    In 1975, as a result of a blockage of the coolant inlet flow, two plates of a fuel element of the BR2 reactor of the Belgian Nuclear Research Centre (SCK•CEN) were partially melted. The fuel element consisted of Al-clad plates with 90% 235U enriched UAl x fuel dispersed in an Al matrix. The element had accumulated a burn up of 21% 235U before it was removed from the reactor. Recently, the damaged fuel plates were sent to the hot laboratory for detailed PIE. Microstructural changes and associated temperature markers were used to identify several stages in the progression to fuel melting. It was found that the temperature in the center of the fuel plate had increased above 900-950 °C before the reactor was scrammed. In view of the limited availability of such datasets, the results of this microstructural analysis provide valuable input in the analysis of accident scenarios for research reactors.

  5. Experimental and analytical studies of melt jet-coolant interactions: a synthesis

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Bui, V.A.; Nourgaliev, R.R.; Green, J.A.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-01-01

    Instability and fragmentation of a core melt jet in water have been actively studied during the past ten years. Several models, and a few computer codes, have been developed. However, there are, still, large uncertainties, both, in interpreting experimental results and in predicting reactor-scale processes. Steam explosion and debris coolability, as reactor safety issues, are related to the jet fragmentation process. A better understanding of the physics of jet instability and fragmentation is crucial for assessments of fuel-coolant interactions (FCIs). This paper presents research, conducted at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS), Stockholm, concerning molten jet-coolant interactions, as a precursor for premixing. First, observations were obtained from scoping experiments with simulant fluids. Second, the linear perturbation method was extended and applied to analyze the interfacial-instability characteristics. Third, two innovative approachs to CFD modeling of jet fragmentation were developed and employed for analysis. The focus of the studies was placed on (a) identifying potential factors, which may affect the jet instability, (b) determining the scaling laws, and (c) predicting the jet behavior for severe accidents conditions. In particular, the effects of melt physical properties, and the thermal hydraulics of the mixing zone, on jet fragmentation were investigated. Finally, with the insights gained from a synthesis of the experimental results and analysis results, a new phenomenological concept, named `macrointeractions concept of jet fragmentation` is proposed. (author)

  6. System approach in the investigation of coolant parametrical oscillations in passive safety injection systems (PSIS)

    Energy Technology Data Exchange (ETDEWEB)

    Proskouriakov, K.N. [Moskovskij Ehnergeticheskij Inst., Moscow (Russian Federation)

    2001-07-01

    The use of thermal-hydraulic computer codes is an important part of the work programme for activities in the field of nuclear power plants (NPP) Safety Research as it will enable to define better the test configuration and parameter range extensions and to extrapolate the results of the small scale experiments towards full scale reactor applications. The CATHARE2, RELAP5, the WCOBRA/TRAC, and APROS codes are the estimate thermal hydraulic codes for the evaluation of large and small break loss of coolant accidents (LOCA). The relatively good agreement experimental data with the calculations have been presented. There was shown also some big mistakes in predicting distribution of flow when two phase are present. Model of parametrical oscillation (P.O.) worked out gives explanation for flow oscillations and indicates that the phenomenon of P.O. appears under certain combination of thermal-hydraulic parameters and structure of heat-removal system. (orig.)

  7. Influence of coolant temperature and pressure on destructive forces at fuel failure in the NSRR experiment

    Energy Technology Data Exchange (ETDEWEB)

    Kusagaya, Kazuyuki [Global Nuclear Fuel - Japan Co., Ltd., Yokosuka, Kanagawa (Japan); Sugiyama, Tomoyuki; Nakamura, Takehiko; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-01-01

    In order to design a new experimental capsule to be used in the NSRR (Nuclear Safety Research Reactor) experiment with the temperature and pressure conditions in a typical commercial BWR, coolant temperature and pressure influence is estimated for destructive forces during fuel rod failure in the experiment simulating reactivity-initiated accident (RIA). Considering steam property dependence on temperature and pressure, it is qualitatively shown that the destructive forces in the BWR operation condition are smaller than those in the room temperature and atmospheric pressure condition. Water column velocity, which determines impact by water hammer, is further investigated quantitatively by modeling the experimental system and water hammer phenomenon. As a result, the maximum velocity of the water column in the BWR operation conditions is calculated to be only about 10% of that in the room temperature and atmospheric pressure condition. (author)

  8. Analytical study on creep behavior of a tube of coolant piping system in nuclear power plant. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Miyazaki, Noriyuki [Kyushu Univ., Fukuoka (Japan); Hagihara, Seiya [Saga Univ., Saga (Japan); Chino, Eiichi; Maeda, Akio [MRI Systems Inc., Tokyo (Japan); Maruyama, Yu; Hashimoto, Kazuichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-10-01

    During severe accident of a light water reactor (LWR), reactor coolant piping would be damaged when the piping is subjected to high internal pressure and very high temperature due to heat transfer from high-temperature gas and decay heat from wall-deposited fission product (FP), both from degraded core. In such a case, high-temperature fast creep deformation could be the main cause for the pipe failure. For the evaluation of piping integrity during severe accidents, a method to predict such high-temperature fast creep deformation should be developed, using a creep constitutive equation considering tertiary creep behavior which has not been considered well in the pipe failure analyses. In this study, a creep constitutive equation was developed first based on the Kachanov-Ravotnov isotropic damage rule that considers the tertiary creep behavior. JAERI creep tensile test data for both nuclear-grade cold-drawn SUS316N and hot-extruded SUS316 materials were used to determine coefficients of the developed constitutive equation. Using the developed constitutive equation, finite element analyses were performed for local creep deformation of coolant piping under two temperature conditions: uniform temperature and temperature gradient. The analytical results indicated the damage variable being integrated following the evolution of creep damage can indicate pipe wall internal damage condition quantitatively. The damage variable was confirmed further to be able to reproduce the observation in JAERI piping failure tests, that is, pipe failure from the wall outside. (author)

  9. 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.

  10. Experimental Setup for Reflood Quench of Accident Tolerant Fuel Claddings

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan; Lee, Kwan Geun; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The concept of accident tolerant fuel (ATF) is a solution to suppress the hydrogen generation in loss of coolant accident (LOCA) situation without safety injection, which was the critical incident in the severe accident in the Fukushima. The changes in fuel and cladding materials may cause a significant difference in reactor performance in long term operation. Properties in terms of material science and engineering have been tested and showed promising results. However, numerous tests are still required to ensure the design performance and safety. Thermal hydraulic tests including boiling and quenching are partly confirmed, but not yet complete. We have been establishing the experimental setup to confirm the properties in the terms of thermal hydraulics. Design considerations and preliminary tests are introduced in this paper. An experimental setup to test thermal hydraulic characteristics of new ATF claddings are established and tested. The W heater set inside the cladding is working properly, exceeding 690 W/m linear power with thermocouples and insulating ceramic sheaths inside. The coolant injection control was also working in good conditions. The setup is about to complete and going to simulate quenching behavior of the ATF in the LOCA situation.

  11. Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-11-15

    The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

  12. 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)

  13. Simulation of 3D Flow in Turbine Blade Rows including the Effects of Coolant Ejection

    Institute of Scientific and Technical Information of China (English)

    Jian-Jun LIU; Bai-Tao AN; Yun-Tao ZENG

    2008-01-01

    This paper describes the numerical simulation of three-dimensional viscous flows in air-cooled turbine blade rows with the effects of coolant ejection. A TVD Navier-Stokes flow solver incorporated with Baldwin-Lomax turbulence model and multi-grid convergence acceleration algorithm are used for the simulation. The influences of coolant ejection on the main flow are accounted by volumetric coolant source terms. Numerical results for a four-stage turbine are presented and discussed.

  14. How a nuclear power plant accident influences acceptance of nuclear power: results of a longitudinal study before and after the Fukushima disaster.

    Science.gov (United States)

    Visschers, Vivianne H M; Siegrist, Michael

    2013-02-01

    Major nuclear accidents, such as the recent accident in Fukushima, Japan, have been shown to decrease the public's acceptance of nuclear power. However, little is known about how a serious accident affects people's acceptance of nuclear power and the determinants of acceptance. We conducted a longitudinal study (N= 790) in Switzerland: one survey was done five months before and one directly after the accident in Fukushima. We assessed acceptance, perceived risks, perceived benefits, and trust related to nuclear power stations. In our model, we assumed that both benefit and risk perceptions determine acceptance of nuclear power. We further hypothesized that trust influences benefit and risk perceptions and that trust before a disaster relates to trust after a disaster. Results showed that the acceptance and perceptions of nuclear power as well as its trust were more negative after the accident. In our model, perceived benefits and risks determined the acceptance of nuclear power stations both before and after Fukushima. Trust had strong effects on perceived benefits and risks, at both times. People's trust before Fukushima strongly influenced their trust after the accident. In addition, perceived benefits before Fukushima correlated with perceived benefits after the accident. Thus, the nuclear accident did not seem to have changed the relations between the determinants of acceptance. Even after a severe accident, the public may still consider the benefits as relevant, and trust remains important for determining their risk and benefit perceptions. A discussion of the benefits of nuclear power seems most likely to affect the public's acceptance of nuclear power, even after a nuclear accident.

  15. Molten Fuel-Coolant Interactions induced by coolant injection into molten fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Yamano, Norihiko; Maruyama, Yu; Moriyama, Kiyofumi; Yang, Y.; Sugimoto, Jun [Severe Accident Research Laboratory, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    1999-07-01

    To investigate Molten Fuel-Coolant Interactions (MFCIs) in various contact geometries, an experimental program, called MUSE (MUlti-configurations in Steam Explosions), has been initiated under the ALPHA program at JAERI in Japan. The first series of MUSE test has been focused on the coolant injection (CI) and stratified modes of FCIs using water as coolant and molten thermite as molten fuel. The effects of water jet subcooling, jet dynamics, jet shape and system constraint on FCIs energetic in these modes were experimentally investigated by precisely measuring their mechanical energy release in the MUSE facility. It was observed that measured mechanical energy increased with increasing of jet subcooling in a weakly constraint system but decreased in a strongly constraint system. FCI energetic also increased with increasing of water jet velocity. These results suggested that the penetration and dispersion phenomena of a water jet inside a melt determined the mixing conditions of FCIs in these contact modes and consequently played important roles on FCI energetics. To understand fundamental physics of these phenomena and possible mixing conditions in the MUSE tests, a set of visualization tests with several pairs of jet-pool liquids in non-boiling and isothermal conditions were carried out. Numerical simulations of a water jet penetrating into a water pool at non-boiling conditions showed similar behaviors to those observed in the visualization tests. (author)

  16. Breakup of jet and drops during premixing phase of fuel coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Haraldsson, Haraldur Oskar

    2000-05-01

    During the course of a hypothetical severe accident in a light water reactor, molten liquid may be introduced into a volatile coolant, which, under certain conditions, results in explosive interactions. Such fuel-coolant interactions (FCI) are characterised by an initial pre-mixing phase during which the molten liquid, metallic or oxidic in nature, undergoes a breakup (fragmentation) process which significantly increase the area available for melt-coolant contact, and thus energy transfer. Although substantial progress in the understanding of phenomenology of the FCI events has been achieved in recent years, there remain uncertainties in describing the primary and secondary breakup processes. The focus of this work is on the melt jet and drop breakup during the premixing phase of FCI. The objectives are to gain insight into the premixing phase of the FCI phenomena, to determine what fraction of the melt fragments and determine the size distribution. The approach is to perform experiments with various simulant materials, at different scales, different conditions and with variation of controlling parameters affecting jet and drop breakup processes. The analysis approach is to investigate processes at different level of detail and complexity to understand the physics, to rationalise experimental results and to develop and validate models. In the first chapter a brief introduction and review of the status of the FCI phenomena is performed. A review of previous and current experimental projects is performed. The status of the experimental projects and major findings are outlined. The first part of the second chapter deals with experimental investigation of jet breakup. Two series of experiments were performed with low and high temperature jets. The low temperature experiments employed cerrobend-70 as jet liquid. A systematic investigation of thermal hydraulic conditions and melt physical properties on the jet fragmentation and particle debris characteristics was

  17. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  18. Influence of coolant motion on structure of hardened steel element

    Directory of Open Access Journals (Sweden)

    A. Kulawik

    2008-08-01

    Full Text Available Presented paper is focused on volumetric hardening process using liquid low melting point metal as a coolant. Effect of convective motion of the coolant on material structure after hardening is investigated. Comparison with results obtained for model neglecting motion of liquid is executed. Mathematical and numerical model based on Finite Element Metod is described. Characteristic Based Split (CBS method is used to uncouple velocities and pressure and finally to solve Navier-Stokes equation. Petrov-Galerkin formulation is employed to stabilize convective term in heat transport equation. Phase transformations model is created on the basis of Johnson-Mehl and Avrami laws. Continuous cooling diagram (CTPc for C45 steel is exploited in presented model of phase transformations. Temporary temperatures, phases participation, thermal and structural strains in hardening element and coolant velocities are shown and discussed.

  19. Experimental study of high temperature particle dropping in coolant liquid

    Institute of Scientific and Technical Information of China (English)

    LI Tianshu; YANG Yanhua; LI Xiaoyan; HU Zhihua

    2007-01-01

    A series of experiments of the premixing stage of fuel-coolant interactions (FCI), namely the particles falling into water, were carried out. The force on the particles during the course of falling has been studied. The dropping character of hot particle was influenced by three main parameters, i.e., particle temperature, particle diameter and coolant subcooling that varied over a wide range. A high-speed camera recorded the falling speed of the particle and the moving curves were obtained. The experimental results showed that for the film boiling on the surface of particle and water, the temperature increase of either particle or coolant would slow down the particle falling velocity. The falling velocity of particle in small diameter is lower than that of the bigger particle. The present work can provide an experimental foundation for further investigation of high-speed transient evaporation heat transfer.

  20. Design and fabrication of magnetic coolant filter

    Science.gov (United States)

    Prashanth, B. N.

    2017-07-01

    Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

  1. Reactor coolant pump shaft seal behavior during blackout conditions

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Traffic Accidents on Slippery Roads

    DEFF Research Database (Denmark)

    Fonnesbech, J. K.; Bolet, Lars

    2014-01-01

    Police registrations from 65 accidents on slippery roads in normally Danish winters have been studied. The study showed: • 1 accident per 100 km when using brine spread with nozzles • 2 accidents per 100 km when using pre wetted salt • 3 accidents per 100 km when using kombi spreaders The results...

  3. Methods for analyzing the uncertainty of a reconstructed result in a traffic accident with interval and probabilistic traces.

    Science.gov (United States)

    Zou, Tiefang; Peng, Xulong; Wu, Wenguang; Cai, Ming

    2017-01-01

    In order to make the reconstructed result more reliable, a method named improved probabilistic-interval method was proposed to analyze the uncertainty of a reconstructed result in a traffic accident with probabilistic and interval traces. In the method, probabilistic traces are replaced by probabilistic sub-intervals firstly; secondly, these probabilistic sub-intervals and those interval traces will be combined to form many new uncertainty analysis problems with only interval traces; thirdly, the upper and lower bound of the reconstructed result and their probability were calculated in each new uncertainty analysis problem, and an algorithm was proposed to shorten the time taken for this step; finally, distribution functions of the upper and lower bound of the reconstructed result were obtained by doing statistic analysis. Through 2 numerical cases, results obtained from the proposed method were almost the same as results obtained from the Monte Carlo method, but the time taken for the proposed method was far less than the time taken for the Monte Carlo method and results obtained from the proposed method were more stable. Through applying the proposed method to a true vehicle-pedestrian accident, not only the upper and lower bound of the impact velocity (v) can be obtained; but also the probability that the upper bound and the lower bound of v falls in an arbitrary interval can be obtained; furthermore, the probability that the interval of v is less than an arbitrary interval can be obtained also. It is concluded that the proposed improved probabilistic-interval method is practical. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

  4. Influence of main variables modifications on accident transient based on AP1000-like MELCOR model

    Science.gov (United States)

    Malicki, M.; Pieńkowski, L.

    2016-09-01

    Analysis of Severe Accidents (SA) is one of the most important parts of nuclear safety researches. MELCOR is a validated system code for severe accident analysis and as such it was used to obtain presented results. Analysed AP1000 model is based on publicly available data only. Sensitivity analysis was done for the main variables of primary reactor coolant system to find their influence on accident transient. This kind of analysis helps to find weak points of reactor design and the model itself. Performed analysis is a base for creation of Small Modular Reactor (SMR) generic model which will be the next step of the investigation aiming to estimate safety level of different reactors. Results clearly help to establish a range of boundary conditions for main the variables in future SMR model.

  5. Technology of high temperature organic coolant

    Energy Technology Data Exchange (ETDEWEB)

    Makin, R.S.; Vorobei, M.P.; Kuprienko, V.A.; Starkov, V.A.; Tsykanov, V.A.; Checketkin, Y.V. [Research Institute of Atomic Reactors, Ulyanovsk (Russian Federation)

    1993-12-31

    Research has been performed on the problems related to the use of high temperature organic coolants in small and medium nuclear power plants. The work performed and also the experience of operating the ARBUS reactor confirmed the inherent safety features, reliability, and enhanced safety margins of the plants with this type of coolants. The advantages of this system and research highlights are presented.

  6. 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.

  7. Characteristics of maxillofacial injuries resulting from road traffic accidents – a 5 year review of the case records from Department of Maxillofacial Surgery in Katowice, Poland

    Directory of Open Access Journals (Sweden)

    Drugacz Jan

    2006-08-01

    Full Text Available Abstract Background In spite of employing numerous devices improving the safety in motor vehicles, traffic accidents are still among the main reasons of maxillofacial injuries. The maxillofacial injuries remain the serious clinical problem because of the specificity of this anatomical region. The knowledge of etiologic factors and mechanisms of injuries can be helpful in a satisfactory trauma prevention. The aim of this study was to find out the incidence and the pattern of maxillofacial injuries resulting from traffic accidents in the patients treated in the Department of Maxillofacial Surgery (Silesian Medical Academy in Katowice, Poland from January 2001 to December 2005. Methods The material consisted of 1024 case records of patients with maxillofacial injuries treated in the Maxillofacial Surgery Department of Silesian Medical Academy. The detailed analysis was carried out on the case records of 198 patients in the age of 3 to 68 with maxillofacial injuries resulting from traffic accidents. On the basis of data from a history, examination on admission, consultations and radiological examinations, patients' age and gender, we obtained the information on a pattern of injury and detailed description of an accident (the date and the time of an accident, the role of the patient in an accident. Results The traffic accidents were the cause of 19,93% maxillofacial injuries in the analyzed period of time. Most of the patients had injuries to the soft tissues of the face (22,21%, followed by tooth and alveolar process injuries (20,71% and mandibular fractures (18,69%. All the types of injuries were more common in men than in women. The majority of the patients were car drivers followed by car passengers, pedestrians, cyclists and motor cyclists. The peak age of the patients was between 18 to 25 years. The prevalent number of accidents resulting in injuries to this region took place in spring, especially between noon and 4 PM. Conclusion Our results

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  9. 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

    framework of the MELCOR analysis will be helpful to establish a basis for the severe accident mitigation strategy. As shown in Table 5, all postulated major scenarios were estimated resulting in the reduction of water inventory level and the mass of hydrogen generation. In the case of loss of cooling accident, time to water level reduction was obviously dependent of the decay heat load. It was revealed that the loss of inventory was dominant factor in the case of loss of coolant accident. With respect to the mass of hydrogen generation, massive amount of hydrogen was generated from the molten corium concrete reaction in MELCOR calculation. Through the MELCOR modeling developed in this study, it was found that MELCOR program has capabilities to predict entire progress of water level, fuel degradation, hydrogen generation, and concrete reaction.

  10. A passively-safe fusion reactor blanket with helium coolant and steel structure

    Energy Technology Data Exchange (ETDEWEB)

    Crosswait, K.M.

    1994-04-01

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.

  11. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T.; Grunwald, G.

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  12. Sensitivity Analysis of Core Damage from Reactor Coolant Pump Seal Leakage during Extended Loss of All AC Power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Da Hee; Kim, Min Gi; Lee, Kyung Jin; Hwang, Su hyun; Lee, Byung Chul [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Yoon, Duk Joo; Lee, Seung Chan [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-10-15

    In this study, in order to comprehend the Fukushima accident, the sensitivity analysis was performed to analyze the behavior of Reactor Coolant System (RCS) during ELAP using the RELAP5/MOD3.3 code. The Fukushima accident was caused by tsunami resulted in Station Black Out (SBO) followed by the reactor core melt-down and release of radioactive materials. After the accident, the equipment and strategies for the Extended Loss of All AC Power (ELAP) were recommended strongly. In this analysis, sensitivity studies for the RCP seal failure of the OPR1000 type NPP were performed by using RELAP5/MOD3.3 code. Six cases with different leakage rate of RCP seal were studied for ELAP with operator action or not. The main findings are summarized as follows: (1) Without the operator action, the core uncovery time is determined by the leakage rate of RCP seal. When the leakage rate per RCP seal are 5 gpm, 50 gpm, and 300 gpm respectively, the core uncovery time are 1.62 hr, 1.58 hr, and 1.29 hr respectively. Namely, If the leakage rate of RCP seal was much bigger, the uncover time of core would be shorter. (2) In case that the cooling by SG secondary side was performed using the TDAFP and SG ADV, the core uncovery time was significantly extended.

  13. Dynamic Analysis of Coolant Channel and Its Internals of Indian 540 MWe PHWR Reactor

    Directory of Open Access Journals (Sweden)

    N. Dharmaraju

    2008-01-01

    Full Text Available The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carries the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be going into operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and modeling studies have been carried out. A good correlation has been achieved between the results of experimental and analytical models. The operating life of a typical coolant channel typically ranges from 10 to 15 full-power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. Through the study of dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. A study has been also carried out to characterize the fuel vibration under different flow condition.

  14. Improvement of Measurement Accuracy of Coolant Flow in a Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Kim, Jong-Bum; Joung, Chang-Young; Ahn, Sung-Ho; Heo, Sung-Ho; Jang, Seoyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, to improve the measurement accuracy of coolant flow in a coolant flow simulator, elimination of external noise are enhanced by adding ground pattern in the control panel and earth around signal cables. In addition, a heating unit is added to strengthen the fluctuation signal by heating the coolant because the source of signals are heat energy. Experimental results using the improved system shows good agreement with the reference flow rate. The measurement error is reduced dramatically compared with the previous measurement accuracy and it will help to analyze the performance of nuclear fuels. For further works, out of pile test will be carried out by fabricating a test rig mockup and inspect the feasibility of the developed system. To verify the performance of a newly developed nuclear fuel, irradiation test needs to be carried out in the research reactor and measure the irradiation behavior such as fuel temperature, fission gas release, neutron dose, coolant temperature, and coolant flow rate. In particular, the heat generation rate of nuclear fuels can be measured indirectly by measuring temperature variation of coolant which passes by the fuel rod and its flow rate. However, it is very difficult to measure the flow rate of coolant at the fuel rod owing to the narrow gap between components of the test rig. In nuclear fields, noise analysis using thermocouples in the test rig has been applied to measure the flow velocity of coolant which circulates through the test loop.

  15. Visualization of Traffic Accidents

    Science.gov (United States)

    Wang, Jie; Shen, Yuzhong; Khattak, Asad

    2010-01-01

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

  16. Development of severe accident analysis code - A study on the molten core-concrete interaction under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Chang Hyun; Lee, Byung Chul; Huh, Chang Wook; Kim, Doh Young; Kim, Ju Yeul [Seoul National University, Seoul (Korea, Republic of)

    1996-07-01

    The purpose of this study is to understand the phenomena of the molten core/concrete interaction during the hypothetical severe accident, and to develop the model for heat transfer and physical phenomena in MCCIs. The contents of this study are analysis of mechanism in MCCIs and assessment of heat transfer models, evaluation of model in CORCON code and verification in CORCON using SWISS and SURC Experiments, and 1000 MWe PWR reactor cavity coolability, and establishment a model for prediction of the crust formation and temperature of melt-pool. The properties and flow condition of melt pool covering with the conditions of severe accident are used to evaluate the heat transfer coefficients in each reviewed model. Also, the scope and limitation of each model for application is assessed. A phenomenological analysis is performed with MELCOR 1.8.2 and MELCOR 1.8.3 And its results is compared with corresponding experimental reports of SWISS and SURC experiments. And the calculation is performed to assess the 1000 MWe PWR reactor cavity coolability. To improve the heat transfer model between melt-pool and overlying coolant and analyze the phase change of melt-pool, 2 dimensional governing equations are established using the enthalpy method and computational program is accomplished in this study. The benchmarking calculation is performed and its results are compared to the experiment which has not considered effects of the coolant boiling and the gas injection. Ultimately, the model shall be developed for considering the gas injection effect and coolant boiling effect. 66 refs., 10 tabs., 29 refs. (author)

  17. Computing Flows Of Coolants In Turbomachines

    Science.gov (United States)

    Meitner, P. L.

    1994-01-01

    Coolant Passage Flow (CPF) computer code developed to predict accurately coolant flow and heat transfer inside turbomachinery cooling passages (either radial or axial blading). Computes flow in one-inlet/one-outlet passage of any shape. Calculates rate of flow of coolant, temperature, pressure, velocity, and heat-transfer coefficients along passage. Integrates one-dimensional momentum and energy equations along defined flow path, taking into account change in area, addition or subtraction of mass, pumping, friction, and transfer of heat. Written in FORTRAN IV.

  18. SIMMER-III applications to fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

  19. Outline of the Desktop Severe Accident Graphic Simulator Module for OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    occurring during accident including core heatup, cladding oxidation and hydrogen generation, core melt progression, vessel failure, fission product release, transport and deposition, and containment failure. Output results are displayed in user friendly graphical format by using text-based (numerical) output of MAAP program.. Window-based simulator of VMAAP is designed to provide graphical displays of the results during the transient simulation so that the users can easily follow the plant dynamics. Figure 1 through 4 show an example of VMAAP graphic display for the reactor coolant system, reactor vessel, containment building, and plotting of important parameters. VMAAP is able to simulate various scenarios very easily and quickly from the input deck of the scenario database of the SARDB. Since hundreds of input decks for severe core damage scenarios are available in SARDB, the simulation for a user-defined scenario can be performed very quickly by using a sub-module of VMAAP Input-editor which is a window-based MAAPspecific input deck generation program. VMAAP consists of following sub-modules: - System menu and tool bar - Project view - Event summary - Interactive control - Parameter help view - Input editor - Reactor vessel view - Reactor coolant system view - Containment building view The plant model used in VMAAP module is oriented to severe accident phenomena and thus it can simulate the in-vessel and ex-vessel behavior for a severe accident. Even though it may not be compatible with the desire to have a best-estimate analysis of an ongoing event, it can be a supporting or supplementary measure to understand the trends of accident progression.

  20. Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Goldmann, Andrew S. (Texas A& M University, College Station, TX); Wagner, Kenneth C.; Powers, Dana Auburn; Ashbaugh, Scott G.; Longmire, Pamela

    2010-04-01

    As part of a Nuclear Regulatory Commission (NRC) research program to evaluate the impact of using mixed-oxide (MOX) fuel in commercial nuclear power plants, a study was undertaken to evaluate the impact of the usage of MOX fuel on the consequences of postulated severe accidents. A series of 23 severe accident calculations was performed using MELCOR 1.8.5 for a four-loop Westinghouse reactor with an ice condenser containment. The calculations covered five basic accident classes that were identified as the risk- and consequence-dominant accident sequences in plant-specific probabilistic risk assessments for the McGuire and Catawba nuclear plants, including station blackouts and loss-of-coolant accidents of various sizes, with both early and late containment failures. Ultimately, the results of these MELCOR simulations will be used to provide a supplement to the NRC's alternative source term described in NUREG-1465. Source term magnitude and timing results are presented consistent with the NUREG-1465 format. For each of the severe accident release phases (coolant release, gap release, in-vessel release, ex-vessel release, and late in-vessel release), source term timing information (onset of release and duration) is presented. For all release phases except for the coolant release phase, magnitudes are presented for each of the NUREG-1465 radionuclide groups. MELCOR results showed variation of noble metal releases between those typical of ruthenium (Ru) and those typical of molybdenum (Mo); therefore, results for the noble metals were presented for Ru and Mo separately. The collection of the source term results can be used as the basis to develop a representative source term (across all accident types) that will be the MOX supplement to NUREG-1465.

  1. Retention of PWR primary coolant trace elements by cation exchange resins during cold shutdown with oxygenation: modelling and experimental results for silver behavior; Retention des elements traces du fluide primaire des REP par les resines echangeuses de cations lors des mises en arret a froid avec oxygenation: modelisation et resultats experimentaux relatifs au comportement de l'argent

    Energy Technology Data Exchange (ETDEWEB)

    Elain, L.; Doury-Berthod, M. [CEA Saclay, INSTN, Institut National des Sciences et Techniques Nucleaires, 91 - Gif-sur-Yvette (France); Genin, J.B. [CEA Cadarache, Dir. de l' Energie Nucleaire (DEN), 13 - Saint-Paul-lez-Durance (France); Berger, M. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France)

    2004-07-01

    In order to minimize the radiochemical impact of the corrosion products on the operation of Pressurized Water Reactors, on-line purification of the primary coolant is carried out. The purification system arranged on the Chemical and Volume Control System is made up of mechanical filters and demineralizers packed with a mixed bed of cation and anion exchange resins. This paper proposes an update on the retention of primary coolant trace elements by the cation exchange resins of the demineralizers during cold shutdowns with oxygenation. The study is first of all devoted to the description of the concentration profiles of the various cation constituents which settle in the demineralizer during purification after oxygenation. For a number of trace elements, localized enrichment zones at the Li{sup +}/Ni(Il) exchange zone are expected to appear in the column. The case of silver is afterwards discussed in detail. Thermodynamic modelling shows that the theoretical retention volume of the metallic element and its degree of enrichment in the column are dependent on the basic composition of the primary coolant and the specific characteristics of the demineralizer cation exchanger. At the Ag{sup +} ion concentration expected in the reactor coolant after oxygenation (between 10{sup -8} mol.L{sup -1} and 10{sup -6} mol.L{sup -1}), the breakthrough of silver should be near-simultaneous with that of nickel. The experimental results, obtained in the laboratory and with a 'Mini-CVCS' pilot instrumentation recently used during the cold shutdown of Tricastin Unit 2,900 MWe PWR NPP, confirm the validity of these theoretical forecasts and enable new hypotheses to be advanced for explaining silver release from a demineralizer. (authors)

  2. Monte Carlo modeling of beta-radiometer device used to measure milk contaminated as a result of the Chernobyl accident

    Energy Technology Data Exchange (ETDEWEB)

    Khrutchinsky, A.; Kutsen, S. [Research Institute for Nuclear Problems, 11 Bobruiskaya Street, Minsk 220050 (Belarus); Minenko, V. [Belarusian Medical Academy of Post-Graduate Education, 3 Brovki Street, Minsk 220714 (Belarus); Zhukova, O. [Center of Radiation Control and Environment Monitoring, 110A Nezalezhnosti Avenue, Minsk 220023 (Belarus); Luckyanov, N.; Bouville, A. [DHHS, NIH, National Cancer Institute, Division of Cancer Epidemiology and Genetics, 6120 Executive Blvd, EPS 7100, Bethesda, MD 20892 (United States); Drozdovitch, V. [DHHS, NIH, National Cancer Institute, Division of Cancer Epidemiology and Genetics, 6120 Executive Blvd, EPS 7100, Bethesda, MD 20892 (United States)], E-mail: drozdovv@mail.nih.gov

    2009-06-15

    This paper presents results of Monte Carlo modeling of the beta-radiometer device with Geiger-Mueller detector used in Belarus and Russia to measure the radioactive contamination of milk after the Chernobyl accident. This type of detector, which is not energy selective, measured the total beta-activity of the radionuclide mix. A mathematical model of the beta-radiometer device, namely DP-100, was developed, and the calibration factors for the different radionuclides that might contribute to the milk contamination were calculated. The estimated calibration factors for {sup 131}I, {sup 137}Cs, {sup 134}Cs, {sup 90}Sr, {sup 144}Ce, and {sup 106}Ru reasonably agree with calibration factors determined experimentally. The calculated calibration factors for {sup 132}Te, {sup 132}I, {sup 133}I, {sup 136}Cs, {sup 89}Sr, {sup 103}Ru, {sup 140}Ba, {sup 140}La, and {sup 141}Ce had not been previously determined experimentally. The obtained results allow to derive the activity of specific radionuclides, in particular {sup 131}I, from the results of the total beta-activity measurements in milk. Results of this study are important for the purposes of retrospective dosimetry that uses measurements of radioactivity in environmental samples performed with beta-radiometer devices.

  3. Contamination of North- and Baltic Sea as a result of the accident of Chernobyl

    Energy Technology Data Exchange (ETDEWEB)

    Nies, H.; Wedekind, C.

    1987-01-01

    The input from 'Chernobyl' can be distinguished from other sources of artificial radionuclides, such as the reprocessing plants at La Hague and Sellafield, by its characteristic nuclide spectrum. The input occurred to quite different strength of activity in the different areas of the North Sea and the Baltic Sea; a rapid vertical dilution within the water column resulted in a rapid decrease of the initial activity concentration. The sediments are contaminated by the vertical transport of sinking suspended particulate matter, which adsorb the activity from the surrounding water.

  4. Application of methods to PLS and PCA analysis transient simulation results thermohydraulic; Aplicacion de los metodos PLS y PCA al analisis de resultados de simulaciones de transitorios termohidraulicos

    Energy Technology Data Exchange (ETDEWEB)

    Martorell, S.; Villamizar, M.; Sanchez, A. I.; Villanueva, J. F.; Carlos, S.; Serradell, V.; Pelayo, F.; Mendizabal, R.; Sol, I.

    2010-07-01

    This paper presents the application of regression partial least squares related to principal component regression, principal components analysis, and statistical analysis of the results obtained in the simulation of thermal-hydraulic transients with a code Best Estimate particularly relevant to a scenario LBLOCA (LargeBreak Loss-of Coolant Accident) in a PWR, considering the effect of uncertainties.

  5. Cervical spine injuries resulting from diving accidents in swimming pools: outcome of 34 patients.

    Science.gov (United States)

    Borius, Pierre-Yves; Gouader, Ismail; Bousquet, Philippe; Draper, Louisa; Roux, Franck-Emmanuel

    2010-04-01

    Cervical spine injuries after diving into private swimming pools can lead to dramatic consequences. We reviewed 34 patients hospitalized in our center between 1996 and 2006. Data was collected from their initial admission and from follow-up appointments. The injuries were sustained by young men in 97% (mean age 27) and the majority happened during the summer (88%). Fractures were at C5-C7 in 70%. American Spinal Injury Association class (ASIA) on admission was A for 8 patients, B for 4, C for 4, D for 1, and E for 17. There were 23 surgical spine stabilizations. Final ASIA class was A for 6 patients, B for 1, C for 3, D for 5, and E for 18. The mean duration of hospitalization was 21.3 days in our neurosurgical center (mean overall cost: 36,000 Euros/patient) plus 10.6 months in rehabilitation center for the 15 patients admitted who had an ASIA class A to C. Mean overall direct cost for a patient with class A is almost 300,000 Euros, compared to around 10,000 Euros for patients with class D and E. In addition, a profound impact on personal and professional life was seen in many cases including 11 divorces and 7 job losses. Dangerous diving into swimming pools can result in spinal injuries with drastic consequences, including permanent physical disability and a profound impact on socio-professional status. Moreover, there are significant financial costs to society. Better prevention strategies should be implemented to reduce the impact of this public health problem.

  6. Modeling of the TMI-2 (Three Mile Island Unit-2) accident with MELPROG/TRAC and calculation results for Phases 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Motley, F.E.; Jenks, R.P.

    1988-01-01

    Work has been performed to develop a Three Mile Island Unit-2 (TMI-2) simulation model for MELPROG/TRAC capable of predicting the observed plant behavior that took place during the accident of March 1979. A description of the TMI-2 plant model is presented and calculation results through 174 min of the accident are discussed. Using the ICBC boundary conditions, the calculation predicts pressurizer draining and core recovering prior to fuel-rod damage. A parametric calculation (reduced makeup flow) is currently underway and is in better agreement with the observed plant behavior. Efforts are underway to resolve current discrepancies and proceed with an accurate simulation through Phases 3 and 4 of the accident (174-227 min and 227-300 min, respectively). 13 refs., 11 figs., 2 tabs.

  7. RADIATION DOSE TO HUMAN AND NON-HUMAN BIOTA IN THE REPUBLIC OF KOREA RESULTING FROM THE FUKUSHIMA NUCLEAR ACCIDENT

    Directory of Open Access Journals (Sweden)

    DONG-KWON KEUM

    2013-02-01

    Full Text Available This paper describes the radiation doses to human and non-human biota in the Republic of Korea, as a result of the Fukushima nuclear accident. By using the measured airborne activity and ground deposition, the effective and thyroid doses of five human age groups (infant, 5 years, 10 years, 15 years and adult were estimated by the ECOSYS code, and the whole body absorbed dose rate of the eight Korean reference animals and plants (RAPs was estimated by the K-BIOTA (the Korean computer code to assess the risk of radioactivity to wildlife. The first-year effective and thyroid human doses ranged from 5.7E-5 mSv in the infant group to 2.0E-4 mSv in the 5 years group, and from 5.0E-4 mSv in the infant group to 3.4E-3 mSv in the 5 years group, respectively. The life-time (70 years effective and thyroid human doses ranged from 1.5E-4 mSv in the infant group to 3.0E-4 mSv in the 5 years group, and from 6.0E-4 mSv in the infant group to 3.5E-3 mSv in the 5 years group, respectively. The estimated maximum whole body absorbed dose rate to the Korean RAPs was 6.7E-7 mGy/d for a snake living in soil (terrestrial biota, and 2.0E-5 mGy/d for freshwater fish (aquatic biota, both of which were far less than the generic dose criteria to protect biota from ionizing radiation. Also, the screening level assessment for ERICA's (Environmental Risks from Ionizing Contaminants: Assessments and management limiting organisms showed that the risk quotient (RQ for the estimated maximum soil and water activity was significantly less than unity for both the terrestrial and freshwater organisms. Conclusively, the radiological risk of the radioactivity released into the environment by the Fukushima nuclear accident to the public and the non-human biota in the republic of Korea is considered negligible.

  8. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

  9. Assessment of Candidate Molten Salt Coolants for the NGNP/NHI Heat-Transfer Loop

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D. F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2006-06-30

    This report provides an assessment of candidate salts proposed as the coolant for the loop that shuttles heat from the Next Generation Nuclear Plant (NGNP) to the Nuclear Hydrogen Initiative (NHI) hydrogen-production plant. The physical properties most relevant for coolant service were reviewed, and key chemical factors that influence material compatibility were also analyzed for the purpose of screening candidate salts. A preliminary assessment of the cost of the raw materials required to produce the coolant is also presented. Salts that are suitable for use as the primary coolant in a high-temperature nuclear reactor were previously analyzed. Some of the fluoride salts identified in the previous study are also appropriate for consideration as the secondary coolant in a heat-transfer loop; therefore, results from the previous report are used in this document. However, alternative coolant salts (i.e., chlorides and fluoroborates) that were not considered in the previous report should be considered for service in the heat-transfer loop. These alternative coolants are considered in this report.

  10. Selection of an Alternate Biocide for the ISS Internal Thermal Control System Coolant, Phase 2

    Science.gov (United States)

    Wilson, Mark E.; Cole, Harold; Weir, Natalee; Oehler, Bill; Steele, John; Varsik, Jerry; Lukens, Clark

    2004-01-01

    The ISS (International Space Station) ITCS (Internal Thermal Control System) includes two internal coolant loops that utilize an aqueous based coolant for heat transfer. A silver salt biocide had previously been utilized as an additive in the coolant formulation to control the growth and proliferation of microorganisms within the coolant loops. Ground-based and in-flight testing demonstrated that the silver salt was rapidly depleted, and did not act as an effective long-term biocide. Efforts to select an optimal alternate biocide for the ITCS coolant application have been underway and are now in the final stages. An extensive evaluation of biocides was conducted to down-select to several candidates for test trials and was reported on previously. Criteria for that down-select included: the need for safe, non-intrusive implementation and operation in a functioning system; the ability to control existing planktonic and biofilm residing microorganisms; a negligible impact on system-wetted materials of construction; and a negligible reactivity with existing coolant additives. Candidate testing to provide data for the selection of an optimal alternate biocide is now in the final stages. That testing has included rapid biocide effectiveness screening using Biolog MT2 plates to determine minimum inhibitory concentration (amount that will inhibit visible growth of microorganisms), time kill studies to determine the exposure time required to completely eliminate organism growth, materials compatibility exposure evaluations, coolant compatibility studies, and bench-top simulated coolant testing. This paper reports the current status of the effort to select an alternate biocide for the ISS ITCS coolant. The results of various test results to select the optimal candidate are presented.

  11. Water coolant supply in relation to different ultrasonic scaler systems, tips and coolant settings

    NARCIS (Netherlands)

    Koster, T.J.G.; Timmerman, M.F.; Feilzer, A.J.; van der Velden, U.; van der Weijden, F.A.

    2009-01-01

    Objective: This study evaluated "in vitro" the consistency of the water coolant supply for five ultrasonic scaler systems in relation to the tip type and different coolant settings. Material and Methods: The systems were: EMS PM-400, EMS PM-600, Satelec P-max, Dürr Vector and Dentsply Cavitron. For

  12. Computation of Space Shuttle high-pressure cryogenic turbopump ball bearing two-phase coolant flow

    Science.gov (United States)

    Chen, Yen-Sen

    1990-01-01

    A homogeneous two-phase fluid flow model, implemented in a three-dimensional Navier-Stokes solver using computational fluid dynamics methodology is described. The application of the model to the analysis of the pump-end bearing coolant flow of the high-pressure oxygen turbopump of the Space Shuttle main engine is studied. Results indicate large boiling zones and hot spots near the ball/race contact points. The extent of the phase change of the liquid oxygen coolant flow due to the frictional and viscous heat fluxes near the contact areas has been investigated for the given inlet conditions of the coolant.

  13. Performance of Helical Coil Heat Recovery Exchanger using Nanofluid as Coolant

    Directory of Open Access Journals (Sweden)

    Navid Bozorgan

    2015-07-01

    Full Text Available Nanofluids are expected to be a promising coolant condidate in chemical processes for heat transfer system size reduction. This paper focuses on reducing the number of turns in a helical coil heat recovery exchanger with a given heat exchange capacity in a biomass heating plant using γ-Al2O3/n-decane nanofluid as coolant. The nanofluid flows through the tubes and the hot n-hexane flows through the shell. The numerical results show that using nanofluid as coolant in a helical coil heat exchanger can reduce the manufacturing cost of the heat exchanger and pumping power by reducing the number of turns of the coil.

  14. Bicycle accidents.

    Science.gov (United States)

    Lind, M G; Wollin, S

    1986-01-01

    Information concerning 520 bicycle accidents and their victims was obtained from medical records and the victims' replies to questionnaires. The analyzed aspects included risk of injury, completeness of accident registrations by police and in hospitals, types of injuries and influence of the cyclists' age and sex, alcohol, fatigue, hunger, haste, physical disability, purpose of cycling, wearing of protective helmet and other clothing, type and quality of road surface, site of accident (road junctions, separate cycle paths, etc.) and turning manoeuvres.

  15. Multiple Boundary Layer Stripping Model by Plateau-Rayleigh Instability for Fuel-Coolant Interactions

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Woo Hyun; Moriyama, Kiyofumi; Park, Hyun Sun [POSTECH, Pohang (Korea, Republic of)

    2016-05-15

    One of them is FCI (Fuel-Coolant Interaction) phenomenon which is resulted from RPV (Reactor Pressure Vessel) failure at high pressure and high temperature condition of molten fuel. If RPV fails, the melt is ejected to the cavity which is flooded by water as a jet form. Then, the ejected melt jet interacts with water causing massive steam generation and resulting in particulate debris bed on the basemat. As a result of FCI, the initial boundary conditions of steam explosion and debris bed coolability are determined and that is the reason why understanding exact mechanism of melt jet breakup is important in this field. That is, FCI can be said as a starting phenomenon in the ex-vessel severe accident scenario. Until now, numerous previous researchers conducted FCI experiments and numerical analysis in small scale and plant scale. In two MATE experiments, the jet breakup lengths are compared and analyzed with the visualization data. From the observation, the new jet breakup model is proposed including the multiple boundary layer stripping mechanism. Combining the existing and new models, the erosion rate fraction for total melt mass rate was obtained. The new model showed that multiple BLS mechanisms contribute approximately 30% of the total melt jet breakup resulting in the short jet breakup length observed in the MATE 00-2 experiment.

  16. HTGR accident initiation and progression analysis status report. Volume V. AIPA fission product source terms

    Energy Technology Data Exchange (ETDEWEB)

    Alberstein, D.; Apperson, C.E. Jr.; Hanson, D.L.; Myers, B.F.; Pfeiffer, W.W.

    1976-02-01

    The primary objective of the Accident Initiation and Progression Analysis (AIPA) Program is to provide guidance for high-temperature gas-cooled reactor (HTGR) safety research and development. Among the parameters considered in estimating the uncertainties in site boundary doses are uncertainties in fission product source terms generated under normal operating conditions, i.e., fuel body inventories, circulating coolant activity, total plateout activity in the primary circuit, and plateout distributions. The volume presented documents the analyses of these source term uncertainties. The results are used for the detailed consequence evaluations, and they provide the basis for evaluation of fission products important for HTGR maintenance and shielding.

  17. 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)

  18. Who by accident? The social morphology of car accidents.

    Science.gov (United States)

    Factor, Roni; Yair, Gad; Mahalel, David

    2010-09-01

    Prior studies in the sociology of accidents have shown that different social groups have different rates of accident involvement. This study extends those studies by implementing Bourdieu's relational perspective of social space to systematically explore the homology between drivers' social characteristics and their involvement in specific types of motor vehicle accident. Using a large database that merges official Israeli road-accident records with socioeconomic data from two censuses, this research maps the social order of road accidents through multiple correspondence analysis. Extending prior studies, the results show that different social groups indeed tend to be involved in motor vehicle accidents of different types and severity. For example, we find that drivers from low socioeconomic backgrounds are overinvolved in severe accidents with fatal outcomes. The new findings reported here shed light on the social regularity of road accidents and expose new facets in the social organization of death. © 2010 Society for Risk Analysis.

  19. Head injury resulting from scooter accidents in Rome: Differences before and after implementing a universal helmet law

    NARCIS (Netherlands)

    G. La Torre (Giuseppe); E.F. van Beeck (Ed); G. Bertazzoni (Giuliano); W. Ricciardi

    2007-01-01

    textabstractObjectives: To estimate the incidence rates and related determinants of head injuries before and after the implementation of a new universal helmet law in Italy. Methods: The investigation took place in the Emergency Room of the Accident and Emergency Department, Teaching Hospital 'Umber

  20. Analysis of Shahid Rajaee hospital administrative data on injuries resulting from car accidents in Shiraz, Iran: 2011–2014 data

    Directory of Open Access Journals (Sweden)

    Mahnaz Yadollahi

    2017-02-01

    Conclusion: Hospital administrative databases of this hospital could be used as reliable sources of information in providing epidemiologic reports of car accidents in terms of severity and outcomes. Improving the quality of recordings at hospital databases is an important initial step towards more comprehensive injury surveillance in Fars, Iran.

  1. Expert system for online surveillance of nuclear reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.

    1992-12-31

    This report describes an expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  2. Sodium hypochlorite accident resulting in life-threatening airway obstruction during root canal treatment: a case report

    Directory of Open Access Journals (Sweden)

    Al-Sebaei MO

    2015-03-01

    Full Text Available Maisa O Al-Sebaei,1 Omar A Halabi,2 Ibrahim E El-Hakim3 1Department of Oral and Maxillofacial Surgery, King Abdulaziz University – Faculty of Dentistry, Jeddah, Kingdom of Saudi Arabia; 2Saudi Board of Oral and Maxillofacial Surgery, Al-Noor Specialist Hospital, Makkah, Kingdom of Saudi Arabia; 3Department of Oral and Maxillofacial Surgery, Riyadh Colleges of Dentistry and Pharmacy, Riyadh, Kingdom of Saudi Arabia Aim: This case report describes a serious and life-threatening complication of the use of sodium hypochlorite as an irrigation solution in root canal therapy. Summary: This case report describes a hypochlorite accident that occurred in a healthy 42-year-old female who was undergoing routine root canal therapy for the lower right central incisor (tooth #41. After approximately 1 hour of irrigation with 3% sodium hypochlorite (for a total of 12 cc, the patient complained of severe pain and burning in the lip. The swelling progressed over the next 8 hours to involve the sublingual and submental fascial spaces with elevation of the tongue and resultant upper airway obstruction. The patient was intubated and remained on mechanical ventilation for 3 days. She recovered without any skin necrosis or nerve deficits. Key learning points: This case report highlights the importance of carefully performing root canal irrigation with sodium hypochlorite to avoid complications. Careful injection without pressure, the use of proper rubber dam isolation, and the use of the endodontic needle are necessary to avoid this type of complication. Although it is a safe root canal irrigation solution, its use may lead to life-threatening complications. Early recognition and management of the untoward effects of sodium hypochlorite are vital for the patient's safety. Keywords: complications of root canal, facial edema, root canal irrigation, root canal therapy, sodium hypochlorite, upper airway obstruction

  3. Accidents - personal factors

    Energy Technology Data Exchange (ETDEWEB)

    Zaitsev, S.L.; Tsygankov, A.V.

    1982-03-01

    This paper evaluates influence of selected personal factors on accident rate in underground coal mines in the USSR. Investigations show that so-called organizational factors cause from 80 to 85% of all accidents. About 70% of the organizational factors is associated with social, personal and economic features of personnel. Selected results of the investigations carried out in Donbass mines are discussed. Causes of miner dissatisfaction are reviewed: 14% is caused by unsatisfactory working conditions, 21% by repeated machine failures, 16% by forced labor during days off, 14% by unsatisfactory material supply, 16% by hard physical labor, 19% by other reasons. About 25% of miners injured during work accidents are characterized as highly professionally qualified with automatic reactions, and about 41% by medium qualifications. About 60% of accidents is caused by miners with less than a 3 year period of service. About 15% of accidents occurs during the first month after a miner has returned from a leave. More than 30% of accidents occurs on the first work day after a day or days off. Distribution of accidents is also presented: 19% of accidents occurs during the first 2 hours of a shift, 36% from the second to the fourth hour, and 45% occurs after the fourth hour and before the shift ends.

  4. ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

    Directory of Open Access Journals (Sweden)

    Veijo KAUPPINEN

    2012-11-01

    Full Text Available Strained environment is a global problem. In metal industries the use of coolant has become more problematic in terms of both employee health and environmental pollution. It is said that the use of coolant forms approximately 8 - 16 % of the total production costs.The traditional methods that use coolants are now obviously becoming obsolete. Hence, it is clear that using a dry cutting system has great implications for resource preservation and waste reduction. For this purpose, a new cooling system is designed for dry cutting. This paper presents the new eco-friendly cooling innovation and the benefits gained by using this method. The new cooling system relies on a unit for ionising ejected air. In order to compare the performance of using this system, cutting experiments were carried out. A series of tests were performed on a horizontal turning machine and on a horizontal machining centre.

  5. Corrosion problems with aqueous coolants, final report

    Energy Technology Data Exchange (ETDEWEB)

    Diegle, R B; Beavers, J A; Clifford, J E

    1980-04-11

    The results of a one year program to characterize corrosion of solar collector alloys in aqueous heat-transfer media are summarized. The program involved a literature review and a laboratory investigation of corrosion in uninhibited solutions. It consisted of three separate tasks, as follows: review of the state-of-the-art of solar collector corrosion processes; study of corrosion in multimetallic systems; and determination of interaction between different waters and chemical antifreeze additives. Task 1 involved a comprehensive review of published literature concerning corrosion under solar collector operating conditions. The reivew also incorporated data from related technologies, specifically, from research performed on automotive cooling systems, cooling towers, and heat exchangers. Task 2 consisted of determining the corrosion behavior of candidate alloys of construction for solar collectors in different types of aqueous coolants containing various concentrations of corrosive ionic species. Task 3 involved measuring the degradation rates of glycol-based heat-transfer media, and also evaluating the effects of degradation on the corrosion behavior of metallic collector materials.

  6. Study on diesel cylinder-head cooling using nanofluid coolant with jet impingement

    Directory of Open Access Journals (Sweden)

    Su Zhong-Gen

    2015-01-01

    Full Text Available To improve the heat-transfer performance of a diesel-engine cylinder head, nanofluid coolant as a new fluid was investigated, and jet impingement technology was then used to study on how to better improve heat-transfer coefficient at the nose bridge area in the diesel-engine cylinder head. Computational fluid dynamic simulation and experiments results demonstrated that using the same jet impingement parameters, the different volume shares of nanofluids showed better cooling effect than traditional coolant, but the good effect of the new cooling method was unsuitable for high volume share of nanofluid. At the same volume share of nanofluid, different jet impingement parameters such as jet angles showed different heat-transfer performance. This result implies that a strong association exists between jet impingement parameters and heat-transfer coefficient. The increase in coolant viscosity of the nanofluid coolant using jet impingement requires the expense of more drive-power cost.

  7. Technology of high-temperature organic coolant

    Energy Technology Data Exchange (ETDEWEB)

    Vorobei, M.P.; Makin, R.S.; Kuprienko, V.A. [and others

    1993-12-31

    A wide range of studies were carried out in RIAR on the problems connected with the use of high-temperature organic coolant at nuclear power plants. The work performed and successful experience gained in persistent operation of the ARBUS reactor confirmed the inherent safety characteristics, high operational reliability, as well as improved safety of stations with similar reactors. A large scope of studies were carried out at the ARBUS pilot reactor and loop with the organic coolant of the MIR reactor and a wide range of problems were solved. The studies are described.

  8. On-Line Coolant Chemistry Analysis

    Energy Technology Data Exchange (ETDEWEB)

    LM Bachman

    2006-07-19

    Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level.

  9. Radionuclide Release after Channel Flow Blockage Accident in CANDU-6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hoon; Jun, Hwang Yong [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The channel flow blockage accident is one of the in core loss of coolant accidents, the release path of radionuclide is very different from conventional loss of coolant accidents. The large amount of radionuclide released from broken channel is being washed during it passes through the moderator in Calandria. The objective of containment behavior analysis for channel flow blockage event is to assess the amount of radionuclide release to the ambient atmosphere. Radionuclide release rates in case of channel flow blockage with all safety system available, that is containment building is intact, as well as with containment system impairment are analyzed with GOTHIC and SMART code

  10. Crack stability analysis of low alloy steel primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

  11. Heat up and potential failure of BWR upper internals during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    In boiling water reactors, the steam dome, steam separators, and dryers above the core are comprised of approximately 100 tons of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. Historically, the upper internals have been modeled using severe accident codes with relatively simple approximations. The upper internals are typically modeled in MELCOR as two lumped volumes with simplified heat transfer characteristics, with no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. This modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. The results indicate that the upper internals can reach high temperatures during a severe accident; they are predicted to reach a high enough temperature such that they lose their structural integrity and relocate. The additional 100 tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.

  12. Use of a PKZh-902 instrument for monitoring solid phases in an organic coolant

    Energy Technology Data Exchange (ETDEWEB)

    Gavrillin, A.I.; Gagarin, S.I.; Sokolov, V.E.; Zabelin, A.I.

    1986-11-01

    Preliminary tests have been performed with a PKZh-902 in checking feed and circulating coolant, and also in evaluating the performance of the cleaning devices. The test program involved determining the stability of the esnsor materials in ditolymethane, examining the effects of radioactive products and those of the optical characteristics of impurities in the first-loop coolant on the readings and errors of measurement, and checking the scope for using the instrument for continuous monitoring of particle concentrations. Results confirm that the PKZh-902 enables one to monitor the composition and concentration of the solid dispersed phase reliably and with adequate accuracy in the feed coolant. The use for monitoring the loop coolant requires additonal research.

  13. Steam as coolant and lubricant in turning of metal matrix composites

    Institute of Scientific and Technical Information of China (English)

    Raviraj SHETTY; Raghuvir PAI; Vasanth KAMATH; Shrikanth S.RAO

    2008-01-01

    Green cutting has become focus of attention in ecological and environmental protection.Steam is cheap.pollution-free and eco-friendly,and then is a good and economical coolant and lubricant.Steam generator and steam feeding system were developed to generate and feed steam.Comparative experiments were carried out in cutting AA6061-15 v0l.%SiC(25 μm particle size),with cubic boron nitride(CBN)insert KB-90 grade under the conditions of compressed air,oil water emulsion,steam as coolant and lubricant,and dry cutting,respectively.The experimental results show that,with steam as coolant and lubricant,gradual reduction in the cutting force,friction coefficient,surface roughness and cutting temperature values were observed.Further,there was reduction in built up edge formation.1t is proved that use of water steam as coolant and lubricant is environmentally friendly.

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

    Directory of Open Access Journals (Sweden)

    Gilberto Espinosa-Paredes

    2012-01-01

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

  15. The Thirty Years’ Results of Radiation Hygienic Monitoring of Tula Region territories contaminated due to the Chernobyl NPP accident

    Directory of Open Access Journals (Sweden)

    V. V. Boldyreva

    2016-01-01

    Full Text Available Over 50% of Tula Region areas were contaminated after the Chernobyl NPP accident. The article provides the thirty years’ results of radiation hygienic monitoring of the Chernobyl accidental fallout - affected areas in Tula Region. The radiation situation is assessed at the initial accidental period and at the current stage. The initial levels of gamma - radiation dose intensity (up to 35 mcSv/hr are identified for the period of the “iodine” hazard along with the tabular data on the dose intensity relative stabilization by the beginning of August 1986 due to iodine-131 decay. The information is presented regarding iodine-131 tentative maximum permissible level exceedance in the dairy products of the two most contaminated regional areas - Plavskoye and Arsenievskoye. The article also provides the laboratory data on the total beta - activity in the foodstuffs in 1986-1987 and cesium-137 maximum permissible level exceedence in 1986. The radionuclide maximum permissible level exceedances in foodstuffs were registered only in 1986 due to the plants surface contamination whereas in the forest mushrooms those exceedances were repeatedly found until 2004. The black earths and grey forest soils had a benign impact upon the intensity of the radionuclide transfer into plants which resulted in the formation of internal radiation doses.At the current stage, the content of cesium-137 and strontium- 90 in the foodstuffs can only be quantified by a radiochemical method. The table covers all the districts within the boundaries of radiation contaminated zones. The radiochemical studies indicate the main dose - forming products. The article contains the table of internal and external radiation doses of the population in Plavsk town over 1986-1990 and displays the factors impacting population’s internal and external exposure. The Chernobyl - affected exposure dose of the population is mostly attributed to the external radiation and, for over twenty years, it

  16. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  17. Safety assessment for electricity generation failure accident of gas cooled nuclear power plant using system dynamics (SD) method

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Tae Ho [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering

    2013-04-15

    The power production failure happens in the loss of coolant of the nuclear power plants (NPPs). The air ingress is a serious accident in gas cooled NPPs. The quantification of the study performed by the system dynamics (SD) method which is processed by the feedback algorithms. The Vensim software package is used for the simulation, which is performed by the Monte-Carlo method. Two kinds of considerations as the economic and safety properties are important in NPPs. The result shows the stability of the operation when the power can be decided. The maximum value of risk is the 11.77 in 43rd and the minimum value is 0.0 in several years. So, the success of the circulation of coolant is simulated by the dynamical values. (orig.)

  18. Analysis of Shahid Rajaee hospital administrative data on injuries resulting from car accidents in Shiraz, Iran: 2011-2014 data.

    Science.gov (United States)

    Yadollahi, Mahnaz; Ghiassee, Aida; Anvar, Mehrdad; Ghaem, Hale; Farahmand, Mohammad

    2017-02-01

    The administrative data from trauma centers could serve as potential sources of invaluable information while studying epidemiologic features of car accidents. In this cross-sectional analysis of Shahid Rajaee hospital administrative data, we aimed to evaluate patients injured in car accidents in terms of age, gender, injury severity, injured body regions and hospitalization outcome in the recent four years (2011-2014). The hospital registry was accessed at Shiraz Trauma Research Center (Shiraz, Iran) and the admission's unit data were merged with the information gathered upon discharge. A total number of 27,222 car accident patients aged over 15 years with International Classification of Diseases 10th revision (ICD-10) external causes of injury codes (V40.9-V49.9) were analyzed. Injury severity score and injured body regions were determined based on converting ICD-10 injury codes to Abbreviated Injury Scale (AIS-98) severity codes using a domestically developed electronic algorithm. A binary logistic regression model was applied to the data to examine the contribution of all independent variables to in-hospital mortality. Men accounted for 68.9% of the injuries and the male to female ratio was 2.2:1. The age of the studied population was (34 ± 15) years, with more than 77.2% of the population located in the 15-45 years old age group. Head and neck was the most commonly injured body region (39.0%) followed by extremities (27.2%). Injury severity score (ISS) was calculated for 13,152 (48.3%) patients, of whom, 80.9% had severity scores less than 9. There were 332 patients (1.2%) admitted to the intensive care units and 422 in-hospital fatalities (1.5%) were recorded during the study period. Age above 65 years [OR = 7.4, 95% CI (5.0-10.9)], ISS above 16 [OR = 9.1, 95% CI (5.5-14.9)], sustaining a thoracic injury [OR = 7.4, 95% CI (4.6-11.9)] and head injury [OR = 4.9, 95% CI (3.1-7.6)] were the most important independent predictors of death following car

  19. NGNP Reactor Coolant Chemistry Control Study

    Energy Technology Data Exchange (ETDEWEB)

    Brian Castle

    2010-11-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor.

  20. An analytical investigation of cold leg small break accidents of the ATLAS facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon-Sik, E-mail: yskim3@kaeri.re.kr; Cho, Seok; Choi, Ki-Yong

    2015-09-15

    A previous parametric study of the direct vessel injection (DVI) line breaks was re-evaluated to see its applicability to that of the cold leg (CL) pipe breaks in advanced thermal-hydraulic test loop for accident simulation (ATLAS). Evaluation results of the tests and analyses for the major parameters, e.g., the pressurizer (PZR) pressure, downcomer water level, collapsed core water level, and clad temperature, were compared for four different CL pipe break scenarios. The overall trends of the major parameters showed reasonable behaviors between the tests and analyses. The clad temperature showed conservative behaviors in the analyses using the suggested options. The suggested counter-current flow limit (CCFL) options for the fuel alignment plate (FAP) and cross-over legs (COLs) can be applicable to any small-break loss-of-coolant accident (SBLOCA) scenario for the CL pipe and DVI line breaks in the ATLAS tests.

  1. Proceedings of the workshop on severe accident research held in Japan (SARJ-97)

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Jun [ed.

    1998-05-01

    The Workshop on Severe Accident Research held in Japan (SARJ-97) was taken place at Pacifico Yokohama on October 6 - 8, 1997, and attended by 180 participants from 15 countries and one international organizations. The 59 papers, which cover wide areas of severe accident research both in experiments and analysis, such as in-vessel melt retention, fuel-coolant interaction, fission products behavior, structural integrity, containment behavior, computer simulations, and accident management, are indexed individually. (J.P.N.)

  2. Proceedings of the workshop on severe accident research held in Japan (SARJ-98)

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Jun [ed.

    1999-07-01

    The Workshop on Severe Accident Research held in Japan (SARJ-98) was taken place at Hotel Lungwood on November 4-6, 1998, and attended by 181 participants from 13 countries. The 63 papers, which cover wide areas of severe accident research both in experiments and analyses, such as in-vessel melt retention, fuel-coolant interaction, fission products behavior, structural integrity, containment behavior, computer simulations, and accident management, are indexed individually. (J.P.N.)

  3. 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...

  4. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project -2006 Update

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2006-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. This paper presents a status of the coolant stability over the past year as well as results from destructive analyses of hardware removed from the on-orbit system and the current approach to coolant remediation.

  5. Ruthenium behaviour in severe nuclear accident conditions. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Backman, U.; Lipponen, M.; Auvinen, A.; Jokiniemi, J.; Zilliacus, R. [VVT Processes (Finland)

    2004-08-01

    During routine nuclear reactor operations, ruthenium will accumulate in the fuel in relatively high concentrations. In a steam atmosphere, ruthenium is not volatile, and it is not likely to be released from the fuel. However, in an air ingress accident during reactor power operation or during maintenance, ruthenium may form volatile species, which may be released into the containment. Oxide forms of ruthenium are more volatile than the metallic form. Radiotoxicity of ruthenium is high both in the short and the long term. The results of this project imply that in oxidising conditions during nuclear reactor core degradation, ruthenium release increases as oxidised gaseous species Ru03 and Ru04 are formed. A significant part of the released ruthenium is then deposited on reactor coolant system piping. However, in the presence of steam and aerosol particles, a substantial amount of ruthenium may be released as gaseous Ru04 into the containment atmosphere. (au)

  6. 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.

  7. Behavior of irradiated BWR fuel under reactivity-initiated-accident conditions; Results of tests FK-1, -2 and -3

    OpenAIRE

    2004-01-01

    Boiling water reactor (BWR) fuels with burnups of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated-accident (RIA) conditions. BWR fuel segment rods of 8times8BJ (STEP I) type from Fukushima-Daiichi Unit 3 nuclear power plant were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g) within about 20 ms. The fuel cladding...

  8. Submerged demineralize system processing of TMI-2 accident waste water

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, H.F.; Quinn, G.J.

    1983-02-01

    Accident-generated radioactive waste at Three Mile Island Unit 2 includes a varity of high and low specific-activity waste. The high-specific-activity waste, particularly over one million gallons of contaminated water, required special processing and secondary waste handling. General public utilities and its contractors developed a zeolite-based ion-exchange system called the Submerged Demineralizer System to reduce contamination levels in the water to below allowable limits. Testing and modifications resulted in an operating system that had successfully processed waste water from the Reactor Coolant Bleed Tanks, the Reactor Building Basement, and the Reactor Coolant System as of August 1982. System design objectives were met and decontamination criteria established in 10 CFR 20 were attained. Additional wastes that could not be handled routinely were generated by another water-processing system, called EPICOR II. EPICOR II wastes are discussed. Low-specific-activity (LSA) wastes such as trash and resin-bed waste canisters are also included in handling. LSA wastes are routinely handled and shipped according to existing industry practice. Plant records are summarized to provide approximate yearly volumes and curie loadings of low-specific-activity wastes being shipped off the Island to a commercial burial site.

  9. Occupational accidents aboard merchant ships

    DEFF Research Database (Denmark)

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

    2002-01-01

    Objectives: To investigate the frequency, circumstances, and causes of occupational accidents aboard merchant ships in international trade, and to identify risk factors for the occurrence of occupational accidents as well as dangerous working situations where possible preventive measures may...... be initiated. Methods: The study is a historical follow up on occupational accidents among crew aboard Danish merchant ships in the period 1993–7. Data were extracted from the Danish Maritime Authority and insurance data. Exact data on time at risk were available. Results: A total of 1993 accidents were...... rate of accidents than Danish citizens. Age was a major risk factor for accidents causing permanent disability. Change of ship and the first period aboard a particular ship were identified as risk factors. Walking from one place to another aboard the ship caused serious accidents. The most serious...

  10. Structure of comorbid diseases in patients with non-alcoholic fatty liver disease, exposed to ionizing radiation as a result of the Chornobyl NPP accident.

    Science.gov (United States)

    Gasanova, O V; Sarkisova, E O; Ovsyannikova, L M; Chumak, A A; Nosach, O V; Nezgovorova, G A; Gromadska, V M

    2014-09-01

    Objective - to study of the structure of comorbid diseases in patients with nonalcoholic fatty liver disease (NAFLD), exposed to ionizing radiation due to the Chornobyl accident, and in the groups of comparison. Materials and methods. Array of surveyed males was divided into 4 groups: the main group - 136 patients with NAFLD affected by the Chornobyl NPP accident, the first comparison group (І CG) - 28 affected by Chornobyl NPP accident without liver disease, the second comparison group (II CG) - 50 patients with NAFLD not exposed to factors of the Chornobyl disaster, and the third comparison group (III CG) - 16 unexposed persons without liver disease. Results. A significant amount of co-morbid pathology was found in all studied groups but III CG: at the average from 4.1 ± 0.4 diseases in II CG to 5.2 ± 0.2 in the main group of patients (p diseases were established. The most common disorders were cardiovascular and cerebrovascular and endocrine diseases. Structure of comorbid pathology in NAFLD in the main study group differed from this in II CG by a significantly greater frequency of detection of cerebrovascular diseases (73.5 and 56 %, p diseases (47.1 %) were more frequent than in II GC: correspondingly 58 % (p diseases and nosological structure. The most common disorders were cardiovascular, cerebrovascular, and endocrine diseases. Significantly higher incidence of essential hypertension, vascular encephalopathy, and thyroid disease was determined.

  11. Fuel-coolant interaction (FCI) phenomena in reactor safety. Current understanding and future research needs

    Energy Technology Data Exchange (ETDEWEB)

    Speis, T.P. [Maryland Univ., College Park, MD (United States); Basu, S.

    1998-01-01

    This paper gives an account of the current understanding of fuel-coolant interaction (FCI) phenomena in the context of reactor safety. With increased emphasis on accident management and with emerging in-vessel core melt retention strategies for advanced light water reactor (ALWR) designs, recent interest in FCI has broadened to include an evaluation of potential threats to the integrity of reactor vessel lower head and ex-vessel structural support, as well as the role of FCI in debris quenching and coolability. The current understanding of FCI with regard to these issues is discussed, and future research needs to address the issues from a risk perspective are identified. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  13. Investigation for Ensuring the Reliability of the MELCOR Analysis Results

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Joonyoung; Maeng, Yunhwan; Lee, Jaeyoung [Handong Global Univ., Pohang (Korea, Republic of)

    2015-10-15

    Flow rate could be also main factor to be proven because it is in charge of a role which takes thermal balance through heat transfer in inner side of fuel assembly. Some problems about a reliability of MELCOR results could be posed in the 2{sup nd} technical report of NSRC project. In order to confirm whether MELCOR results are dependable, experimental data of Sandia Fuel Project 1 phase were used to be compared to be a reference. In Spent Fuel Pool (SFP) severe accident, especially in case of boil-off, partial loss of coolant accident, and complete loss of coolant accident; heat source and flow rate could be main points to analyze the MELCOR results. Heat source might be composed as decay heat and oxidation heat. Because heat source makes it possible to lead a zirconium fire situation if it is satisfied that heat accumulates in spent fuel rod and then cladding temperature could be raised continuously to be generated an oxidation heat, this might be a main factor to be confirmed. This work was proposed to investigate reliability of MELCOR results in order to confirm physical phenomena if SFP severe accident is occurred. Almost results showed that MELCOR results were significantly different by minute change of main parameter in identical condition. Therefore it could be necessary that oxidation coefficients have to be chosen as value to delineate real phenomena as possible.

  14. Loss of vacuum accident (LOVA): Comparison of computational fluid dynamics (CFD) flow velocities against experimental data for the model validation

    Energy Technology Data Exchange (ETDEWEB)

    Bellecci, C.; Gaudio, P.; Lupelli, I. [Faculty of Engineering, University of Rome ' Tor Vergata' , Via del Politecnico 1, 00133 Rome (Italy); Malizia, A., E-mail: malizia@ing.uniroma2.it [Faculty of Engineering, University of Rome ' Tor Vergata' , Via del Politecnico 1, 00133 Rome (Italy); Porfiri, M.T. [ENEA Nuclear Fusion Technologies, Via Enrico Fermi 45 I, 00044, Frascati (Italy); Quaranta, R.; Richetta, M. [Faculty of Engineering, University of Rome ' Tor Vergata' , Via del Politecnico 1, 00133 Rome (Italy)

    2011-06-15

    A recognized safety issue for future fusion reactors fueled with deuterium and tritium is the generation of sizeable quantities of dust. Several mechanisms resulting from material response to plasma bombardment in normal and off-normal conditions are responsible for generating dust of micron and sub-micron length scales inside the VV (Vacuum Vessel) of experimental fusion facilities. The loss of coolant accidents (LOCA), loss of coolant flow accidents (LOFA) and loss of vacuum accidents (LOVA) are types of accidents, expected in experimental fusion reactors like ITER, that may jeopardize components and plasma vessel integrity and cause dust mobilization risky for workers and public. The air velocity is the driven parameter for dust resuspension and its characterization, in the very first phase of the accidents, is critical for the dust release. To study the air velocity trend a small facility, Small Tank for Aerosol Removal and Dust (STARDUST), was set up at the University of Rome 'Tor Vergata', in collaboration with ENEA Frascati laboratories. It simulates a low pressurization rate (300 Pa/s) LOVA event in ITER due to a small air inlet from two different positions of the leak: at the equatorial port level and at the divertor port level. The velocity magnitude in STARDUST was investigated in order to map the velocity field by means of a punctual capacitive transducer placed inside STARDUST without obstacles. FLUENT was used to simulate the flow behavior for the same LOVA scenarios used during the experimental tests. The results of these simulations were compared against the experimental data for CFD code validation. For validation purposes, the CFD simulation data were extracted at the same locations as the experimental data were collected for the first four seconds, because at the beginning of the experiments the maximum velocity values (that could cause the almost complete dust mobilization) have been measured. In this paper the authors present and

  15. Recovery studies for plutonium machining oil coolant

    Energy Technology Data Exchange (ETDEWEB)

    Navratil, J. D.; Baldwin, C. E.

    1977-04-27

    Lathe coolant oil, contaminated with plutonium and having a carbon tetrachloride diluent, is generated in plutonium machining areas at Rocky Flats. A research program was initiated to determine the nature of plutonium in this mixture of oil and carbon tetrachloride. Appropriate methods then could be developed to remove the plutonium and to recycle the oil and carbon tetrachloride. Studies showed that the mixtures of spent oil and carbon tetrachloride contained particulate plutonium and plutonium species that are soluble in water or in oil and carbon tetrachloride. The particulate plutonium was removed by filtration; the nonfilterable plutonium was removed by adsorption on various materials. Laboratory-scale tests indicated the lathe-coolant oil mixture could be separated by distilling the carbon tetrachloride to yield recyclable products.

  16. Use of Distribution Devices for Hydraulic Profiling of Coolant Flow in Core Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2014-01-01

    Full Text Available In setting up a reactor plant for the transportation-power module of the megawatt class an important task is to optimize the path of flow, i.e. providing moderate hydraulic resistance, uniform distribution of the coolant. Significant contribution to the hydraulic losses makes one selected design of the coolant supplies. It is, in particular, hemispherical or semi-elliptical shape of the supply reservoir, which is selected to reduce its mass, resulting in the formation of torusshaped vortex in the inlet manifold, that leads to uneven coolant velocity at the inlet into the core, the flow pulsations, hydraulic losses.To control the flow redistribution in the core according to the level of energy are used the switchgear - deflectors installed in a hemispherical reservoir supplying coolant to the fuel elements (FE of the core of gas-cooled reactor. This design solution has an effect on the structure of the flow, rate in the cooling duct, and the flow resistance of the collector.In this paper we present the results of experiments carried out on the gas dynamic model of coolant paths, deflectors, and core, comprising 55 fuel rod simulators. Numerical simulation of flow in two-parameter model, using the k-ε turbulence model, and the software package ANSYS CFX v14.0 is performed. The paper demonstrates that experimental results are in compliance with calculated ones.The results obtained suggest that the use of switchgear ensures a coolant flow balance directly at the core inlet, thereby providing temperature reduction of fuel rods with a uniform power release in the cross-section. Considered options to find constructive solutions for deflectors give an idea to solve the problem of reducing hydraulic losses in the coolant paths, to decrease pulsation components of flow in the core and length of initial section of flow stabilization.

  17. Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance

    Directory of Open Access Journals (Sweden)

    Cheng Bo

    2016-01-01

    Full Text Available Molybdenum based alloy is selected as a candidate to enhance tolerance of fuel to severe loss of coolant accidents due to its high melting temperature of ∼2600 °C and ability to maintain sufficient mechanical strength at temperatures exceeding 1200 °C. An outer layer of either a Zr-alloy or Al-containing stainless steel is designed to provide corrosion resistance under normal operation and oxidation resistance in steam exceeding 1000 °C for 24 hours under severe loss of coolant accidents. Due to its higher neutron absorption cross-sections, the Mo-alloy cladding is designed to be less than half the thickness of the current Zr-alloy cladding. A feasibility study has been undertaken to demonstrate (1 fabricability of long, thin wall Mo-alloy tubes, (2 formability of a protective outer coating, (3 weldability of Mo tube to endcaps, (4 corrosion resistance in autoclaves with simulated LWR coolant, (5 oxidation resistance to steam at 1000–1500 °C, and (6 sufficient axial and diametral strength and ductility. High purity Mo as well as Mo + La2O3 ODS alloy have been successfully fabricated into ∼2-meter long tubes for the feasibility study. Preliminary results are encouraging, and hence rodlets with Mo-alloy cladding containing fuel pellets have been under preparation for irradiation at the Advanced Test Reactor (ATR in Idaho National Laboratory. Additional efforts are underway to enhance the Mo cladding mechanical properties via process optimization. Oxidation tests to temperatures up to 1500 °C, and burst and creep tests up to 1000 °C are also underway. In addition, some Mo disks in close contact with UO2 from a previous irradiation program (to >100 GWd/MTU at the Halden Reactor have been subjected to post-irradiation examination to evaluate the chemical compatibility of Mo with irradiated UO2 and fission products. This paper will provide an update on results from the feasibility study and discuss the attributes of the

  18. System Study: High-Pressure Coolant Injection 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

  19. System Study: High-Pressure Coolant Injection 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-01-31

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

  20. Whole-body counter survey results 4 months after the Fukushima Daiichi NPP accident in Minamisoma City, Fukushima

    CERN Document Server

    Hayano, Ryugo S; Nomura, Shuhei; Nemoto, Tsuyoshi; Tsubokura, Masaharu; Hanai, Tatsuo; Kumemoto, Yuki; Kowata, Satoshi; Oikawa, Tomoyoshi; Kanazawa, Yukio

    2014-01-01

    Using the first WBC unit installed in Fukushima Prefecture after the accident, the radiocesium body burdens of 566 high-risk residents of Minamisoma city were measured in July 2011 at the Minamisoma Municipal General Hospital. The analysis of the data was challenging because this chair-type WBC unit did not have sufficient shielding against background gamma rays, and methods had to be developed to reliably compensate for the body-attenuated background radiation. Fortunately, data for repeated tests of hospital staff members using both the chair-type and well-shielded FASTSCAN WBC units, installed in September 2011, were available, and could be used to check the validity of the analysis. The CEDs of all subjects, estimated under the assumption of acute inhalation in March 2011, were found to be less than 1 mSv.

  1. EEG patterns in persons exposed to ionizing radiation as a result of the Chernobyl accident: part 1: conventional EEG analysis.

    Science.gov (United States)

    Loganovsky, K N; Yuryev, K L

    2001-01-01

    Prospective conventional EEG study was carried out 3-5 and 10-13 years after the Chernobyl accident (1986) in patients who had acute radiation sickness and in emergency workers in 1986 ("liquidators"). Control groups comprised healthy volunteers; veterans of the Afghanistan war with posttraumatic stress disorder; veterans with mild traumatic brain injury; and patients with dyscirculatory encephalopathy. In 3-5 years after irradiation, there were irritated EEG changes with paroxysmal activity shifted to the left frontotemporal region (cortical-limbic overactivation) that were transformed 10-13 years after irradiation toward a low-voltage EEG pattern with excess of fast (beta) and slow (delta) activity together with depression of alpha and theta activity (organic brain damage with inhibition of the cortical-limbic system). Quantitative EEG is likely to be very informative for investigation of dose-effect relationships.

  2. Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Barrio del Juanes, M.T.; Garcia Cuesta, J.C.; Vallejo Diaz, I.; Herranz Puebla

    2001-07-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

  3. 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); McMurray, Jake W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    Since 2011, Oak Ridge National Laboratory (ORNL) has been conducting high temperature steam oxidation testing of candidate alloys for accident tolerant fuel (ATF) cladding. These concepts are designed to enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the US ATF community, the Severe Accident Test Station (SATS) has been evaluating candidate materials (including coatings) since 2012. Compared to the current UO2/Zr-based alloy fuel system, alternative cladding materials need to offer slower oxidation kinetics and a smaller enthalpy of oxidation in order to significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models. However, prior 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. In some cases, the results have been difficult to interpret and more fundamental information is needed such as the stability of alumina in flowing steam at 1400°-1500°C. This report summarizes recent work to measure the steam oxidation kinetics of candidate alloys, the evaporation rate of alumina in steam and the development of integral data on FeCrAl compared to conventional Zr-based cladding.

  4. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A. [ETSI Industriales-Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal, 2. 28006 Madrid (Spain)

    2010-07-01

    Breeder reactors are considered the unique tool for fully exploiting the natural nuclear resources. In current LWR, only a 0.5% of the primary energy contained in the nuclei removed from the mine is converted into useful heat, with the rest remaining in the depleted uranium or in the spent fuel. The objective of resource-efficiency stimulated the interest in Fast- Reactor-based fuel cycles which can exploit a much higher fraction of the energy content of the mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles would also offers several potential advantages over a uranium fuel cycle. The coolant initially chosen for most of the FBR programs launched in the 60's was sodium, which still is considered the best candidate for these reactors. However, Na-cooled FBR have a positive void reactivity coefficient, which has been among others, a fundamental drawback that has cancelled the deployment of these reactors. Therefore, it seems reasonable to explore totally new options on coolants for breeders. In this paper, a proposal is presented on a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would need an extensive R and D programme, this paper presents the very appealing properties of this salt, in the case of using a specific type of fuel, similar to that of pebble bed reactors. The concept will be studied over a typical MOX composition and extended to a Thorium-based cycle. The general analysis takes into account requirements for criticality (opening the option of hybrid subcritical systems); requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window is found in the definition of a F{sub 2}Be cooled reactor where the safety requirement is met, unlike for molten metal cooled reactors which always have positive void

  5. A preliminary study for the implementation of general accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of each strategy are also investigated. 11 refs., 3 figs., 3 tabs. (Author)

  6. 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.

  7. Comparative Evaluation of Coolant Mixing Experiments at the ROCOM, Vattenfall, and Gidropress Test Facilities

    Directory of Open Access Journals (Sweden)

    S. Kliem

    2007-01-01

    Full Text Available Coolant mixing is an important mitigative mechanism against reactivity accidents caused by local boron dilution. Experiments on coolant mixing were carried out at three different test facilities representing three different reactor types. These are the ROCOM test facility modelling a German KONVOI-type reactor, the Vattenfall test facility being a model of a Westinghouse three-loop PWR, and the Gidropress test facility modelling a VVER-1000 PWR. The scenario of the start-up of the first main coolant pump was investigated in all three facilities. The experiments were accompanied by velocity measurements in the downcomer for the same scenario in the ROCOM and the Vattenfall test facilities. A similar flow structure was found in these measurements in both cases. A maximum of the velocity is measured at the opposite side in regard to the position of the loop with the starting-up pump whilst a recirculation area was found just below this inlet nozzle in both facilities. The analysis of the slug mixing experiments showed also comparable flow behaviour. In accordance with the velocity measurements, the first part of the deboration is also found on the opposite side. In this region, the maximum deboration is measured in all three cases. These maximum values are in the same order of magnitude for nearly identical initial slug volumes.

  8. Research in the Ciemat on severe accidents: strategy and recent results; Investigaciones en el Ciemat sobre accidentes severos: estrategia y resultados recientes

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.

    2012-11-01

    Severe accident research is a fundamental brick in the nuclear technology wall. Its complexity entails huge challenges that require international cooperation to be overcome. CIEMAT has accumulated more than 40 years of experience in the field. By setting a structured research strategy and a continuous enhancement of theoretical an experimental capabilities, CIEMAT has recently produced the results on which this article builds up. Through them, both its working domains and its firm commitment for a continuous growth of knowledge and know-how are outlined. (Author) 24 refs.

  9. Radiological impact of the nuclear power plant accident on freshwater fish in Fukushima: An overview of monitoring results.

    Science.gov (United States)

    Wada, Toshihiro; Tomiya, Atsushi; Enomoto, Masahiro; Sato, Toshiyuki; Morishita, Daigo; Izumi, Shigehiko; Niizeki, Kouji; Suzuki, Shunji; Morita, Takami; Kawata, Gyo

    2016-01-01

    Radionuclide ((131)I, (134)Cs, and (137)Cs) concentrations of monitored freshwater fish species collected from different habitats (rivers, lakes, and culture ponds) in Fukushima Prefecture during March 2011-December 2014 (total 16 species, n = 2692) were analyzed to present a detailed description of radionuclide contamination after the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) accident, and to elucidate species-specific spatiotemporal declining trends of (137)Cs concentration for their respective habitats. Low concentrations of (131)I (≤24 Bq kg(-1)-wet) were detected from only 11 samples collected during March-June 2011, demonstrating that (131)I transferred to freshwater fish were not intense. In river and lake fishes, a more gradual decrease and higher radiocesium ((134)Cs, (137)Cs) concentrations were observed than in culture pond fishes, which strongly implied that radiocesium in freshwater fish species was mainly bioaccumulated through the food web in the wild. During 2011-2014, percentages above the Japanese regulatory limit of 100 Bq kg(-1)-wet for radiocesium in river and lake fish (14.0% and 39.6%, respectively) were higher than in monitored marine fish (9.9%), indicating longer-term contamination of freshwater fish species, especially in lakes. Higher radiocesium concentrations (maximum 18.7 kBq kg(-1)-wet in Oncorhynchus masou) were found in the northwestern areas from the FDNPP with higher deposition. However, radiocesium contamination levels were regarded as 1-2 orders of magnitude less than those after the Chernobyl accident. Lagged increase of (137)Cs concentration and longer ecological half-lives (Teco: 1.2-2.6 y in the central part of Fukushima Prefecture) were observed in carnivorous salmonids (O. masou, Salvelinus leucomaenis), whereas a rapid increase and decrease of (137)Cs concentration and shorter Teco (0.99 and 0.69 y) were found in herbivorous and planktivorous osmerids (Plecoglossus altivelis, Hypomesus nipponensis) with

  10. Assessment of environmental public exposure from a hypothetical nuclear accident for Unit-1 Bushehr nuclear power plant.

    Science.gov (United States)

    Sohrabi, M; Ghasemi, M; Amrollahi, R; Khamooshi, C; Parsouzi, Z

    2013-05-01

    Unit-1 of the Bushehr nuclear power plant (BNPP-1) is a VVER-type reactor with 1,000-MWe power constructed near Bushehr city at the coast of the Persian Gulf, Iran. The reactor has been recently operational to near its full power. The radiological impact of nuclear power plant (NPP) accidents is of public concern, and the assessment of radiological consequences of any hypothetical nuclear accident on public exposure is vital. The hypothetical accident scenario considered in this paper is a design-basis accident, that is, a primary coolant leakage to the secondary circuit. This scenario was selected in order to compare and verify the results obtained in the present paper with those reported in the Final Safety Analysis Report (FSAR 2007) of the BNPP-1 and to develop a well-proven methodology that can be used to study other and more severe hypothetical accident scenarios for this reactor. In the present study, the version 2.01 of the PC COSYMA code was applied. In the early phase of the accidental releases, effective doses (from external and internal exposures) as well as individual and collective doses (due to the late phase of accidental releases) were evaluated. The surrounding area of the BNPP-1 within a radius of 80 km was subdivided into seven concentric rings and 16 sectors, and distribution of population and agricultural products was calculated for this grid. The results show that during the first year following the modeled hypothetical accident, the effective doses do not exceed the limit of 5 mSv, for the considered distances from the BNPP-1. The results obtained in this study are in good agreement with those in the FSAR-2007 report. The agreement obtained is in light of many inherent uncertainties and variables existing in the two modeling procedures applied and proves that the methodology applied here can also be used to model other severe hypothetical accident scenarios of the BNPP-1 such as a small and large break in the reactor coolant system as well

  11. MELCOR accident analysis for ARIES-ACT

    Energy Technology Data Exchange (ETDEWEB)

    Paul W. Humrickhouse; Brad J. Merrill

    2012-08-01

    We model a loss of flow accident (LOFA) in the ARIES-ACT1 tokamak design. ARIES-ACT1 features an advanced SiC blanket with LiPb as coolant and breeder, a helium cooled steel structural ring and tungsten divertors, a thin-walled, helium cooled vacuum vessel, and a room temperature water-cooled shield outside the vacuum vessel. The water heat transfer system is designed to remove heat by natural circulation during a LOFA. The MELCOR model uses time-dependent decay heats for each component determined by 1-D modeling. The MELCOR model shows that, despite periodic boiling of the water coolant, that structures are kept adequately cool by the passive safety system.

  12. A Study on the Operation Strategy for Combined Accident including TLOFW accident

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    It is difficult for operators to recognize the necessity of a feed-and-bleed (F-B) operation when the loss of coolant accident and failure of secondary side occur. An F-B operation directly cools down the reactor coolant system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. The plant is not always necessary the F-B operation when the secondary side is failed. It is not necessary to initiate an F-B operation in the case of a medium or large break because these cases correspond to low RCS pressure sequences when the secondary side is failed. If the break size is too small to sufficiently decrease the RCS pressure, the F-B operation is necessary. Therefore, in the case of a combined accident including a secondary cooling system failure, the provision of clear information will play a critical role in the operators' decision to initiate an F-B operation. This study focuses on the how we establish the operation strategy for combined accident including the failure of secondary side in consideration of plant and operating conditions. Previous studies have usually focused on accidents involving a TLOFW accident. The plant conditions to make the operators confused seriously are usually the combined accident because the ORP only focuses on a single accident and FRP is less familiar with operators. The relationship between CET and PCT under various plant conditions is important to decide the limitation of initiating the F-B operation to prevent core damage.

  13. Effects of staggered blades on the hydraulic characteristics of a 1400-MW canned nuclear coolant pump

    Directory of Open Access Journals (Sweden)

    Fang-Ming Zhou

    2016-08-01

    Full Text Available A canned nuclear coolant pump is used in an advanced third-generation pressurized water reactor. Impeller is a key component of a canned nuclear coolant pump. Usually, the blade is installed between the hub and the shroud as an entire part. The blade is divided into two parts and is staggered in the circumferential direction is an approach of blade design. To understand the effects of staggered blades on a canned nuclear coolant pump, this article numerically investigated different types of staggering. The validity of the numerical simulation was confirmed by comparing the numerical and experimental results. The performance change of a canned nuclear coolant pump with staggered blades was acquired. Hydraulic performance curves, axial force curves, static pressure distributions at the impeller outlet, and static pressure pulsations were performed to investigate the performance changes caused by the staggered blades. The results show that the staggered blade has an important influence on the performance of canned nuclear coolant pumps. A staggered blade does not improve hydraulic performance but does improve the axial force and pressure pulsation. Specifically, the staggered blades can significantly reduce the pressure pulsation amplitude on the impeller pass frequency.

  14. Rotor dynamic analysis of main coolant pump

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chong Won; Seo, Jeong Hwan; Kim, Choong Hwan; Shin, Jae Chul; Wang, Lei Tian [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    1999-03-01

    A rotor dynamic analysis program DARBS/MCP, for the main coolant pump of the integral reactor, has been developed. The dynamic analysis model of the main coolant pump includes a vertical shaft, three grooved radial journal bearings and gaps that represent the structure-fluid interaction effects between the rotor and the lubricant fluid. The electromagnetic force from the motor and the hydro-dynamic force induced by impeller are the major sources of vibration that may affect the rotor system stability. DARBS/MCP is a software that is developed to effectively analyze the dynamics of MCP rotor systems effectively by applying powerful numerical algorithms such as FEM with modal truncation and {lambda}-matrix method for harmonic analysis. Main design control parameters, that have much influence to the dynamic stability, have been found by Taguchi's sensitivity analysis method. Design suggestions to improve the stability of MCP rotor system have been documented. The dynamic bearing parameters of the journal bearings used for main coolant pump have been determined by directly solving the Reynolds equation using FDM method. Fluid-structure interaction effect that occurs at the small gaps between the rotor and the stator were modeled as equivalent seals, the electromagnetic force effect was regarded as a linear negative radial spring and the impeller was modeled as a rigid disk with hydrodynamic and static radial force. Although there exist critical speeds in the range of operational speeds for type I and II rotor systems, the amplitude of vibration appears to be less than the vibration limit set by the API standards. Further more, it has been verified that the main design parameters such as the clearance and length of journal bearings, and the static radial force of impeller should be properly adjusted, in order to the improve dynamic stability of the rotor system. (author). 39 refs., 81 figs., 17 tabs.

  15. [The significance of the results of crash-tests with the use of the models of the pedestrians' lower extremities for the prevention of the traffic road accidents].

    Science.gov (United States)

    Smirenin, S A; Fetisov, V A; Grigoryan, V G; Gusarov, A A; Kucheryavets, Yu O

    The disabling injuries inflicted during road traffic accidents (RTA) create a serious challenge for the public health services and are at the same time a major socio-economic problem in the majority of the countries throughout the world. The injuries to the lower extremities of the pedestrians make up the largest fraction of the total number of the non-lethal RTA injuries. Most of them are responsible for the considerable deterioration of the quality of life for the participants in the accidents during the subsequent period. The objective of the present study was to summarize the currently available results of experimental testing of the biomechanical models of the pedestrians' lower extremities in the framework of the program for the prevention of the road traffic accidents as proposed by the World Health Organization (WHO, 2004). The European Enhanced Safety Vehicle Committee (EEVC) has developed a series of crash-tests with the use of the models of the pedestrians' lower extremities simulating the vehicle bumper-pedestrian impact. The models are intended for the assessment of the risk of the tibia fractures and the injuries to the knee joint ligaments. The experts of EEVC proposed the biomechanical criteria for the acceleration of the knee and talocrural parts of the lower limbs as well as for the shear displacement of the knee and knee-bending angle. The engineering solution of this problem is based on numerous innovation proposals being implemented in the machine-building industry with the purpose of reducing the stiffness of structural elements of the bumper and other front components of a modern vehicle designed to protect the pedestrians from severe injuries that can be inflicted in the road traffic accidents. The activities of the public health authorities (in the first place, bureaus of forensic medical expertise and analogous facilities) have a direct bearing on the solution of the problem of control of road traffic injuries because they are possessed of

  16. Predicting the conditions under which vibroacoustic resonances with external periodic loads occur in the primary coolant circuits of VVER-based NPPs

    Science.gov (United States)

    Proskuryakov, K. N.; Fedorov, A. I.; Zaporozhets, M. V.

    2015-08-01

    The accident at the Japanese Fukushima Daiichi nuclear power plant (NPP) caused by an earthquake showed the need of taking further efforts aimed at improving the design and engineering solutions for ensuring seismic resistance of NPPs with due regard to mutual influence of the dynamic processes occurring in the NPP building structures and process systems. Resonance interaction between the vibrations of NPP equipment and coolant pressure pulsations leads to an abnormal growth of dynamic stresses in structural materials, accelerated exhaustion of equipment service life, and increased number of sudden equipment failures. The article presents the results from a combined calculation-theoretical and experimental substantiation of mutual amplification of two kinds of external periodic loads caused by rotation of the reactor coolant pump (RCP) rotor and an earthquake. The data of vibration measurements at an NPP are presented, which confirm the predicted multiple amplification of vibrations in the steam generator and RCP at a certain combination of coolant thermal-hydraulic parameters. It is shown that the vibration frequencies of the main equipment may fall in the frequency band corresponding to the maximal values in the envelope response spectra constructed on the basis of floor accelerograms. The article presents the results from prediction of conditions under which vibroacoustic resonances with external periodic loads take place, which confirm the occurrence of additional earthquake-induced multiple growth of pressure pulsation intensity in the steam generator at the 8.3 Hz frequency and additional multiple growth of vibrations of the RCP and the steam generator cold header at the 16.6 Hz frequency. It is shown that at the elastic wave frequency equal to 8.3 Hz in the coolant, resonance occurs with the frequency of forced vibrations caused by the rotation of the RCP rotor. A conclusion is drawn about the possibility of exceeding the design level of equipment vibrations

  17. Cryogenic-coolant He-4-superconductor interaction

    Science.gov (United States)

    Caspi, S.; Lee, J. Y.; Kim, Y. I.; Allen, R. J.; Frederking, T. H. K.

    1978-01-01

    The thermodynamic and thermal interaction between a type 2 composite alloy and cryo-coolant He4 was studied with emphasis on post quench phenomena of formvar coated conductors. The latter were investigated using a heater simulation technique. Overall heat transfer coefficients were evaluated for the quench onset point. Heat flux densities were determined for phenomena of thermal switching between a peak and a recovery value. The study covered near saturated liquid, pressurized He4, both above and below the lambda transition, and above and below the thermodynamic critical pressure. In addition, friction coefficients for relative motion between formvar insulated conductors were determined.

  18. Accidents in nuclear ships

    Energy Technology Data Exchange (ETDEWEB)

    Oelgaard, P.L. [Risoe National Lab., Roskilde (Denmark)]|[Technical Univ. of Denmark, Lyngby (Denmark)

    1996-12-01

    This report starts with a discussion of the types of nuclear vessels accidents, in particular accidents which involve the nuclear propulsion systems. Next available information on 61 reported nuclear ship events in considered. Of these 6 deals with U.S. ships, 54 with USSR ships and 1 with a French ship. The ships are in almost all cases nuclear submarines. Only events that involve the sinking of vessels, the nuclear propulsion plants, radiation exposures, fires/explosions, sea-water leaks into the submarines and sinking of vessels are considered. For each event a summary of available information is presented, and comments are added. In some cases the available information is not credible, and these events are neglected. This reduces the number of events to 5 U.S. events, 35 USSR/Russian events and 1 French event. A comparison is made between the reported Soviet accidents and information available on dumped and damaged Soviet naval reactors. It seems possible to obtain good correlation between the two types of events. An analysis is made of the accident and estimates are made of the accident probabilities which are found to be of the order of 10{sup -3} per ship reactor years. It if finally pointed out that the consequences of nuclear ship accidents are fairly local and does in no way not approach the magnitude of the Chernobyl accident. It is emphasized that some of the information on which this report is based, may not be correct. Consequently some of the results of the assessments made may not be correct. (au).

  19. An approach for IC engine coolant energy recovery based on low-temperature organic Rankine cycle

    Institute of Scientific and Technical Information of China (English)

    付建勤; 刘敬平; 徐政欣; 邓帮林; 刘琦

    2015-01-01

    To promote the fuel utilization efficiency of IC engine, an approach was proposed for IC engine coolant energy recovery based on low-temperature organic Rankine cycle (ORC). The ORC system uses IC engine coolant as heat source, and it is coupled to the IC engine cooling system. After various kinds of organic working media were compared, R124 was selected as the ORC working medium. According to IC engine operating conditions and coolant energy characteristics, the major parameters of ORC system were preliminary designed. Then, the effects of various parameters on cycle performance and recovery potential of coolant energy were analyzed via cycle process calculation. The results indicate that cycle efficiency is mainly influenced by the working pressure of ORC, while the maximum working pressure is limited by IC engine coolant temperature. At the same working pressure, cycle efficiency is hardly affected by both the mass flow rate and temperature of working medium. When the bottom cycle working pressure arrives at the maximum allowable value of 1.6 MPa, the fuel utilization efficiency of IC engine could be improved by 12.1%. All these demonstrate that this low-temperature ORC is a useful energy-saving technology for IC engine.

  20. Public dose calculations in case of accident: major differences between American and German standards and influence on the results; Calculos de dosis al publico en caso de accidente: principales diferencias entre normativas americanas y alemanas e influencia sobre los resultados

    Energy Technology Data Exchange (ETDEWEB)

    Munoz Cervantes, A.

    2010-07-01

    Estimates of the radiological consequences of accidents are part of the safety studies of a nuclear power plant. The criterion used today is the effective dose received by the public as a result of an accident. The country standards set a limit on that dose, and that value can vary significantly from one country to another, as seen in the second chapter of this paper.

  1. Accident: Reminder

    CERN Document Server

    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

  2. Advances in operational safety and severe accident research

    Energy Technology Data Exchange (ETDEWEB)

    Simola, K. [VTT Automation (Finland)

    2002-02-01

    A project on reactor safety was carried out as a part of the NKS programme during 1999-2001. The objective of the project was to obtain a shared Nordic view of certain key safety issues related to the operating nuclear power plants in Finland and Sweden. The focus of the project was on selected central aspects of nuclear reactor safety that are of common interest for the Nordic nuclear authorities, utilities and research bodies. The project consisted of three sub-projects. One of them concentrated on the problems related to risk-informed deci- sion making, especially on the uncertainties and incompleteness of probabilistic safety assessments and their impact on the possibilities to use the PSA results in decision making. Another sub-project dealt with questions related to maintenance, such as human and organisational factors in maintenance and maintenance management. The focus of the third sub-project was on severe accidents. This sub-project concentrated on phenomenological studies of hydrogen combustion, formation of organic iodine, and core re-criticality due to molten core coolant interaction in the lower head of reactor vessel. Moreover, the current status of severe accident research and management was reviewed. (au)

  3. Source term analysis for a nuclear submarine accident

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B.J.; Hugron, J.J.M.R. [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    1999-07-01

    A source term analysis has been conducted to determine the activity release into the environment as a result of a large-break loss-of-coolant accident aboard a visiting nuclear-powered submarine to a Canadian port. This best-estimate analysis considers the fractional release from the core, and fission product transport in the primary heat transport system, primary containment (i.e. reactor compartment) and submarine hull. Physical removal mechanisms such as vapour and aerosol deposition are treated in the calculation. Since a thermalhydraulic analysis indicated that the integrity of the reactor compartment is maintained, release from the reactor compartment will only occur by leakage; however, it is conservatively assumed that the secondary containment is not isolated for a 24-h period where release occurs through an open hatch in the submarine hull. Consequently, during this period, the activity release into the atmosphere is estimated as 4.6 TBq, leading to a maximum individual dose equivalent of 0.5 mSv at 800 metres from the berthing location. This activity release is comparable to that obtained in the BEREX TSA study (for a similar accident scenario) but is four orders of magnitude less than that reported in the earlier Davis study where, unrealistically, no credit had been taken for the containment system or for any physical removal processes. (author)

  4. Membrane systems and their use in nuclear power plants. Treatment of primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Kus, Pavel; Bartova, Sarka; Skala, Martin; Vonkova, Katerina [Research Centre Rez, Husinec-Rez (Czech Republic). Technological Circuits Innovation Dept.; Zach, Vaclav; Kopa, Roman [CEZ a.s., Temelin (Czech Republic). Nuclear Power Plant Temelin

    2016-03-15

    In nuclear power plants, drained primary coolant containing boric acid is currently treated in the system of evaporators and by ion exchangers. Replacement of the system of evaporators by membrane system (MS) will result in lower operating cost mainly due to lower operation temperature. In membrane systems the feed primary coolant is separated into two output streams: retentate and permeate. Retentate stream consists of the concentrated boric acid solution together with other components, while permeate stream consists of purified water. Results are presented achieved by testing a pilot-plant unit of reverse osmosis in nuclear power plant (NPP) Temelin.

  5. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  6. Measurement of Coolant in a Flat Heat Pipe Using Neutron Radiography

    Science.gov (United States)

    Mizuta, Kei; Saito, Yasushi; Goshima, Takashi; Tsutsui, Toshio

    A newly developed flat heat pipe FGHPTM (Morex Kiire Co.) was experimentally investigated by using neutron radiography. The test sample of the FGHP heat spreader was 65 × 65 × 2 mm3 composed of several etched copper plates and pure water was used as the coolant. Neutron radiography was performed at the E-2 port of the Kyoto University Research Reactor (KUR). The coolant distributions in the wick area of the FGHP and its heat transfer characteristics were measured at heating conditions. Experimental results show that the coolant distributions depend slightly on its installation posture and that the liquid thickness in the wick region remains constant with increasing heat input to the FGHP. In addition, it is found that the wick surface does not dry out even in the vertical posture at present experimental conditions.

  7. Simulating the corrosion of zirconium alloys in the water coolant of VVER reactors

    Science.gov (United States)

    Kritskii, V. G.; Berezina, I. G.; Motkova, E. A.

    2013-07-01

    A model for predicting the corrosion of cladding zirconium alloys depending on their composition and operating conditions is proposed. Laws of thermodynamics and chemical kinetics of the reactions through which the multicomponent zirconium alloy is oxidized in the reactor coolant constitute the physicochemical heart of the model. The developed version of the model is verified against the results obtained from tests of fuel rod claddings made of commercial-grade and experimental zirconium alloys carried out by different researchers under autoclave and reactor conditions. It is shown that the proposed model adequately describes the corrosion of alloys in coolants used at nuclear power stations. It is determined that, owing to boiling of coolant and its acidification in a VVER-1200 reactor, Zr-1% Nb alloys with additions of iron and oxygen must be more resistant to corrosion than the commercial-grade alloy E110.

  8. Pressure Load Analysis during Severe Accidents for the Evaluation of Late Containment Failure in OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The MAAP code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a level 2 probabilistic safety assessment or severe accident management strategy developments. The code employs lots of user-options for supporting a sensitivity and uncertainty analysis. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to in-vessel hydrogen generation, gas combustion in the containment, corium distribution in the containment after a reactor vessel failure, corium coolability in the reactor cavity, and molten-corium interaction with concrete. The phenomenology of severe accidents is extremely complex. In this paper, a sampling-based phenomenological uncertainty analysis was performed to statistically quantify uncertainties associated with the pressure load of a containment building for a late containment failure evaluation, based on the key modeling parameters employed in the MAAP code and random samples for those parameters. Phenomenological issues surrounding the late containment failure mode are highly complex. Included are the pressurization owing to steam generation in the cavity, molten corium-concrete interaction, late hydrogen burn in the containment, and the secondary heat removal availability. The methodology and calculation results can be applied for the optimum assessment of a late containment failure model. The accident sequences considered were a loss of coolant accidents and loss of offsite accidents expected in the OPR-1000 plant. As a result, uncertainties addressed in the pressure load of the containment building were quantified as a function of time. A realistic evaluation of the mean and variance estimates provides a more complete

  9. Improvements of primary coolant shutdown chemistry and reactor coolant system cleanup

    Energy Technology Data Exchange (ETDEWEB)

    Gaudard, G.; Gilles, B.; Mesnage, F. [EDF/GDL (France); Cattant, F. [EDF R and D (France)

    2002-07-01

    In the framework of a radiation exposure management program entitled <>, EDF aims at decreasing the mass dosimetry of nuclear power plants workers. So, the annual dose per unit, which has improved from 2.44 m.Sv in 1991 to 1.08 in 2000, should target 0.8 mSv in the year 2005 term in order to meet the results of the best nuclear operators. One of the guidelines for irradiation source term reduction is the optimization of operation parameters, including reactor coolant system (RCS) chemistry in operation, RCS shutdown chemistry and RCS cleanup improvement. This paper presents the EDF strategy for the shutdown and start up RCS chemistry optimization. All the shutdown modes have been reviewed and for each of them, the chemical specifications will be fine tuned. A survey of some US PWRs shutdown practices has been conducted for an acid and reducing shutdown chemistry implementation test at one EDF unit. This survey shows that deviating from the EPRI recommended practice for acid and reducing shutdown chemistry is possible and that critical path impact can be minimized. The paper also presents some investigations about soluble and insoluble species behavior and characterization; the study focuses here on {sup 110m}Ag, {sup 122}Sb, {sup 124}Sb and iodine contamination. Concerning RCS cleanup improvement, the paper presents two studies. The first one highlights some limited design modifications that are either underway or planned, for an increased flow rate during the most critical periods of the shutdown. The second one focuses on the strategy EDF envisions for filters and resins selection criteria. Matching the study on contaminants behavior with the study of filters and resins selection criteria should allow improving the cleanup efficiency. (authors)

  10. Experimental investigation of thermoelectric power generation versus coolant pumping power in a microchannel heat sink

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Rosendahl, Lasse; Andreasen, Søren Juhl

    2012-01-01

    The coolant heat sinks in thermoelectric generators (TEG) play an important role in order to power generation in the energy systems. This paper explores the effective pumping power required for the TEGs cooling at five temperature difference of the hot and cold sides of the TEG. In addition......, the temperature distribution and the pressure drop in sample microchannels are considered at four sample coolant flow rates. The heat sink contains twenty plate-fin microchannels with hydraulic diameter equal to 0.93 mm. The experimental results show that there is a unique flow rate that gives maximum net...

  11. Thermostat-controlled coolant pump - a new concept for fuel saving

    Energy Technology Data Exchange (ETDEWEB)

    Etemad, S. [Volvo Car Components Corp., Gothenburg (Sweden); Anderson, A. [Volvo Truck Corp., Gothenburg (Sweden)

    1999-07-01

    A new coolant pump concept has been developed for better fuel economy. The flow returning from the radiator is fed coaxially into the pump. The by-pass flow is fed tangentially into the pump, generating a pre-swirl with the same direction of rotation as the coolant pump impeller. The relative velocity between the flow and the impeller decreases. This reduces the transferred momentum from the impeller to the fluid, reducing the power consumption. The flow split between the radiator and the by-pass channel is controlled by the ordinary thermostat. Results from analysis and measurements are presented. (author)

  12. Experimental simulation of low rate primary coolant leaks. For the case of vessel head penetrations affected by through wall cracking

    Energy Technology Data Exchange (ETDEWEB)

    You, D.; Feron, D. [CEA-Saclay - DEN/DPC/SCCME, 91 - Gif-sur-Yvette (France); Turluer, G. [CEA-Fontenay-aux-Roses - IPSN/DES/SAMS, 92 - Fontenay-aux-Roses (France)

    2002-07-01

    An experimental simulation of primary coolant leaks was carried out to determine how the composition of the leaking liquid would change. The experiment used the EVA experimental setup, specially designed for quantitatively investigating concentration phenomena driven by evaporation. The test showed that the final composition, obtained from a solution representative of the primary coolant at the beginning of the cycle, is highly concentrated and slightly acid. The experimental results are compared with those obtained using the MULTEQ software. (authors)

  13. Use of ethanolamine for alkalization of secondary coolant. First experience at VVER reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smiesko, I. [NPP Jaslovske Bohunice (Slovakia); Bystriansky, J. [TEDIS-KOR, Dobra (Czech Republic); Szalo, A. [NPPRI Trnava (Slovakia)

    2002-07-01

    The paper summarises preparatory work and results of six-week plant trial aimed at use of ethanolamine for alkalization of secondary coolant. Operational data in pre-test and test period are given and outage inspection results are commented. Future plans are outlined. (authors)

  14. Partial Discharge Measurements in HV Rotating Machines in Dependence on Pressure of Coolant

    Directory of Open Access Journals (Sweden)

    I. Kršňák

    2002-01-01

    Full Text Available The influence of the pressure of the coolant used in high voltage rotating machines on partial discharges occurring in stator insulation is discussed in this paper. The first part deals with a theoretical analysis of the topic. The second part deals with the results obtained on a real generator in industrial conditions. Finally, theoretical assumptions and obtained results are compared.

  15. Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.

    2016-08-01

    The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

  16. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A. [St. Petersburg State Technical Univ. (Russian Federation)

    1997-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

  17. Simulation of fuel dispersion in the MYRRHA-FASTEF primary coolant with CFD and SIMMER-IV

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, Sophia, E-mail: sophia.buckingham@vki.ac.be [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Eboli, Marica [University of Pisa, Largo Lucio Lazzarino 2, 56122 Pisa (Italy); Moreau, Vincent [CRS4, Science and Technology Park Polaris – Piscina Manna, 09010 Pula (Italy); Van Tichelen, Katrien [SCK-CEN, Boeretang 200, 2400 Mol (Belgium)

    2015-12-15

    Highlights: • A comparison between CFD and system codes applied to long-term dispersion of fuel particles inside the MYRRHA reactor is proposed. • Important accumulations at the free-surface level are to be expected. • The risk of core blockage should not be neglected. • Numerical approach and modeling assumptions have a strong influence on the simulation results and accuracy. - Abstract: The objective of this work is to assess the behavior of fuel redistribution in heavy liquid metal nuclear systems under fuel pin failure conditions. Two different modeling approaches are considered using Computational Fluid Dynamics (CFD) codes and a system code, applied to the MYRRHA facility primary coolant loop version 1.4. Two different CFD models are constructed: the first is a single-phase steady model prepared in ANSYS Fluent, while the second is a two-phase model based on the volume of fluid (VOF) method in STARCCM+ to capture the upper free-surface dynamics. Both use a Lagrangian tracking approach with oneway coupling to follow the particles throughout the reactor. The system code SIMMER-IV is used for the third model, without neutronic coupling. Although limited regarding the fluid dynamic aspects compared to the CFD codes, comparisons of particle distributions highlight strong similarities despite quantitative discrepancies in the size of fuel accumulations. These disparities should be taken into account while performing the safety analysis of nuclear systems and developing strategies for accident mitigation.

  18. Failure rates in Barsebaeck-1 reactor coolant pressure boundary piping. An application of a piping failure database

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, B. [RSA Technologies, Vista, CA (United States)

    1999-05-01

    This report documents an application of a piping failure database to estimate the frequency of leak and rupture in reactor coolant pressure boundary piping. The study used Barsebaeck-1 as reference plant. The study tried two different approaches to piping failure rate estimation: 1) PSA-style, simple estimation using Bayesian statistics, and 2) fitting of statistical distribution to failure data. A large, validated database on piping failures (like the SKI-PIPE database) supports both approaches. In addition to documenting leak and rupture frequencies, the SKI report describes the use of piping failure data to estimate frequency of medium and large loss of coolant accidents (LOCAs). This application study was co sponsored by Barsebaeck Kraft AB and SKI Research 41 refs, figs, tabs

  19. Power module assemblies with staggered coolant channels

    Science.gov (United States)

    Herron, Nicholas Hayden; Mann, Brooks S; Korich, Mark D

    2013-07-16

    A manifold is provided for supporting a power module assembly with a plurality of power modules. The manifold includes a first manifold section. The first face of the first manifold section is configured to receive the first power module, and the second face of the first manifold section defines a first cavity with a first baseplate thermally coupled to the first power module. The first face of the second manifold section is configured to receive the second power module, and the second face of the second manifold section defines a second cavity with a second baseplate thermally coupled to the second power module. The second face of the first manifold section and the second face of the second manifold section are coupled together such that the first cavity and the second cavity form a coolant channel. The first cavity is at least partially staggered with respect to second cavity.

  20. Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

    OpenAIRE

    Bo Cheng; Young-Jin Kim; Peter Chou

    2016-01-01

    In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident managem...

  1. Traffic Accidents on Slippery Roads

    DEFF Research Database (Denmark)

    Fonnesbech, J. K.; Bolet, Lars

    2014-01-01

    Police registrations from 65 accidents on slippery roads in normally Danish winters have been studied. The study showed: • 1 accident per 100 km when using brine spread with nozzles • 2 accidents per 100 km when using pre wetted salt • 3 accidents per 100 km when using kombi spreaders The results...... of accidents in normally Danish winter seasons are remarkable alike the amount of salt used in praxis in the winter 2011/2012. • 2.7 ton NaCl/km when using brine spread with nozzles • 5 ton NaCl/km when using pre wetted salt. • 5.7 ton NaCl/km when using kombi spreaders The explanation is that spreading...

  2. Industrial accidents triggered by lightning.

    Science.gov (United States)

    Renni, Elisabetta; Krausmann, Elisabeth; Cozzani, Valerio

    2010-12-15

    Natural disasters can cause major accidents in chemical facilities where they can lead to the release of hazardous materials which in turn can result in fires, explosions or toxic dispersion. Lightning strikes are the most frequent cause of major accidents triggered by natural events. In order to contribute towards the development of a quantitative approach for assessing lightning risk at industrial facilities, lightning-triggered accident case histories were retrieved from the major industrial accident databases and analysed to extract information on types of vulnerable equipment, failure dynamics and damage states, as well as on the final consequences of the event. The most vulnerable category of equipment is storage tanks. Lightning damage is incurred by immediate ignition, electrical and electronic systems failure or structural damage with subsequent release. Toxic releases and tank fires tend to be the most common scenarios associated with lightning strikes. Oil, diesel and gasoline are the substances most frequently released during lightning-triggered Natech accidents.

  3. How to reduce the number of accidents

    CERN Multimedia

    2012-01-01

    Among the safety objectives that the Director-General has established for CERN in 2012 is a reduction in the number of workplace accidents.   The best way to prevent workplace accidents is to learn from experience. This is why any accident, fire, instance of pollution, or even a near-miss, should be reported using the EDH form that can be found here. All accident reports are followed up. The departments investigate all accidents that result in sick leave, as well as all the more common categories of accidents at CERN, essentially falls (slipping, falling on stairs, etc.), regardless of whether or not they lead to sick leave. By studying the accident causes that come to light in this way, it is possible to take preventive action to avoid such accidents in the future. If you have any questions, the HSE Unit will be happy to answer them. Contact us at safety-general@cern.ch. HSE Unit

  4. Reactor coolant pump shaft seal behavior during station blackout

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. 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....

  6. ACCIDENT AT «FUKUSHIMA-»I NPP: FIRST RESULTS OF EMERGENCY RESPONSE REPORT 2: ACTIVITIES OF THE ROSPOTREBNADZOR AUTHORITIES FOR THE RADIATION PROTECTION OF THE RUSSIAN FEDERATION POPULATION ON THE EARLY STAGE OF ACCIDENT

    Directory of Open Access Journals (Sweden)

    G. G. Onischenko

    2011-01-01

    Full Text Available Basic measures of the Rospotrebnadzor emergency response during the early stage of the «Fukushima-I» NPP radiation accident are being analyzed in the article. Radiation monitoring of the environmental objects of the territories of the Russian Federation Far East Federal District, radiation control of arriving from Japan vehicles, freights and passengers as well as imported from Japan food products were promptly organized. This allowed to get reliable evaluations of the levels of radioactive contamination at the Russian Federation territory and population exposure doses due to the «Fukushima-I» NPP accident, timely exclude the possibility of import to the Russian territory for the freights, vehicles, food products having contamination exceeding established in the Russian Federation standards.

  7. Leak rate analysis of the Westinghouse Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Boardman, T.; Jeanmougin, N.; Lofaro, R.; Prevost, J.

    1985-07-01

    An independent analysis was performed by ETEC to determine what the seal leakage rates would be for the Westinghouse Reactor Coolant Pump (RCP) during a postulated station blackout resulting from loss of ac electric power. The object of the study was to determine leakage rates for the following conditions: Case 1: All three seals function. Case 2: No. 1 seal fails open while Nos. 2 and 3 seals function. Case 3: All three seals fail open. The ETEC analysis confirmed Westinghouse calculations on RCP seal performance for the conditions investigated. The leak rates predicted by ETEC were slightly lower than those predicted by Westinghouse for each of the three cases as summarized below. Case 1: ETEC predicted 19.6 gpm, Westinghouse predicted 21.1 gpm. Case 2: ETEC predicted 64.7 gpm, Westinghouse predicted 75.6 gpm. Case 3: ETEC predicted 422 gpm, Westinghouse predicted 480 gpm. 3 refs., 22 figs., 6 tabs.

  8. Potassium iodide for thyroid blockade in a reactor accident: administrative policies that govern its use.

    Science.gov (United States)

    Becker, D V; Zanzonico, P

    1997-04-01

    A marked increase in thyroid cancer among young children who were in the vicinity of the Chernobyl nuclear power plant at the time of the 1986 accident strongly suggests a possible causal relationship to the large amounts of radioactive iodine isotopes in the resulting fallout. Although remaining indoors, restricting consumption of locally produced milk and foodstuffs, and evacuation are important strategies in a major breach-of-containment accident, stable potassium iodide (KI) prophylaxis given shortly before or immediately after exposure can reduce greatly the thyroidal accumulation of radioiodines and the resulting radiation dose. Concerns about possible side effects of large-scale, medically unsupervised KI consumption largely have been allayed in light of the favorable experience in Poland following the Chernobyl accident; 16 million persons received single administrations of KI with only rare occurrence of side effects and with a probable 40% reduction in projected thyroid radiation dose. Despite the universal acceptance of KI as an effective thyroid protective agent, supplies of KI in the US are not available for public distribution in the event of a reactor accident largely because government agencies have argued that stockpiling and distribution of KI to other than emergency workers cannot be recommended in light of difficult distribution logistics, problematic administrative issue, and a calculated low cost-effectiveness. However, KI in tablet form is expensive and has a long shelf life, and many countries have largely stockpiles and distribution programs. The World Health Organization recognizes the benefits of stable KI and urges its general availability. At present there are 110 operating nuclear power plants in the US and more than 300 in the rest of the world. These reactors product 17% of the world's electricity and in some countries up to 60-70% of the total electrical energy. Almost all US nuclear power plants have multistage containment

  9. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  10. Diesel engine coolant analysis, new application for established instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D.P.; Lukas, M.; Lynch, B.K. [Spectro Incorporated, Littleton, MA (United States)

    1997-12-31

    Rotating disk electrode (RDE) arc emission spectrometers are user` many commercial, industrial and military laboratories throughout the world to analyze millions of oil and fuel samples each year. In fact, RDE spectrometers have been used exclusively for oil and fuel analysis for so long that it has nearly been forgotten by most practitioners that when RDE spectrometers were first introduced more than 40 years ago, they were routinely used for aqueous samples as well. This presentation reviews early methods of aqueous sample analysis using RDE technology. This presentation also describes recent work to calibrate an RDE spectrometer for both water samples and for engine coolant samples which are a mixture of approximately 50 % water and 50 % ethylene or propylene glycol. Limits of detection determined for aqueous standards are comparable to limits of detection for oil standards. Repeatability of aqueous samples is comparable to the repeatability achieved for oil samples. A comparison of results for coolant samples measured by both inductively coupled plasma (ICP) and rotating disk electrode (RDE) spectrometers is presented. Not surprisingly, RDE results are significantly higher for samples containing particles larger than a few micrometers. Although limits of detection for aqueous samples are not as low as can be achieved using the more modern ICP spectrometric method or the more cumbersome atomic absorption (AA) method, this presentation suggests that RDE spectrometers may be appropriate for certain types of aqueous samples in situations where the more sensitive ICP or AA spectrometers and the laboratory environment and skilled personnel needed for them to operate are not conveniently available. (orig.) 4 refs.

  11. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2005-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

  12. Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

    Energy Technology Data Exchange (ETDEWEB)

    Banati, J. [Lappeenranta Univ. of Technology, Lappeenranta (Finland)

    1995-09-01

    This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

  13. Analysis of Early Severe Accident Initiated by LBLOCA for Qinshan Phase II Nuclear Power Project

    Directory of Open Access Journals (Sweden)

    Shi Xing-Wei

    2013-07-01

    Full Text Available The purpose of this study is to simulate an early Severe Accident (SA scenario more detail through transferring the thermal-hydraulic status of the plant predicted by RELAP5 computer code to SA Program (SAP. Based on the criterion of date extract time, the RELAP5 thermal-hydraulic calculation data is extracted to form a file for SAP input card at 1477K of cladding surface. Relying on the thermal-hydraulic boundary parameters calculated by RELAP5 code, analysis of early SA initiated by the Large Break Loss-of-Coolant Accident (LBLOCA without mitigation measures for Qinshan Phase II Nuclear Power Plant (QSP-II performed by SAP through finding the key events of accident sequence, estimating the amount of hydrogen generation and oxidation behavior of the cladding and evaluating the relocation order of the materials collapsed in the central region of the core. The results of this study are expected to improve the SA analysis methodology more detail through analyzing early SA scenario.

  14. Study on effects of mixing vane grids on coolant temperature distribution by subchannel analysis

    Energy Technology Data Exchange (ETDEWEB)

    Mao, H.; Yang, B.W.; Han, B. [Xi' an Jiaotong Univ., Shaanxi (China). Science and Technology Center for Advanced Nuclear Fuel Research

    2016-07-15

    Mixing vane grids (MVG) have great influence on coolant temperature field in the rod bundle. The MVG could enhance convective heat transfer between the fuel rod wall and the coolant, and promote inter-subchannel mixing at the same time. For the influence of the MVG on convective heat transfer enhancement, many experiments have been done and several correlations have been developed based on the experimental data. However, inter-subchannel mixing promotion caused by the MVG is not well estimated in subchannel analysis because the information of mixing vanes is totally missing in most subchannel codes. This paper analyzes the influence of mixing vanes on coolant temperature distribution using the improved MVG model in subchannel analysis. The coolant temperature distributions with the MVG are analyzed, and the results show that mixing vanes lead to a more uniform temperature distribution. The performances of split vane grids under different power conditions are evaluated. The results are compared with those of spacer grids without mixing vanes and some conclusions are obtained.

  15. The upgrade of intense pulsed neutron source (IPNS) through the change of coolant and reflector

    CERN Document Server

    Baek, I C; Iverson, E B

    2002-01-01

    The current intense pulsed neutron source (IPNS) depleted uranium target is cooled by light water. The inner reflector material is graphite and the outer reflector material is beryllium. The presence of H sub 2 O in the target moderates neutrons and leads to a higher absorption loss in the target than is necessary. D sub 2 O coolant in the small quantities required minimizes this effect. We have studied the possible improvement in IPNS beam fluxes that would result from changing the coolant from H sub 2 O to D sub 2 O and the inner reflector from graphite to beryllium. Neutron intensities were calculated for directions normal to the viewed surface of each moderator for four different cases of combinations of target coolant and reflector materials. The simulations reported here were performed using the MCNPX (version 2.1.5) computer program. Our results show that substantial gains in neutron beam intensities can be achieved by appropriate combination of target coolant and reflector materials. The combination o...

  16. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences.

  17. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

    1995-10-01

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  18. Long-term environmental radioactive contamination of Europe due to the Chernobyl accident - Results of the Joint Danube Survey 2013.

    Science.gov (United States)

    Maringer, Franz Josef; Ackerl, Claudia; Baumgartner, Andreas; Burger-Scheidlin, Christopher; Kocadag, Maria; Sterba, Johannes H; Stietka, Michael; Welch, Jan Matthew

    2017-02-27

    In the course of the Joint Danube Survey 3 (JDS3), coordinated by the International Commission for the Protection of the Danube River (ICPDR), laboratory ships travelled 2375km down the Danube River engaging in sampling, processing and on-board analyses during the summer of 2013. The results of the radiometric analysis of (90)Sr, (137)Cs and natural radionuclides in recent riverbed sediment are presented. The activity concentrations of (90)Sr and (137)Cs in Danube sediments have been found below 100Bq/kg.

  19. 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.

  20. ANS severe accident program overview & planning document

    Energy Technology Data Exchange (ETDEWEB)

    Taleyarkhan, R.P.

    1995-09-01

    The Advanced Neutron Source (ANS) severe accident document was developed to provide a concise and coherent mechanism for presenting the ANS SAP goals, a strategy satisfying these goals, a succinct summary of the work done to date, and what needs to be done in the future to ensure timely licensability. Guidance was received from various bodies [viz., panel members of the ANS severe accident workshop and safety review committee, Department of Energy (DOE) orders, Nuclear Regulatory Commission (NRC) requirements for ALWRs and advanced reactors, ACRS comments, world-wide trends] were utilized to set up the ANS-relevant SAS goals and strategy. An in-containment worker protection goal was also set up to account for the routine experimenters and other workers within containment. The strategy for achieving the goals is centered upon closing the severe accident issues that have the potential for becoming certification issues when assessed against realistic bounding events. Realistic bounding events are defined as events with an occurrency frequency greater than 10{sup {minus}6}/y. Currently, based upon the level-1 probabilistic risk assessment studies, the realistic bounding events for application for issue closure are flow blockage of fuel element coolant channels, and rapid depressurization-related accidents.

  1. Reentry response of the light weight radioisotope heater unit resulting from a Venus-Earth-Earth Gravity Assist maneuver accident

    Energy Technology Data Exchange (ETDEWEB)

    Hagan, J.C.

    1988-10-01

    Reentry analyses consisting of ablation response, thermal response and thermal stress response have been conducted on the Light Weight Radioisotope Heater Unit for Galileo/VEEGA reentry conditions. Sequential ablation analyses of the LWRHU aeroshell, the fuel clad, and the fuel pellet have been conducted in reentry regimes where the aeroshell has been deemed to fail. The failure criterion for ablation is assumed to be recession corresponding to 50% of the wall thickness (the design criterion recommended in the DOE Overall Safety Manual). Although the analyses have been carried far beyond this limit (as presented and discussed herein), JHU/APL endorses the position that failure may occur at the time that this recession is achieved or at lower altitudes within the heat pulse considering the uncertainties in the aerodynamic, thermodynamic, and thermo-structural analyses and modeling. These uncertainties result mainly because of the high energies involved in the VEEGA reentries compared to orbital decay reentries. Risk evaluations should consider the fact that for shallow flight paths the unit may disassemble at high-altitude as a result of ablation or may remain intact until it impacts with a clad that had been molten. 80 refs., 46 figs., 16 tabs.

  2. Aerial Survey Results for 131I Deposition on the Ground after the Fukushima Daiichi Nuclear Power Plant Accident

    Energy Technology Data Exchange (ETDEWEB)

    Torii, Tatsuo [JAEA; Sugita, Takeshi [JAEA; Okada, Colin E. [NSTec; Reed, Michael S. [NSTec; Blumenthal, Daniel J. [NNSA

    2013-08-01

    In March 2011 the second largest accidental release of radioactivity in history occurred at the Fukushima Daiichi nuclear power plant following a magnitude 9.0 earthquake and subsequent tsunami. Teams from the U.S. Department of Energy, National Nuclear Security Administration Office of Emergency Response performed aerial surveys to provide initial maps of the dispersal of radioactive material in Japan. The initial results from the surveys did not report the concentration of 131I. This work reports on analyses performed on the initial survey data by a joint Japan-US collaboration to determine 131I ground concentration. This information is potentially useful in reconstruction of the inhalation and external exposure doses from this short-lived radionuclide. The deposited concentration of 134Cs is also reported.

  3. Lubricant-coolant fluid for machining metals

    Energy Technology Data Exchange (ETDEWEB)

    Berlin, A.A.; Epshtein, V.R.; Pastunov, V.A.; Sherle, A.I.; Shpin' kov, V.A.; Sladkova, T.A.

    1981-03-10

    For improving the antiwear and anticorrosion properties, the lubricant-coolant fluid (LCF) based on water, triethanolamine, and NaNO/sub 2/ contains additionally the sodium salt of an acid ester of maleic acid and substituted oligooxyethylenes (NMO) with the following proportions of the components: triethanolamine 0.3-0.5%, NaNO/sub 2/ 0.3-0.5%, NMO 0.5-2.0%, and water the remainder. In the case of using the proposed LCF on high-speed machine tools, it can contain additionally a foam suppressor in an amount of 0.005-0.1%. For preventing microbiological contamination of the LCF, bactericides of the type furacillin, formalin, vazin (transliteration), and others in an amount of 0.005-0.1% can be added to its composition. Introduction of the NMO additive ensures high wetting and lubricating characteristics in the LCF, which is characterized by stability during storage and service and good anticorrosion properties. Use of the proposed LCF makes it possible to increase the life of the cutting tool by a factor of 2.2 in machining Steel 40Kh and by a factor of 1.3 in machining corroding steel by comparison with the prototype; at the same time the service life of the LCF is increased twofold. The LCF can be used in machining parts of alloyed construction and corrosionresistant steels with cutting-edge and abrasive tools.

  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. CANDU heat sinks improvements as a follow up to Fukushima Daiichi accident ''the regulator perspective''

    Energy Technology Data Exchange (ETDEWEB)

    Mesmous, Noreddine; Harwood, Chris [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2015-06-15

    The purpose of this paper is to provide a summary of the Canadian Nuclear Safety Commission (CNSC) recommendations related to improving the heat sink strategy as a follow up to the Fukushima Daiichi Accident (FDA). As a follow up to FDA, CNSC staff tasked the Nuclear Power Plant (NPP) licensees to review the lessons learned from the FDA and re-examine the NPP safety cases. The reviews have examined the CANDU defence-in-depth strategy and considered events more severe than those that have historically been regarded as credible, and evaluated their impact on the NPPs safety. Availability of emergency equipment was shown to be crucial during the FDA and its availability could have arrested the accident progression early enough to minimize any radioactive release to the environment. As a result, licensees presented appropriate evaluations of the means to provide coolant make-up to the primary Heat Transport System (HTS), boilers, moderator, calandria vault, and irradiated fuel pools.

  6. Atmospheric transport and deposition of radionuclides released after the Fukushima Dai-chi accident and resulting effective dose

    Science.gov (United States)

    Marzo, Giuseppe A.

    2014-09-01

    On 11 March 2011 an earthquake off the Pacific coast of the Fukushima prefecture generated a tsunami that hit Fukushima Dai-ichi and Fukushima Da-ini Nuclear Power Plants. From 12 March a significant amount of radioactive material was released into the atmosphere and dispersed worldwide. Among the most abundant radioactive species released were iodine and cesium isotopes. By means of an atmospheric dispersion Lagrangian code and publicly available meteorological data, the atmospheric dispersion of 131I, 134Cs, and 137Cs have been simulated for three months after the event with a spatial resolution of 0.5° × 0.5° globally. The simulation has been validated by comparison to publicly available measurements collected in 206 locations worldwide. Sensitivity analysis shows that release height of the radionuclides, wet deposition velocity, and source term are the parameters with the most impact on the simulation results. The simulation shows that the radioactive plume, consisting of about 200 PBq by adding contributions from 131I, 134Cs, and 137Cs, has been transported over the entire northern hemisphere depositing up to 1.2 MBq m-2 nearby the NPPs to less than 20 Bq m-2 in Europe. The consequent effective dose to the population over a 50-year period, calculated by considering both external and internal pathways of exposure, is found to be about 40 mSv in the surroundings of Fukushima Dai-ichi, while other countries in the northern hemisphere experienced doses several orders of magnitude lower suggesting a small impact on the population health elsewhere.

  7. Chernobyl accident and its consequences

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.; Bonell, P.G.; Hicks, D.

    1987-01-01

    The USSR power reactor programme is first described. The reasons for the accident at the Chernobyl-4 RBMK nuclear reactor on 26 April 1986, the sequence of events that took place, and the immediate and long-term consequences are considered. A description of the RBMK-type reactors is given and the design changes resulting from the experience of the accident are explained. The source terms describing the details of the radioactivity release associated with the accident and the environmental consequences are covered in the last two sections of the report. Throughout the text comments referring to the UK Nuclear Installations Inspectorate Safety assessment principles have been inserted. (U.K.).

  8. Analysis of the results of environmental monitoring in Spain after the Chernobyl and Fukushima accidents; Analisis de los resultados de la vigilancia radiologica ambiental en Espana tras los accidentes de Chernobil y Fukusima

    Energy Technology Data Exchange (ETDEWEB)

    Luque Heredia, S.; Salas Collantes, R.; Rey del castillo, C.; Marugar tovar, I.; Sterling carmona, A.; Ramos Salvador, L.; Lorente Lorente, P.

    2013-07-01

    As a result of accidents at Chernobyl and Fukushima nuclear power plants, CSN launched special surveillance devices to monitor radioactive contamination through the values provided by the various networks and environmental monitoring programs. The aim of this study is to compare and analyze the results corresponding to the exposure pathways and matrices in which contamination is detected. (Author)

  9. On the validation of an Eulerian model with the Fukushima Nuclear Power Plant (FNPP) accident. Global and local results for Europe and Japan

    Energy Technology Data Exchange (ETDEWEB)

    Evangeliou, Nikolaos; Balkanski, Yves; Cozic, Anne [CEA-CNRS-UVSQ UMR 8212, IPSL/LSCE - Laboratoire des Sciences du Climat et de l' Environnement, L' Orme des Merisiers, 91191 Gif-sur-Yvette Cedex (France); Florou, Heleni; Eleftheriadis, Konstantinos; Kritidis, Panayotis [NCSR ' Demokritos' , Institute of Nuclear and Radiological Sciences and Technology, Energy and Safety (INRASTES), Environmental Radioactivity Laboratory, 15310 Athens (Greece)

    2014-07-01

    A large debate about the exact emissions after the Fukushima NPP accident is still ongoing more than 3 years after the original disaster. Terada et al. (2012) reported the total release of {sup 137}Cs to be 13 PBq (x10{sup 15} Bq), based on an inverse modelling using Japanese data only, whereas the IRSN reported releases of {sup 137}Cs to be 20.6 PBq (IRSN, 2011). In the present study, we used the emission inventories for {sup 137}Cs and {sup 133}Xe reported by Stohl et al. (2012) estimated by inverse modelling using the CTBTO (Comprehensive Nuclear Test Ban Treaty Organisation) and Japanese networks (36.7 PBq of {sup 137}Cs and 15.3 EBq of {sup 133}Xe). For the simulations of the accident, three different versions of the LMDZORINCA model were used; A regular one with a grid resolution of 2.50 deg. x 1.27 deg. for the global comparison with the CTBTO network (19 and 39 vertical layers), and a zoom version over Europe and Asia (0.45 deg. x 0.51 deg. for 19 levels) resulting after 'stretching' the grid using the same number of grid points to assess what happened in Greece, and Japan. Cesium isotopes were treated as sub-micronic aerosols, whereas {sup 133}Xe as a passive tracer within the model, whereas several other radionuclides were estimated from reported isotopic ratios. Our results for the global assessment fit well to the observations. They differ about 0.04% from the measurements for {sup 137}Cs, and around 40% for xenon. The most significant deviations were observed for the northernmost stations due to both scavenging processes and transport over the Arctic. Scattered measurements of several radionuclides from Japan were adopted from literature (Shininaga et al., 2014). The comparison showed a significant quality of our model, although some isotopes are miscalculated. This shows that the reported isotopic ratios might be biased somehow. Finally, in Greece, few measurements of {sup 131}I, {sup 134}Cs and {sup 137}Cs were adopted from Potiriadis et al

  10. Rupture of the aorta following road traffic accidents in the United Kingdom 1992-1999. The results of the co-operative crash injury study.

    Science.gov (United States)

    Richens, D; Kotidis, K; Neale, M; Oakley, C; Fails, A

    2003-02-01

    The true incidence and survivability of blunt traumatic aortic rupture following road traffic accidents in the UK is unclear. The objective of this study was to determine the extent of blunt traumatic aortic rupture in the UK after road traffic accidents and the conditions under which it occurs. Data for the study was obtained from the Co-operative Crash Injury Study database. Road traffic accidents that happened between 1992 and 1999 and included in the Co-operative Crash Injury Study database were retrospectively investigated. A total of 8285 vehicles carrying 14,435 occupants were involved in 7067 accidents. There were 132 cases of blunt traumatic aortic rupture, of which the scene survival was 9% and the overall mortality was 98%. Twenty-one percent of all fatalities had blunt traumatic aortic rupture (130/613). Twenty-nine percent were due to frontal impacts and 44% were due to side impacts. Twelve percent of the blunt traumatic aortic rupture cases in frontal vehicle impacts were wearing seat belts and had airbag protection and 19% had no restraint mechanism. The Equivalent Test Speed of the accident vehicles, (where equivalent test speed provides an estimate of the vehicle impact severity and not an estimate of the vehicle speed at the time of the accident), ranged from 30 to 110 km/h in frontal impacts and from 15 to 82 km/h in side impacts. Blunt traumatic aortic rupture carries a high mortality and occurred in 21% of car occupant deaths in this sample of road traffic accidents. Impact scenarios varied but were most common from the side. The use of an airbag or seat belt does not eliminate risk. The injury can occur at low severity impacts particularly in side impact. Copyright 2002 Elsevier Science B.V.

  11. Summary of intensive monitoring for radionuclides in fishery products after Fukushima accident and comparison to the results of long term monitoring program in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Ken; Morita, Takami; Shigenobu, Yuya; Takagi, Kaori; Miki, Shizuho; Kaeriyama, Hideki; Ambe, Daisuke; Ono, Tsuneo [National Research Institute of Fisheries Science, Fisheries Research Agency, 2-12-4, Fukuura, Kanazawa, Yokohama, Kanagawa 236-8648 (Japan)

    2014-07-01

    Monitoring of artificial radionuclides in fishery products started in late 1950's and has over half a century history in Japan. Fisheries Agency (FA), Fisheries Research Agency (FRA) and prefectural organizations have been conducting the monitoring. The intensive monitoring of radioactive material in fishery products started in the late of March 2011 immediately after the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) accident for the coastal area faced to the Pacific Ocean in the eastern area of Japan and for the offshore area. The purpose of this monitoring is to avoid the distribution of fishery products containing radioactive cesium in concentration over the Japanese standard limit (Cs-134+Cs-137: 100 Bq/kg). Japanese monitoring data are opened to public on the FA's and FRA's web page. In this study, we resume the results of intensive monitoring started in 2011 and show the temporal change of concentration of radioactive cesium in fishery product. In Fukushima Prefecture, the ratio of samples exceeding the Japanese standard limit shows a steadily decreasing trend from 53% at Mar.- Jun. 2011 to below 3% at Jul. - Sep. 2013. In other prefectures, the ratio was 6.5% at Mar.- Jun. 2011 and fell gradually, and it has been below to 1% at Jul. - Sep. 2013. In presentation, we show the results of monitoring conducted in the coastal area and deeper water around Japan from long gamma ray measurements with ashed samples by high purified germanium gamma spectrometry. (authors)

  12. In-vessel Zircaloy oxidation/hydrogen generation behavior during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Cronenberg, A.W. (Science and Engineering Associates, Inc., Albuquerque, NM (USA))

    1990-09-01

    In-vessel Zircaloy oxidation and hydrogen generation data from various US Nuclear Regulatory Commission severe-fuel damage test programs are presented and compared, where the effects of Zircaloy melting, bundle reconfiguration, and bundle quenching by reflooding are assessed for common findings. The experiments evaluated include fuel bundles incorporating fresh and previously irradiated fuel rods, as well as control rods. Findings indicate that the extent of bundle oxidation is largely controlled by steam supply conditions and that high rates of hydrogen generation continued after melt formation and relocation. Likewise, no retardation of hydrogen generation was noted for experiments which incorporated control rods. Metallographic findings indicate extensive oxidation of once-molten Zircaloy bearing test debris. Such test results indicate no apparent limitations to Zircaloy oxidation for fuel bundles subjected to severe-accident coolant-boiloff conditions. 46 refs., 22 figs., 12 tabs.

  13. Transient two-phase performance of LOFT reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed.

  14. Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

    Directory of Open Access Journals (Sweden)

    Avinash J. Gaikwad

    2008-01-01

    Full Text Available Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs, like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs is resorted to mitigate consequences of station blackout (SBO. In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT, SGs, and PDHRs under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections. On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

  15. Optimized Coolant-Flow Diverter For Increased Bearing Life

    Science.gov (United States)

    Subbaraman, Maria R.; Butner, Myles F.

    1995-01-01

    Coolant-flow diverter for rolling-element bearings in cryogenic turbopump designed to enhance cooling power of flow in contact with bearings and thereby reduce bearing wear. Delivers jets of coolant as close as possible to hot spots at points of contact between balls and race. Also imparts swirl that enhances beneficial pumping effect. Used with success in end ball bearing of high-pressure-oxidizer turbopump.

  16. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States). Dept. of Mechanical Engineering

    1995-12-31

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines; however there is practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  17. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    Science.gov (United States)

    2006-12-01

    1992) PFR UK 250 MWe - 14 Shut Down (1994) Rapsodie France 40 MWe - 40 Shut Down (1983) Phenix France 233 MWe - 22 In Operation BOR-60 Russia...107years.98 • Problems with radioactive waste management and coolant disposal during decommissioning .99 O th er • Lead is abundantly available in...is high due to Bi-210, half-life 3.6 106years.102 • Problems with radioactive waste management and coolant disposal during decommissioning . 103 O

  18. Evaluation of In-Vessel Corium Retention under a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae-Joon; Kang, Kyung-Ho; Ha, Kwang-Soon; Kim, Jong-Tae; Koo, Kil-Mo; Cho, Young-Ro; Hong, Seong-Wan; Kim, Sang-Baik; Kim, Hee-Dong

    2008-02-15

    The current study on In-Vessel corium Retention and its application activities to the actual nuclear power plant have been reviewed and discussed in this study. Severe accident sequence which determines an initial condition of the IVR has been evaluated and late phase melt progression, heat transfer on the outer reactor vessel, and in-vessel corium cooling mechanism have been estimated in detail. During the high pressure sequence of the reactor coolant system, a natural circulation flow of the hot steam leads to a failure of the pressurizer surge line before the reactor vessel failure, which leads to a rapid decrease of the reactor coolant system pressure. The results of RASPLAV/MASCA study by OECD/NEA have shown that a melt stratification has occurred in the lower plenum of the reactor vessel. In particular, laver inversion has occurred, which is that a high density of the metal melt moves to the lower part of the oxidic melt layer. A method of heat transfer enhancement on the outer reactor vessel is an optimal design of the reactor vessel insulation for an increase of the natural circulation flow between the outer reactor vessel and the its insulation, and an increase of the critical Heat flux on the outer reactor vessel by using various method, such as Nono fluid, coated reactor vessel, and so on. An increase method of the in-vessel melt cooling is a development of the In-vessel core catcher and a decrease of focusing effect in the metal layer.

  19. The dominance of accidents caused by banalities

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten

    . Nevertheless, the fact is that the simpler accidents normally caused by what might be regarded as banalities occur at a much higher frequencies and with many more fatalities and invalidities than any of what are usually regarded as the most dangerous kinds of accidents. In depth analysis of national statistics...... on accidents could reveal the kind of accidents we are talking about, where they happen, to whom, how, and what can be done about them. This would require a special registration system of the events leading up to the accident. The main results for the four most frequent types of accident will be described...... as an example of how much information such systems can offer in general for the work of accident prevention in more traditional and common enterprises....

  20. Thyroid exposure of Belarusian and Ukrainian children due to the Chernobyl accident and resulting thyroid cancer risk. Final report of BfS project StSch 4240

    Energy Technology Data Exchange (ETDEWEB)

    Jacob, P.; Meckbach, R.; Ulanovski, A.; Schotola, C.; Proehl, G. [GSF-Institute of Radiation Protection, Neuherberg (Germany); Kenigsberg, J.; Buglova, E.; Kruk, J. [Institute of Radiation Medicine and Endocrinology, Minsk (Belarus); Likhtarev, I.; Kovgan, L.; Vavilov, S.; Chepurniy, M. [Ukrainian Radiation Protection Inst., Kyiv (Ukraine); Tronko, M.; Bogdanova, T. [Institute of Endocrinolgoy and Metabolism of the Academy of Medical Sciences of Ukraine, Kyiv (Ukraine); Shinkarev, S.; Gavrilin, Y. [All-Russian Public Organization of Invalids ' Chernobylets' , Scientific Center ' FENIX' , Moscow (Russian Federation); Demidchik, Y. [Thyroid Cancer Center, Minsk (Belarus)

    2005-07-01

    Main objectives of the BfS Project StSch4240 Thyroid Exposure of Belarusian and Ukrainian Children due to the Chernobyl Accident and Resulting Thyroid Cancer Risk were: to establish improved estimates of average thyroid dose for both genders and for each birth-year cohort of the period 1968 - 1985 in Ukrainian and Belarusian settlements, in which more than 10 measurements of the {sup 131}I activity in the human thyroid have been performed in May/June 1986, to explore, whether this dosimetric database can be extended to neighboring settlements, to establish improved estimates of average thyroid dose for both genders and for each birth-year cohort of the period 1968 - 1985 in Ukrainian and Belarusian oblasts (regions) and larger cities, to document the thyroid cancer incidence for the period 1986 - 2001 in Ukraine and Belarus and describe morphological characteristics of the cancer cases, to assess the contribution of the baseline incidence to the total thyroid cancer incidence in the two countries and identify regional and temporal dependencies, to perform analyses of excess risks in settlements with more than 10 measurements of the {sup 131}I activity in the human thyroid. The project has been conducted in the period 6 December 1999 to 31 March 2004. (orig.)

  1. RENEB accident simulation exercise

    OpenAIRE

    Brzozowska, Beata; Ainsbury, Elizabeth; Baert, Annelot; Beaton-Green, Lindsay; Barrios, Leonardo; Barquinero, Joan Francesc; Bassinet, Celine; Beinke, Christina; Benedek, Anett; Beukes, Philip; Bortolin, Emanuela; Buraczewska, Iwona; Burbidge, Christopher; De Amicis, Andrea; De Angelis, Cinzia

    2017-01-01

    Purpose: The RENEB accident exercise was carried out in order to train the RENEB participants in coordinating and managing potentially large data sets that would be generated in case of a major radiological event. Materials and methods: Each participant was offered the possibility to activate the network by sending an alerting email about a simulated radiation emergency. The same participant had to collect, compile and report capacity, triage categorization and exposure scenario results ob...

  2. Test Data for USEPR Severe Accident Code Validation

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe

    2007-05-01

    This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: • Fuel Heatup and Melt Progression • Reactor Coolant System (RCS) Thermal Hydraulics • In-Vessel Molten Pool Formation and Heat Transfer • Fuel/Coolant Interactions during Relocation • Debris Heat Loads to the Vessel • Vessel Failure • Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure • Melt Spreading and Coolability • Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

  3. Modeling vertical loads in pools resulting from fluid injection. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lai, W.; McCauley, E.W.

    1978-06-15

    Table-top model experiments were performed to investigate pressure suppression pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peachbottom Mark I boiling water reactor containment system. The results guided subsequent conduct of experiments in the /sup 1///sub 5/-scale facility and provided new insight into the vertical load function (VLF). Model experiments show an oscillatory VLF with the download typically double-spiked followed by a more gradual sinusoidal upload. The load function contains a high frequency oscillation superimposed on a low frequency one; evidence from measurements indicates that the oscillations are initiated by fluid dynamics phenomena.

  4. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)

    2007-07-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.

  5. Development of core design and analysis technology for integral reactor; development of coolant activity and dose evaluation program

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Sun; Kim, Byeong Soo; Go, Hyun Seok; Lee, Young Wook; Jang, Mee [Seoul National University, Seoul (Korea)

    2002-03-01

    SMART, small- medium-sized integral reactor, is different from the customary electricity-generation PWR in design concepts and structures. The conventional coolant activity evaluation codes used in customary PWRs cannot be applied to SMART. In this study, SAEP(Specific Activity Evaluation Program) is developed that can be applied to both customary PWR and advanced reactor such as SMART. SAEP uses three methods(SAEP Ver.02, Ver.05, Ver.06) to evaluate coolant activity. They solve inhomogeneous linearly-coupled differential equations generated by considering nuclear system as N sub-components. Coolant activities of customary PWR are evaluated by use of SAEP. The results show good agreement with FSAR data. SAEP is used to evaluate coolant activities for SMART and the results are proposed in this study. These results show that SAEP is able to perform coolant activity evaluation for both customary PWR and advanced reactor such as SMART. In addition, with respect to radiation shielding optimization, conventional optimization methods and their characteristics related to radiation shielding are reviewed and analyzed. Strategies for proper usage of conventional methods are proposed to agree with the shielding design cases. 30 refs., 25 figs., 6 tabs. (Author)

  6. Development of core design and analysis technology for integral reactor; development of coolant activity and dose evaluation program

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Sun; Kim, Byeong Soo; Go, Hyun Seok; Lee, Young Wook; Jang, Mee [Seoul National University, Seoul (Korea)

    2002-03-01

    SMART, small- medium-sized integral reactor, is different from the customary electricity-generation PWR in design concepts and structures. The conventional coolant activity evaluation codes used in customary PWRs cannot be applied to SMART. In this study, SAEP(Specific Activity Evaluation Program) is developed that can be applied to both customary PWR and advanced reactor such as SMART. SAEP uses three methods(SAEP Ver.02, Ver.05, Ver.06) to evaluate coolant activity. They solve inhomogeneous linearly-coupled differential equations generated by considering nuclear system as N sub-components. Coolant activities of customary PWR are evaluated by use of SAEP. The results show good agreement with FSAR data. SAEP is used to evaluate coolant activities for SMART and the results are proposed in this study. These results show that SAEP is able to perform coolant activity evaluation for both customary PWR and advanced reactor such as SMART. In addition, with respect to radiation shielding optimization, conventional optimization methods and their characteristics related to radiation shielding are reviewed and analyzed. Strategies for proper usage of conventional methods are proposed to agree with the shielding design cases. 30 refs., 25 figs., 6 tabs. (Author)

  7. Experimental studies into the fluid dynamic performance of the coolant flow in the mixed core of the Temelin NPP VVER-1000 reactor

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-11-01

    Full Text Available The paper presents the results of studies into the interassembly coolant interaction in the Temelin nuclear power plant (NPP VVER-1000 reactor core. An aerodynamic test bench was used to study the coolant flow processes in a TVSA-type fuel assembly bundle. To obtain more detailed information on the coolant flow dynamics, a VVER-1000 reactor core fragment was selected as the test model, which comprised two segments of a TVSA-12 PLUS fuel assembly and one segment of a TVSA-T assembly with stiffening angles and an interassembly gap. The studies into the coolant fluid dynamics consisted in measuring the velocity vector both in representative TVSA regions and inside the interassembly gap using a five-channel pneumometric probe. An analysis into the spatial distribution of the absolute flow velocity projections made it possible to detail the TVSA spacer, mixing and combined spacer grid flow pattern, identify the regions with the maximum transverse coolant flow, and determine the depth of the coolant flow disturbance propagation and redistribution in adjacent TVSA assemblies. The results of the studies into the interassembly coolant interaction among the adjacent TVSA assemblies are used at OKBM Afrikantov to update the VVER-1000 core thermal-hydraulic analysis procedures and have been added to the database for verification of computational fluid dynamics (CFD codes and for detailed cellwise analyses of the VVER-100 reactor cores.

  8. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    Science.gov (United States)

    Reyes, S.; Latkowski, J. F.; Gomez del Rio, J.; Sanz, J.

    2001-05-01

    Previous studies of the safety and environmental aspects of the HYLIFE-II inertial fusion energy power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work, computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) have been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here we consider a severe loss of coolant accident (LOCA) in conjunction with simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the confinement) and of the two barriers surrounding the chamber (inner shielding and confinement building itself). Even though confinement failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product transport and release. The results of these calculations show that the estimated off-site dose is less than 5 mSv (0.5 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  9. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, S; Gomez del Rio, J; Sanz, J

    2000-02-23

    Previous studies of the safety and environmental (S and E) aspects of the HYLIFE-II inertial fusion energy (IFE) power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work a set of computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) has been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here the authors consider a severe lost of coolant accident (LOCA) producing simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the containment) and of the two barriers surrounding the chamber (inner shielding and containment building it self). Even though containment failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product release and transport. The results of these calculations show that the estimated off-site dose is less than 6 mSv (0.6 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  10. Experimental simulation of asymmetric heat up of coolant channel under small break LOCA condition for PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Yadav, Ashwini K., E-mail: ashwinikumaryadav@gmail.com [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kumar, Ravi, E-mail: ravikfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Chatterjee, B., E-mail: barun@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Gupta, Akhilesh, E-mail: akhilfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Mukhopadhyay, D., E-mail: dmukho@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2013-02-15

    Highlights: ► Circumferential temperature gradient of PT for asymmetric heat-up was 440 °C. ► At 2 MPa ballooning initiated at 450 °C and with strain rate of 0.0277%/s. ► At 4 MPa ballooning initiated at 390 °C and with strain rate of 0.0305%/s. ► At 4 MPa, PT ruptured under uneven strain and steep temperature gradient. ► Integrity of PT depends on internal pressure and magnitude of decay power. -- Abstract: During postulated small break loss of coolant accident (SBLOCA) for Pressurised Heavy Water Reactors (PHWRs) as well as for postulated SBLOCA coincident with loss of ECCS, a stratified flow condition can arise in the coolant channels as the gravitational force dominates over the low inertial flow arising from small break flow. A Station Blackout condition without operator intervention can also lead to stratified flow condition during a slow channel boil-off condition. For all these conditions the pressure remains high and under stratified flow condition, the horizontal fuel bundles experience different heat transfer environments with respect to the stratified flow level. This causes the bundle upper portion to get heated up higher as compared to the submerged portion. This kind of asymmetrical heating of the bundle is having a direct bearing on the circumferential temperature gradient of pressure tube (PT) component of the coolant channel. The integrity of the PT is important under normal conditions as well as at different accident loading conditions as this component houses the fuel bundles and serves as a coolant pressure boundary of the reactors. An assessment of PT is required with respect to different accident loading conditions. The present investigation aims to study thermo-mechanical behaviour of PT (Zr, 2.5 wt% Nb) under a stratified flow condition under different internal pressures. The component is subjected to an asymmetrical heat-up conditions as expected during the said situation under different pressure conditions which varies from 2

  11. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Summary of results. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Whitehead, D.W. [ed.; Staple, B.D.; Daniel, S.L. [and others

    1995-07-01

    During 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to examine the potential risks during low power and shutdown operations. Two plants, Surry and Grand Gulf, were selected as the plants to be studied by Brookhaven National Laboratory (Surry) and Sandia National Laboratories (Grand Gulf). This report documents the work performed during the analysis of the Grand Gulf plant. A phased approach was used for the overall study. In Phase 1, the objectives were to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenario frequencies and risks, and to provide a foundation for a detailed Phase 2 analysis. It was in Phase 1 that the concept of plant operational states (POSs) was developed to allow the analysts to better represent the plant as it transitions from power operation to nonpower operation than was possible with the traditional technical specification divisions of modes of operation. This phase consisted of a coarse screening analysis performed for all POSs, including seismic and internal fire and flood for some POSs. In Phase 2, POS 5 (approximately cold shutdown as defined by Grand Gulf Technical Specifications) during a refueling outage was selected as the plant configuration to be analyzed based on the results of the Phase 1 study. The scope of the Level 1 study includes plant damage state analysis and uncertainty analysis and is documented in a multi-volume NUREG/CR report (i.e., NUREG/CR-6143). The internal events analysis is documented in Volume 2. Internal fire and internal flood analyses are documented in Volumes 3 and 4, respectively. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. The Level 2/3 study of the traditional internal events is documented in Volume 6, and a summary of the results for all analyses is documented in Volume 1.

  12. PREVENTION OF OCCUPATIONAL ACCIDENTS

    Directory of Open Access Journals (Sweden)

    Jovica Jovanovic

    2004-01-01

    Full Text Available Medical services, physicians and nurses play an essential role in the plant safety program through primary treatment of injured workers and by helping to identify workplace hazards. The physician and nurse should participate in the worksite investigations to identify specific hazard or stresses potentially causing the occupational accidents and injuries and in planning the subsequent hazard control program. Physicians and nurses must work closely and cooperatively with supervisors to ensure the prompt reporting and treatment of all work related health and safety problems. Occupational accidents, work related injuries and fatalities result from multiple causes, affect different segments of the working population, and occur in a myriad of occupations and industrial settings. Multiple factors and risks contribute to traumatic injuries, such as hazardous exposures, workplace and process design, work organization and environment, economics, and other social factors. With such a diversity of theories, it will not be difficult to understand that there does not exist one single theory that is considered right or correct and is universally accepted. These theories are nonetheless necessary, but not sufficient, for developing a frame of reference for understanding accident occurrences. Prevention strategies are also varied, and multiple strategies may be applicable to many settings, including engineering controls, protective equipment and technologies, management commitment to and investment in safety, regulatory controls, and education and training. Research needs are thus broad, and the development and application of interventions involve many disciplines and organizations.

  13. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  14. 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)

  15. 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.

  16. Optimization of the Severe Accident Management Strategy for Domestic Plants and Validation Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. B.; Kim, H. D.; Koo, K. M.; Park, R. J.; Hong, S. H.; Cho, Y. R.; Kim, J. T.; Ha, K. S.; Kang, K. H

    2007-04-15

    nuclear power plants, a technical basis report and computational aid tools were developed in parallel with the experimental and analytical works for the resolution of the uncertain safety issues. ELIAS experiments were carried out to quantify the boiling heat removal rate at the upper surface of a metallic layer for precise evaluations on the effect of a late in-vessel coolant injection. T-HERMES experiments were performed to examine the two-phase natural circulation phenomena through the gap between the reactor vessel and the insulator in the APR1400. Detailed analyses on the hydrogen control in the APR1400 containment were performed focused on the effect of spray system actuation on the hydrogen burning and the evaluation of the hydrogen behavior in the IRWST. To develop the technical basis report for the severe accident management, analyses using SCDAP/RELAP5 code were performed for the accident sequences of the OPR1000. Based on the experimental and analytical results performed in this study, the computational aids for the evaluations of hydrogen flammability in the containment, criteria of the in-vessel corium cooling, criteria of the external reactor vessel cooling were developed. An ASSA code was developed to validate the signal from the instrumentations during the severe accidents and to process the abnormal signal. Since ASSA can perform the signal processing from the direct input of the nuclear power plant during the severe accident, it can be platform of the computational aids. In this study, the ASSA was linked with the computaional aids for the hydrogen flammability.

  17. The premixing and propagation phases of fuel-coolant interactions: a review of recent experimental studies and code developments

    Energy Technology Data Exchange (ETDEWEB)

    Antariksawan, A.R. [Reactor Safety Technology Research Center of BATAN (Indonesia); Moriyama, Kiyofumi; Park, Hyun-sun; Maruyama, Yu; Yang, Yanhua; Sugimoto, Jun

    1998-09-01

    A vapor explosion (or an energetic fuel-coolant interactions, FCIs) is a process in which hot liquid (fuel) transfers its internal energy to colder, more volatile liquid (coolant); thus the coolant vaporizes at high pressure and expands and does works on its surroundings. Traditionally, the energetic fuel-coolant interactions could be distinguished in subsequent stages: premixing (or coarse mixing), triggering, propagation and expansion. Realizing that better and realistic prediction of fuel-coolant interaction consequences will be available understanding the phenomenology in the premixing and propagation stages, many experimental and analytical studies have been performed during more than two decades. A lot of important achievements are obtained during the time. However, some fundamental aspects are still not clear enough; thus the works are directed to that direction. In conjunction, the model/code development is pursuit. This is aimed to provide a scaling tool to bridge the experimental results to the real geometries, e.g. reactor pressure vessel, reactor containment. The present review intends to collect the available information on the recent works performed to study the premixing and propagation phases. (author). 97 refs.

  18. Shipboard determination of radiocesium in seawater after the Fukushima accident: results from the 2011-2012 Russian expeditions to the Sea of Japan and western North Pacific Ocean.

    Science.gov (United States)

    Ramzaev, V; Nikitin, A; Sevastyanov, A; Artemiev, G; Bruk, G; Ivanov, S

    2014-09-01

    waters had decreased for all seven stations repeatedly sampled in the study. A detailed observation of radiocesium distribution within the water column down to the depth of 200 m at nine stations from the Kuroshio-Oyashio Interfrontal Zone and Kuroshio Extension in 2012 revealed maximal activity concentrations of both cesium radionuclides in the 100-200 m depth layer. The average inventory of Fukushima-derived (137)Cs in the top 200 m of the water column for the nine stations was estimated as 1.19 kBq m(-2) (decay corrected to 15 March 2011) which is 4.6 times higher than the background value of 0.26 kBq m(-2) expected for this depth. The monitoring results obtained in the study and relevant data published by others show that following the Fukushima accident, the Oyashio current acts as a provider of low-contaminated subarctic waters to the heavily contaminated Kuroshio-Oyashio mixed water region.

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

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, In Seop, E-mail: inseopjeon@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Heo, Sun, E-mail: sunnysunny@khnp.co.kr [Central Research Institute, Korea Hydro & Nuclear Power Co., 70 Yuseong-daero 1312 beon-gil, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Kang, Hyun Gook, E-mail: hyungook@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2016-02-15

    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.

  20. Temporal Statistic of Traffic Accidents in Turkey

    Science.gov (United States)

    Erdogan, S.; Yalcin, M.; Yilmaz, M.; Korkmaz Takim, A.

    2015-10-01

    Traffic accidents form clusters in terms of geographic space and over time which themselves exhibit distinct spatial and temporal patterns. There is an imperative need to understand how, where and when traffic accidents occur in order to develop appropriate accident reduction strategies. An improved understanding of the location, time and reasons for traffic accidents makes a significant contribution to preventing them. Traffic accident occurrences have been extensively studied from different spatial and temporal points of view using a variety of methodological approaches. In literature, less research has been dedicated to the temporal patterns of traffic accidents. In this paper, the numbers of traffic accidents are normalized according to the traffic volume and the distribution and fluctuation of these accidents is examined in terms of Islamic time intervals. The daily activities and worship of Muslims are arranged according to these time intervals that are spaced fairly throughout the day according to the position of the sun. The Islamic time intervals are never been used before to identify the critical hour for traffic accidents in the world. The results show that the sunrise is the critical time that acts as a threshold in the rate of traffic accidents throughout Turkey in Islamic time intervals.

  1. TEMPORAL STATISTIC OF TRAFFIC ACCIDENTS IN TURKEY

    Directory of Open Access Journals (Sweden)

    S. Erdogan

    2015-10-01

    Full Text Available Traffic accidents form clusters in terms of geographic space and over time which themselves exhibit distinct spatial and temporal patterns. There is an imperative need to understand how, where and when traffic accidents occur in order to develop appropriate accident reduction strategies. An improved understanding of the location, time and reasons for traffic accidents makes a significant contribution to preventing them. Traffic accident occurrences have been extensively studied from different spatial and temporal points of view using a variety of methodological approaches. In literature, less research has been dedicated to the temporal patterns of traffic accidents. In this paper, the numbers of traffic accidents are normalized according to the traffic volume and the distribution and fluctuation of these accidents is examined in terms of Islamic time intervals. The daily activities and worship of Muslims are arranged according to these time intervals that are spaced fairly throughout the day according to the position of the sun. The Islamic time intervals are never been used before to identify the critical hour for traffic accidents in the world. The results show that the sunrise is the critical time that acts as a threshold in the rate of traffic accidents throughout Turkey in Islamic time intervals.

  2. Radiological accidents balance in medicine; Bilan des accidents radiologiques en medecine

    Energy Technology Data Exchange (ETDEWEB)

    Nenot, J.C.

    1995-12-31

    This work deals with the radiological accidents in medicine. In medicine, the radiation accidents on medical personnel and patients can be the result of over dosage and bad focusing of radiotherapy sealed sources. Sometimes, the accidents, if they are unknown during a time enough for the source to be spread and to expose a lot of persons (in the case of source dismantling for instance) can take considerable dimensions. Others accidents can come from bad handling of linear accelerators and from radionuclide kinetics in some therapies. Some examples of accidents are given. (O.L.). 11 refs.

  3. Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)

    Energy Technology Data Exchange (ETDEWEB)

    Larson, J.R.; Evans, D.R.; McCardell, R.K.

    1978-01-01

    This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR.

  4. Measurement of long-lived radionuclides in surface soil around F1NPP accident site by Accelerator Mass Spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Miyake, Yasuto; Matsuzaki, Hiroyuki [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo (Japan); Sasa, Kimikazu; Takahashi, Tsutomu [AMS Group, Tandem Accelerator Complex, Research Facility Center for Science and Technology, University of Tsukuba, Tsukuba, Ibaraki 305-8577 (Japan)

    2015-10-15

    In March 2011, vast amounts of radionuclides were released into the environment due to the Fukushima Daiichi Nuclear Power Plant (F1NPP) accident. However, very little work has been done concerning accident-derived long-lived nuclides such as {sup 129}I (T{sub 1/2} = 1.57 × 10{sup 7} year) and {sup 36}Cl (T{sub 1/2} = 3.01 × 10{sup 5} year). {sup 129}I and {sup 131}I are both produced by {sup 235}U fission in nuclear reactors. Being isotopes of iodine, these nuclides are expected to behave similarly in the environment. This makes {sup 129}I useful for retrospective reconstruction of {sup 131}I distribution during the initial stages of the accident. On the other hand, {sup 36}Cl is generated during reactor operation via neutron capture reaction of {sup 35}Cl, an impurity in the coolant or reactor component. Resulting {sup 36}Cl/Cl ratio within the reactor is thus much higher compared to that in environment. Similar to {sup 129}I, {sup 36}Cl is expected to have leaked out during the accident and it is important to evaluate its effects. In this study, {sup 129}I concentrations were determined in several surface soil samples collected around F1NPP. Average {sup 129}I/{sup 131}I ratio was estimated to be 26.1 ± 5.8 as of March 11, 2011, consistent with calculations using ORIGEN2 code and other published data. {sup 36}Cl/Cl ratios in some of the soil samples were likewise measured and ranged from 1.1 × 10{sup −12} to 2.6 × 10{sup −11}. These are higher compared to ratios measured around F1NPP before the accident. A positive correlation between {sup 36}Cl and {sup 129}I concentration was observed.

  5. Measurement of long-lived radionuclides in surface soil around F1NPP accident site by Accelerator Mass Spectrometry

    Science.gov (United States)

    Miyake, Yasuto; Matsuzaki, Hiroyuki; Sasa, Kimikazu; Takahashi, Tsutomu

    2015-10-01

    In March 2011, vast amounts of radionuclides were released into the environment due to the Fukushima Daiichi Nuclear Power Plant (F1NPP) accident. However, very little work has been done concerning accident-derived long-lived nuclides such as 129I (T1/2 = 1.57 × 107 year) and 36Cl (T1/2 = 3.01 × 105 year). 129I and 131I are both produced by 235U fission in nuclear reactors. Being isotopes of iodine, these nuclides are expected to behave similarly in the environment. This makes 129I useful for retrospective reconstruction of 131I distribution during the initial stages of the accident. On the other hand, 36Cl is generated during reactor operation via neutron capture reaction of 35Cl, an impurity in the coolant or reactor component. Resulting 36Cl/Cl ratio within the reactor is thus much higher compared to that in environment. Similar to 129I, 36Cl is expected to have leaked out during the accident and it is important to evaluate its effects. In this study, 129I concentrations were determined in several surface soil samples collected around F1NPP. Average 129I/131I ratio was estimated to be 26.1 ± 5.8 as of March 11, 2011, consistent with calculations using ORIGEN2 code and other published data. 36Cl/Cl ratios in some of the soil samples were likewise measured and ranged from 1.1 × 10-12 to 2.6 × 10-11. These are higher compared to ratios measured around F1NPP before the accident. A positive correlation between 36Cl and 129I concentration was observed.

  6. Why Airplanes Crash: Causes of Accidents Worldwide

    OpenAIRE

    Oster, Clinton V.; John S. Strong; Zorn, Kurt

    2010-01-01

    In this paper, we report the results of a detailed examination of the causes of 700 fatal aviation accidents that occurred worldwide between 1990 and 2006 in commercial passenger service. We look at both scheduled and nonscheduled and both domestic and international service. We also categorize the accident aircraft as large jets, regional and medium jets, small jets, turboprops, and piston powered aircraft. We find that the mix of causes of those accidents vary substantially across regions of...

  7. Key Characteristics of Combined Accident including TLOFW accident for PSA Modeling

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

    The conventional PSA techniques cannot adequately evaluate all events. The conventional PSA models usually focus on single internal events such as DBAs, the external hazards such as fire, seismic. However, the Fukushima accident of Japan in 2011 reveals that very rare event is necessary to be considered in the PSA model to prevent the radioactive release to environment caused by poor treatment based on lack of the information, and to improve the emergency operation procedure. Especially, the results from PSA can be used to decision making for regulators. Moreover, designers can consider the weakness of plant safety based on the quantified results and understand accident sequence based on human actions and system availability. This study is for PSA modeling of combined accidents including total loss of feedwater (TLOFW) accident. The TLOFW accident is a representative accident involving the failure of cooling through secondary side. If the amount of heat transfer is not enough due to the failure of secondary side, the heat will be accumulated to the primary side by continuous core decay heat. Transients with loss of feedwater include total loss of feedwater accident, loss of condenser vacuum accident, and closure of all MSIVs. When residual heat removal by the secondary side is terminated, the safety injection into the RCS with direct primary depressurization would provide alternative heat removal. This operation is called feed and bleed (F and B) operation. Combined accidents including TLOFW accident are very rare event and partially considered in conventional PSA model. Since the necessity of F and B operation is related to plant conditions, the PSA modeling for combined accidents including TLOFW accident is necessary to identify the design and operational vulnerabilities.The PSA is significant to assess the risk of NPPs, and to identify the design and operational vulnerabilities. Even though the combined accident is very rare event, the consequence of combined

  8. Actively controlling coolant-cooled cold plate configuration

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J.; Parida, Pritish R.

    2015-07-28

    A method is provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The method includes: monitoring a variable associated with at least one of the coolant-cooled cold plate or one or more electronic components being cooled by the cold plate; and dynamically varying, based on the monitored variable, a physical configuration of the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the one or more electronic components, and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the coolant-cooled cold plate, the positioning of which may be adjusted based on the monitored variable.

  9. Fuel cell cooling system using a non-dielectric coolant

    Energy Technology Data Exchange (ETDEWEB)

    Grevstad, P.E.; Gelting, R.L.

    1976-07-13

    A cooler for removing waste heat from a stack of fuel cells uses a non-dielectric coolant which is carried in a plurality of tubes passing through one or more separator plates in the stack. Preferably the coolant is water so that heat removal is by evaporation of the water within the tubes by boiling. The tubes are electrically insulated from the cells by a coating of dielectric material such as polytetrafluoroethylene. In one embodiment of the invention the cooler tubes are connected to the stack coolant supply conduits by dielectric hoses having a high length to diameter ratio to provide a several hundred thousand ohm impedance path in case of a flaw in the protective dielectric coating, in order that a short circuit of the stack does not occur.

  10. Simulation of isothermal multi-phase fuel-coolant interaction using MPS method with GPU acceleration

    Energy Technology Data Exchange (ETDEWEB)

    Gou, W.; Zhang, S.; Zheng, Y. [Zhejiang Univ., Hangzhou (China). Center for Engineering and Scientific Computation

    2016-07-15

    The energetic fuel-coolant interaction (FCI) has been one of the primary safety concerns in nuclear power plants. Graphical processing unit (GPU) implementation of the moving particle semi-implicit (MPS) method is presented and used to simulate the fuel coolant interaction problem. The governing equations are discretized with the particle interaction model of MPS. Detailed implementation on single-GPU is introduced. The three-dimensional broken dam is simulated to verify the developed GPU acceleration MPS method. The proposed GPU acceleration algorithm and developed code are then used to simulate the FCI problem. As a summary of results, the developed GPU-MPS method showed a good agreement with the experimental observation and theoretical prediction.

  11. PIV measurements of coolant flow field in a diesel engine cylinder head

    Science.gov (United States)

    Ma, Hongwei; Zhang, Zhenyang; Xue, Cheng; Huang, Yunlong

    2015-04-01

    This paper presents experimental measurements of coolant flow field in the water jacket of a diesel engine cylinder head. The test was conducted at three different flow rates using a 2-D PIV system. Appropriate tracing particles were selected and delivery device was designed and manufactured before the test. The flow parameters, such as velocity, vorticity and turbulence, were used to analyze the flow field. The effects of vortex which was located between the intake valve and the exhaust valve were discussed. The experimental results showed an asymmetric distribution of velocity in the water jacket. This led to an asymmetric thermal distribution, which would shorten the service life of the cylinder head. The structure optimization to the water jacket of cylinder head was proposed in this paper. The experimental system, especially the 2-D PIV system, is a great help to study the coolant flow structure and analyze cooling mechanism in the diesel engine cylinder head.

  12. The effect of coolants on the performance of magnetic micro-refrigerators.

    Science.gov (United States)

    Silva, D J; Bordalo, B D; Pereira, A M; Ventura, J; Oliveira, J C R E; Araújo, J P

    2014-06-01

    Magnetic refrigeration is an alternative cooling technique with envisaged technological applications on micro- and opto-electronic devices. Here, we present a magnetic micro-refrigerator cooling device with embedded micro-channels and based on the magnetocaloric effect. We studied the influence of the coolant fluid in the refrigeration process by numerically simulating the heat transfer processes using the finite element method. This allowed us to calculate the cooling power of the device. Our results show that gallium is the most efficient coolant fluid and, when used with Gd5Si2Ge2, a maximum power of 11.2 W/mm3 at a working frequency of -5 kHz can be reached. However, for operation frequencies around 50 Hz, water is the most efficient fluid with a cooling power of 0.137 W/mm3.

  13. Quantifying the risk of extreme aviation accidents

    Science.gov (United States)

    Das, Kumer Pial; Dey, Asim Kumer

    2016-12-01

    Air travel is considered a safe means of transportation. But when aviation accidents do occur they often result in fatalities. Fortunately, the most extreme accidents occur rarely. However, 2014 was the deadliest year in the past decade causing 111 plane crashes, and among them worst four crashes cause 298, 239, 162 and 116 deaths. In this study, we want to assess the risk of the catastrophic aviation accidents by studying historical aviation accidents. Applying a generalized Pareto model we predict the maximum fatalities from an aviation accident in future. The fitted model is compared with some of its competitive models. The uncertainty in the inferences are quantified using simulated aviation accident series, generated by bootstrap resampling and Monte Carlo simulations.

  14. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  15. Diagnostic system for identification of accident scenarios in nuclear power plants using artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Santosh, T.V. [Health, Safety and Environment Group, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)], E-mail: santutv@barc.gov.in; Srivastava, A.; Sanyasi Rao, V.V.S.; Ghosh, A.K.; Kushwaha, H.S. [Health, Safety and Environment Group, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2009-03-15

    This paper presents the work carried out towards developing a diagnostic system for the identification of accident scenarios in 220 MWe Indian PHWRs. The objective of this study is to develop a methodology based on artificial neural networks (ANNs), which assists in identifying a transient quickly and suggests the operator to initiate the corrective actions during abnormal operations of the reactor. An operator support system, known as symptom-based diagnostic system (SBDS), has been developed using ANN that diagnoses the transients based on reactor process parameters, and continuously displays the status of the reactor. As a pilot study, the large break loss of coolant accident (LOCA) with and without the emergency core cooling system (ECCS) in reactor headers has been considered. Several break scenarios of large break LOCA have been analyzed. The time-dependent transient data have been generated using the RELAP5 thermal hydraulic code assuming an equilibrium core, which conforms to a realistic estimation. The diagnostic results obtained from the ANN study are satisfactory. These results have been incorporated in the SBDS software for operator assistance. A few important outputs of the SBDS have been discussed in this paper.

  16. Systematics of Reconstructed Process Facility Criticality Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Pruvost, N.L.; McLaughlin, T.P.; Monahan, S.P.

    1999-09-19

    The systematics of the characteristics of twenty-one criticality accidents occurring in nuclear processing facilities of the Russian Federation, the United States, and the United Kingdom are examined. By systematics the authors mean the degree of consistency or agreement between the factual parameters reported for the accidents and the experimentally known conditions for criticality. The twenty-one reported process criticality accidents are not sufficiently well described to justify attempting detailed neutronic modeling. However, results of classic hand calculations confirm the credibility of the reported accident conditions.

  17. Radioactive materials transport accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    McSweeney, T.I.; Maheras, S.J.; Ross, S.B. [Battelle Memorial Inst. (United States)

    2004-07-01

    Over the last 25 years, one of the major issues raised regarding radioactive material transportation has been the risk of severe accidents. While numerous studies have shown that traffic fatalities dominate the risk, modeling the risk of severe accidents has remained one of the most difficult analysis problems. This paper will show how models that were developed for nuclear spent fuel transport accident analysis can be adopted to obtain estimates of release fractions for other types of radioactive material such as vitrified highlevel radioactive waste. The paper will also show how some experimental results from fire experiments involving low level waste packaging can be used in modeling transport accident analysis with this waste form. The results of the analysis enable an analyst to clearly show the differences in the release fractions as a function of accident severity. The paper will also show that by placing the data in a database such as ACCESS trademark, it is possible to obtain risk measures for transporting the waste forms along proposed routes from the generator site to potential final disposal sites.

  18. Nuclear accident dosimetry intercomparison studies.

    Science.gov (United States)

    Sims, C S

    1989-09-01

    Twenty-two nuclear accident dosimetry intercomparison studies utilizing the fast-pulse Health Physics Research Reactor at the Oak Ridge National Laboratory have been conducted since 1965. These studies have provided a total of 62 different organizations a forum for discussion of criticality accident dosimetry, an opportunity to test their neutron and gamma-ray dosimetry systems under a variety of simulated criticality accident conditions, and the experience of comparing results with reference dose values as well as with the measured results obtained by others making measurements under identical conditions. Sixty-nine nuclear accidents (27 with unmoderated neutron energy spectra and 42 with eight different shielded spectra) have been simulated in the studies. Neutron doses were in the 0.2-8.5 Gy range and gamma doses in the 0.1-2.0 Gy range. A total of 2,289 dose measurements (1,311 neutron, 978 gamma) were made during the intercomparisons. The primary methods of neutron dosimetry were activation foils, thermoluminescent dosimeters, and blood sodium activation. The main methods of gamma dose measurement were thermoluminescent dosimeters, radiophotoluminescent glass, and film. About 68% of the neutron measurements met the accuracy guidelines (+/- 25%) and about 52% of the gamma measurements met the accuracy criterion (+/- 20%) for accident dosimetry.

  19. Additional requirements for leak-before-break application to primary coolant piping in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Roussel, G. [AIB Vincotte Nuclear, Brussels (Belgium)

    1997-04-01

    Leak-Before-Break (LBB) technology has not been applied in the first design of the seven Pressurized Water Reactors the Belgian utility is currently operating. The design basis of these plants required to consider the dynamic effects associated with the ruptures to be postulated in the high energy piping. The application of the LBB technology to the existing plants has been recently approved by the Belgian Safety Authorities but with a limitation to the primary coolant loop. LBB analysis has been initiated for the Doel 3 and Tihange 2 plants to allow the withdrawal of some of the reactor coolant pump snubbers at both plants and not reinstall some of the restraints after steam generator replacement at Doel 3. LBB analysis was also found beneficial to demonstrate the acceptability of the primary components and piping to the new conditions resulting from power uprating and stretch-out operation. LBB analysis has been subsequently performed on the primary coolant loop of the Tihange I plant and is currently being performed for the Doel 4 plant. Application of the LBB to the primary coolant loop is based in Belgium on the U.S. Nuclear Regulatory Commission requirements. However the Belgian Safety Authorities required some additional analyses and put some restrictions on the benefits of the LBB analysis to maintain the global safety of the plant at a sufficient level. This paper develops the main steps of the safety evaluation performed by the Belgian Safety Authorities for accepting the application of the LBB technology to existing plants and summarizes the requirements asked for in addition to the U.S. Nuclear Regulatory Commission rules.

  20. Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Shack, W.J. [Argonne National Lab., IL (United States)

    1998-03-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented.

  1. MATLAB/Simulink Framework for Modeling Complex Coolant Flow Configurations of Advanced Automotive Thermal Management Systems

    Energy Technology Data Exchange (ETDEWEB)

    Titov, Gene; Lustbader, Jason; Leighton, Daniel; Kiss, Tibor

    2016-04-05

    The National Renewable Energy Laboratory's (NREL's) CoolSim MATLAB/Simulink modeling framework was extended by including a newly developed coolant loop solution method aimed at reducing the simulation effort for arbitrarily complex thermal management systems. The new approach does not require the user to identify specific coolant loops and their flow. The user only needs to connect the fluid network elements in a manner consistent with the desired schematic. Using the new solution method, a model of NREL's advanced combined coolant loop system for electric vehicles was created that reflected the test system architecture. This system was built using components provided by the MAHLE Group and included both air conditioning and heat pump modes. Validation with test bench data and verification with the previous solution method were performed for 10 operating points spanning a range of ambient temperatures between -2 degrees C and 43 degrees C. The largest root mean square difference between pressure, temperature, energy and mass flow rate data and simulation results was less than 7%.

  2. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  3. Regulation Plans on Severe Accidents developed by KINS Severe Accident Regulation Preparation TFT

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyun Tae; Chung, Ku Young; Na, Han Bee [KINS, Daejeon (Korea, Republic of)

    2016-05-15

    Some nuclear power plants in Fukushima Daiichi site had lost their emergency reactor cooling function for long-time so the fuels inside the reactors were molten, and the integrity of containment was damaged. Therefore, large amount of radioactive material was released to environment. Because the social and economic effects of severe accidents are enormous, Korean Government already issued 'Severe Accident Policy' in 2001 which requires nuclear power plant operators to set up 'Quantitative Safety Goal', to do 'Probabilistic Safety Analysis', to install 'Severe Accident Countermeasures' and to make 'Severe Accident Management Plan'. After the Fukushima disaster, a Special Safety Inspection was performed for all operating nuclear power plants of Korea. The inspection team from industry, academia, and research institutes assessed Korean NPPs capabilities to cope with or respond to severe accidents and emergency situation caused by natural disasters such as a large earthquake or tsunami. As a result of the special inspection, about 50 action items were identified to increase the capability to cope with natural disaster and severe accidents. Nuclear Safety Act has been amended to require NPP operators to submit Accident Management Plant as part of operating license application. The KINS Severe Accident Regulation Preparation TFT had first investigated oversea severe accident regulation trend before and after the Fukushima accident. Then, the TFT has developed regulation draft for severe accidents such as Severe accident Management Plans, the required design features for new NPPs to prevent severe accident against multiple failures and beyond-design external events, countermeasures to mitigate severe accident and to keep the integrity of containment, and assessment methodology on safety assessment plan and probabilistic safety assessment.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  5. Accident history, risk perception and traffic safe behaviour.

    Science.gov (United States)

    Ngueutsa, Robert; Kouabenan, Dongo Rémi

    2017-09-01

    This study clarifies the associations between accident history, perception of the riskiness of road travel and traffic safety behaviours by taking into account the number and severity of accidents experienced. A sample of 525 road users in Cameroon answered a questionnaire comprising items on perception of risk, safe behaviour and personal accident history. Participants who reported involvement in more than three accidents or involvement in a severe accident perceived road travel as less risky and also reported behaving less safely compared with those involved in fewer, or less severe accidents. The results have practical implications for the prevention of traffic accidents. Practitioner Summary: The associations between accident history, perceived risk of road travel and safe behaviour were investigated using self-report questionnaire data. Participants involved in more than three accidents, or in severe accidents, perceived road travel as less risky and also reported more unsafe behaviour compared with those involved in fewer, or less severe accidents. Campaigns targeting people with a less serious, less extensive accident history should aim to increase awareness of hazards and the potential severity of their consequences, as well as emphasising how easy it is to take the recommended preventive actions. Campaigns targeting those involved in more frequent accidents, and survivors of serious accidents, should address feelings of invulnerability and helplessness.

  6. A study on industrial accident rate forecasting and program development of estimated zero accident time in Korea.

    Science.gov (United States)

    Kim, Tae-gu; Kang, Young-sig; Lee, Hyung-won

    2011-01-01

    To begin a zero accident campaign for industry, the first thing is to estimate the industrial accident rate and the zero accident time systematically. This paper considers the social and technical change of the business environment after beginning the zero accident campaign through quantitative time series analysis methods. These methods include sum of squared errors (SSE), regression analysis method (RAM), exponential smoothing method (ESM), double exponential smoothing method (DESM), auto-regressive integrated moving average (ARIMA) model, and the proposed analytic function method (AFM). The program is developed to estimate the accident rate, zero accident time and achievement probability of an efficient industrial environment. In this paper, MFC (Microsoft Foundation Class) software of Visual Studio 2008 was used to develop a zero accident program. The results of this paper will provide major information for industrial accident prevention and be an important part of stimulating the zero accident campaign within all industrial environments.

  7. Validation of GAMMA+ model for Evaluating Heat Transfer of VHTR core in Accident Conditions by CFD analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Dongho; Yoon, Sujong; Park, Gooncherl; Cho, Hyoungkyu [Seoul National Univ., Seoul (Korea, Republic of)

    2013-05-15

    KAERI has established a plan to demonstrate massive production of hydrogen using a VHTR by the early 2020s. In addition the GAMMA+ code is developed to analyze VHTR thermo-fluid transients at KAERI. One of the candidate reactor designs for VHTR is prismatic modular reactor (PMR), of which reference reactor is the 600MWth GT-MHR. This type of reactor has a passive safety system. During the High Pressure Conduction Cooling (HPCC) or Low Pressure Conduction Cooling (LPCC) accident, the core heats up by decay heat and then starts to cool down by conduction and radiation cooling to the Reactor Cavity Cooling System (RCCS) through the prismatic core. In this mechanism, the solid conduction occurs in graphite and fuel blocks, and the gas conduction and radiation occurs in coolant holes and bypass gaps. It is important to predict conduction and radiation heat transfer in the core for safety analysis. Effective thermal conductivity is derived by Maxwell's far-field methodology Radiation effect is expressed as corresponding conductivity and added to gas conductivity. In this study, ETC model used in GAMMA+ code is validated with the commercial CFD code, CFX-13. In this study, the effective thermal conductivity model of the GAMMA+ was evaluated by comparison of CFD analysis. The CFD analysis was conducted for various numbers and volume fractions of coolant holes and temperatures. Although slight disagreement was shown for the cases run with small number of holes, the result of GAMMA+ model is accurate for the large numbers of holes sufficiently. Since there are 102 coolant holes and 210 fuel holes in a fuel block, it is concluded that GAMMA+ model is proper formula for predicting effective thermal conductivity of the VHTR fuel block. However, in high temperature region above 500 .deg. C, the GAMMA+ model underestimates the effective thermal conductivity since radiation heat transfer is not reflected precisely. Further researches on it seem to be necessary.

  8. Accident Page

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    @@ Explosion in Nanjing Dingshan On February 4th, 2007 an isobutylene tank containing 14 tons of isobutylene exploded and caught fire at Nanjing Dingshan Fine Chemical Plant (Nanjing Dingshan) located in Qixia District, Nanjing, Jiangsu province. The explosion had effect on the nearby Nanjing Hongyan Plastic Plant and resulted in one death and six injuries.

  9. AUTOMOTIVE AND HEAVY-DUTY ENGINE COOLANT RECYCLING BY DISTILLATION

    Science.gov (United States)

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants for a facility such as the New Jersey Department of Transportation garage in Ewing, New Jersey. he specific recycling evaluated is b...

  10. EVALUATION OF FILTRATION AND DISTILLATION METHODS FOR RECYCLING AUTOMOTIVE COOLANT.

    Science.gov (United States)

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants at a New Jersey Department of Transportation garage. The specific recycling units evaluated are based on the technologies of filtrat...

  11. Fuels, Lubricants, and Coolants. FOS: Fundamentals of Service.

    Science.gov (United States)

    John Deere Co., Moline, IL.

    This manual on fuels, lubricants, and coolants is one of a series of power mechanics tests and visual aids on automotive and off-the-road agricultural and construction equipment. Materials present basic information with illustrations for use by vocational students and teachers as well as shop servicemen and laymen. Focusing on fuels, the first of…

  12. Integral coolant channels supply made by melt-out method

    Science.gov (United States)

    Escher, W. J. D.

    1964-01-01

    Melt-out method of constructing strong, pressure-tight fluid coolant channels for chambers is accomplished by cementing pins to the surface and by depositing a melt-out material on the surface followed by two layers of epoxy-resin impregnated glass fibers. The structure is heated to melt out the low-melting alloy.

  13. EVALUATION OF FILTRATION AND DISTILLATION METHODS FOR RECYCLING AUTOMOTIVE COOLANT.

    Science.gov (United States)

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants at a New Jersey Department of Transportation garage. The specific recycling units evaluated are based on the technologies of filtrat...

  14. AUTOMOTIVE AND HEAVY-DUTY ENGINE COOLANT RECYCLING BY DISTILLATION

    Science.gov (United States)

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants for a facility such as the New Jersey Department of Transportation garage in Ewing, New Jersey. he specific recycling evaluated is b...

  15. Mean-field theory for car accidents

    Science.gov (United States)

    Huang, Ding-Wei; Tseng, Wei-Chung

    2001-11-01

    We study analytically the occurrence of car accidents in the Nagel-Schreckenberg traffic model. We obtain exact results for the occurrence of car accidents Pac as a function of the car density ρ and the degree of stochastic braking p1 in the case of speed limit vmax=1. Various quantities are calculated analytically. The nontrivial limit p1-->0 is discussed.

  16. Fukushima accident study using MELCOR

    Institute of Scientific and Technical Information of China (English)

    Randall O Gauntt

    2013-01-01

    The accidents at the Fukushima Daiichi nuclear power station stunned the world as the sequences played out over severals days and videos of hydrogen explosions were televised as they took place.The accidents all resulted in severe damage to the reactor cores and releases of radioactivity to the environment despite heroic measures had taken by the operating personnel.The following paper provides some background into the development of these accidents and their root causes,chief among them,the prolonged station blackout conditions that isolated the reactors from their ultimate heat sink — the ocean.The interpretations given in this paper are summarized from a recently completed report funded by the United States Department of Energy (USDOE).

  17. Correlation between Ni base alloys surface conditioning and cation release mitigation in primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Clauzel, M.; Guillodo, M.; Foucault, M. [AREVA NP SAS, Technical Centre, Le Creusot (France); Engler, N.; Chahma, F.; Brun, C. [AREVA NP SAS, Chemistry and Radiochemistry Group, Paris La Defense (France)

    2010-07-01

    The mastering of the reactor coolant system radioactive contamination is a real stake of performance for operating plants and new builds. The reduction of activated corrosion products deposited on RCS surfaces allows minimizing the global dose integrated by workers which supports the ALARA approach. Moreover, the contamination mastering limits the volumic activities in the primary coolant and thus optimizes the reactor shutdown duration and environment releases. The main contamination sources on PWR are due to Co-60 and Co-58 nuclides which come respectively Co-59 and Ni-58, naturally present in alloys used in the RCS. Co is naturally present as an impurity in alloys or as the main component of hardfacing materials (Stellites™). Ni is released mainly by SG tubes which represent the most important surface of the RCS. PWR steam generators (SG), due to the huge wetted surface are the main source of corrosion products release in the primary coolant circuit. As corrosion products may be transported throughout the whole circuit, activated in the core, and redeposited all over circuit surfaces, resulting in an increase of activity buildup, it is of primary importance to gain a better understanding of phenomenon leading to corrosion product release from SG tubes before setting up mitigation measures. Previous studies have shown that SG tubing made of the same material had different release rates. To find the origin of these discrepancies, investigations have been performed on tubes at the as-received state and after exposure to a nominal primary chemistry in titanium recirculating loop. These investigations highlighted the existence of a correlation between the inner surface metallurgical properties and the release of corrosion products in primary coolant. Oxide films formed in nominal primary chemistry are always protective, their morphology and their composition depending strongly on the geometrical, metallurgical and physico-chemical state of the surface on which they

  18. Review of the correlation between results of cytogenetic dosimetry from blood lymphocytes and EPR dosimetry from tooth enamel for victims of radiation accidents.

    Science.gov (United States)

    Khvostunov, I K; Ivannikov, A I; Skvortsov, V G; Nugis, V Yu; Golub, E V

    2015-03-01

    The goal of this study was to compare dose estimates from electron paramagnetic resonance (EPR) dosimetry with teeth and cytogenetic dosimetry with blood lymphocytes for 30 victims of radiation accidents. The whole-body exposures estimated by tooth enamel EPR dosimetry were ranging from 0.01 to 9.3 Gy. Study group comprised victims exposed to acute and prolonged irradiation at high and low dose rate in different accidents. Blood samples were taken from each of them for cytogenetic analysis. Aberrations were scored and analysed according to International Atomic Energy Agency (IAEA) guidelines for conventional and FISH analysis. Tooth samples were collected in dental clinics after they had been extracted during ordinary practice. EPR dosimetry was performed according to the IAEA protocol. EPR dosimetry showed good correlation with dosimetry based on chromosomal analysis. All estimations of cytogenetic dose below detection limit coincide with EPR dose estimates within the ranges of uncertainty. The differences between cytogenetic and EPR assays may occur in a case of previous unaccounted exposure, non-homogeneous irradiation and due to contribution to absorbed dose from neutron irradiation.

  19. [Blood amylase: a biological marker in irradiation accidents? Preliminary results obtained at the Gustave-Roussy Institut (GRI) and a literature review].

    Science.gov (United States)

    Hennequin, C; Cosset, J M; Cailleux, P E; Girinsky, T; Ganem, G; Hubert, D; Comoy, E; Dutreix, J

    1989-01-01

    The retrospective evaluation of the dose after an irradiation accident is of paramount importance; it allows an adequate selection of patients and the most appropriate treatment can then be proposed. Classical physical dosimetry often lacks precision for dose assessment in such accidents. Cytogenetics, usually more reliable, is not 100% accurate and cannot be used in some particular instances. At the Institut Gustave-Roussy, we studied amylasemia in 15 patients who received a total body irradiation (TBI) for bone marrow grafting, at various dose levels (10, 2 and 1.35 Gy). Hyperamylasemia was found to be constant and dose-dependent. Ten additional patients given a localized irradiation of 2 Gy in the Waldeyer ring had a similar rise in amylasemia as did TBI patients who had received the same dose. In contrast, 13 patients given a pancreatic irradiation (as part of a localized abdominal irradiation) did not show any increase in amylasemia. This study seems to confirm reported data, which suggested that post-TBI hyperamylasemia is almost only related to salivary gland irradiation. Amylasemia could possibly be used as a "biological dosimeter"; however, the dose-effect relationship should be more precisely defined, as well as individual variations. Moreover, the definition of a "threshold-dose" below which hyperamylasemia can never be detected, would be of interest for radioprotection.

  20. Directly connected heat exchanger tube section and coolant-cooled structure

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J.; Coico, Patrick A.; Graybill, David P.; Iyengar, Madhusudan K.; Kamath, Vinod; Kochuparambil, Bejoy J.; Schmidt, Roger R.; Steinke, Mark E.

    2015-09-15

    A method is provided for fabricating a cooling apparatus for cooling an electronics rack, which includes an air-to-liquid heat exchanger, one or more coolant-cooled structures, and a tube. The heat exchanger is associated with the electronics rack and disposed to cool air passing through the rack, includes a plurality of coolant-carrying tube sections, each tube section having a coolant inlet and outlet, one of which is coupled in fluid communication with a coolant loop to facilitate flow of coolant through the tube section. The coolant-cooled structure(s) is in thermal contact with an electronic component(s) of the rack, and facilitates transfer of heat from the component(s) to the coolant. The tube connects in fluid communication one coolant-cooled structure and the other of the coolant inlet or outlet of the one tube section, and facilitates flow of coolant directly between that coolant-carrying tube section of the heat exchanger and the coolant-cooled structure.

  1. Laser accidents: Being Prepared

    Energy Technology Data Exchange (ETDEWEB)

    Barat, K

    2003-01-24

    The goal of the Laser Safety Officer and any laser safety program is to prevent a laser accident from occurring, in particular an injury to a person's eyes. Most laser safety courses talk about laser accidents, causes, and types of injury. The purpose of this presentation is to present a plan for safety offices and users to follow in case of accident or injury from laser radiation.

  2. [Accidents with the "paraglider"].

    Science.gov (United States)

    Lang, T H; Dengg, C; Gabl, M

    1988-09-01

    With a collective of 46 patients we show the details and kinds of accidents caused by paragliding. The base for the casuistry of the accidents was a questionnaire which was answered by most of the injured persons. These were questions about the theoretical and practical training, the course of the flight during the different phases, and the subjective point of view of the course of the accident. The patterns of the injuries showed a high incidence of injuries of the spinal column and high risks for the ankles. At the end, we give some advice how to prevent these accidents.

  3. Accidents (FARS) (National)

    Data.gov (United States)

    Department of Transportation — Accident - (1975-current): This data file (NTAD) contains information about crash characteristics and environmental conditions at the time of the crash. There is one...

  4. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  5. The electrochemistry of IGSCC mitigation in BWR coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, D.D. [Center for Electrochemical Science and Technology, The Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    A brief review is presented of the electrochemical mitigation of IGSCC in water-cooled reactor heat transport circuit structural materials. Electrochemical control and mitigation is possible, because of the existence of a critical potential for IGSCC and by the feasibility of modifying the environment to displace the corrosion potential (ECP) to a value that is more negative than the critical value. However, even in cases where the ECP cannot be displaced sufficiently in the negative direction to become more negative than the critical potential, considerable advantage is accrued, because of the roughly exponential dependence of crack growth rate on potential. The most important parameters in affecting electrochemical control over the ECP and crack growth rate are the kinetic parameters (exchange current densities and Tafel constants) for the redox reactions involving the principal radiolysis products of water (O{sub 2}, H{sub 2}, H{sub 2}O{sub 2}), external solution composition (concentrations of O{sub 2}, H{sub 2}O{sub 2}, and H{sub 2}), flow velocity, and the conductivity of the bulk environment. The kinetic parameters for the redox reactions essentially determine the charge transfer impedance of the steel surface, which is shown to be one of the key parameters in affecting the magnitude of the coupling current and hence the crack growth rate. The exchange current densities, in particular, are amenable to control by catalysis or inhibition, with the result that surface modification techniques are highly effective in controlling and mitigating IGSCC in reactor coolant circuit materials. (authors)

  6. Fuel-Coolant Interaction visualization in TROI test facility

    Energy Technology Data Exchange (ETDEWEB)

    Na, Young Su; Hong, Seong-Ho; Song, Jin Ho; Hong, Seong-Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    It is necessary to observe the FCI (Fuel-Coolant Interaction) phenomena at the condition of vessel failure to IVR. We carried out a visualization test on the interaction of a corium melt and water to observe the premixing phase without a free fall of a melt jet in a gas phase before contacting the cooling water. This paper is based on the previous study presented at Ninth Korea-Japan Symposium on Nuclear Hydraulics and Safety, we added the results on sieved debris distribution. The visualization test on the FCI without a free fall of a corium melt jet in a gas phase was conducted carefully in the TROI test facility. A prototypic corium consisting of uranium oxide and zirconium oxide with a weight ratio of UO{sub 2} to ZrO{sub 2} of 80 to 20, respectively, was heated up using the induction heating method. It was observed that a corium melt jet penetrated into water with 1000 mm in depth, and it took about 0.6 seconds from opening the releasing valve, which was confirmed by the sequential variation of the temperature measured by the sacrificial thermocouples installed in the direction of a falling melt jet. The cumulative mass fraction of the debris smaller than 1.0 mm was 15%, and the mass mean diameter of the debris was 2.9 mm. This visualization test can generate the valuable information such as the behavior of the corium melt jet and the size of mixing zone for validating the computer code.

  7. Elevated-pressure mixed-coolants Joule Thomson cryocooling

    Science.gov (United States)

    Maytal, B.-Z.; Nellis, G. F.; Klein, S. A.; Pfotenhauer, J. M.

    2006-01-01

    This paper explores the potential of mixed coolants at elevated pressures for Joule-Thomson cryocooling. A numerical model of a Joule-Thomson cryocooler is developed that is capable of simulating operation with mixtures of up to 9 components consisting of hydrocarbons, non-flammable halogenated refrigerants, and inert gases. The numerical model is integrated with a genetic optimization algorithm, which has a high capability for convergence in an environment of discontinuities, constraints and local optima. The genetic optimization algorithm is used to select the optimal mixture compositions that separately maximizes following two objective functions at each elevated pressure for 80, 90 and 95 K cryocooling: the molar specific cooling capacity (the highest attainable is 3200 J/mol) and the produced cooling capacity per thermal conductance which is a measure of the compactness of the recuperator. The optimized cooling capacity for a non-flammable halogenated refrigerant mixture is smaller than for a hydrocarbon mixture; however, the cooling capacity of the two types of mixtures approach one another as pressure becomes higher. The coefficient of performance, the required heat transfer area and the effect of the number of components in the mixture is investigated as a function of the pressure. It is shown that mixtures with more components provide a higher cooling capacity but require larger recuperative heat exchangers. Optimized mixtures for 90 K cryocooling have similar cooling capacity as those for 80 K. Optimized compactness for 80 K is about 50% higher than can be achieved by pure nitrogen. For 90 K, no mixture provides a more compact recuperator than can be achieved using pure argon. The results are discussed in the context of potential applications for closed and open cycle cryocoolers.

  8. BACKUP AND ULTIMATE HEAT SINKS IN CANDU REACTORS FOR PROLONGED SBO ACCIDENTS

    OpenAIRE

    T. NITHEANANDAN; BROWN, M J

    2013-01-01

    In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. ...

  9. Scram discharge volume break studies accident sequence analysis

    Energy Technology Data Exchange (ETDEWEB)

    Harrington, R.M.; Hodge, S.A.

    1982-01-01

    This paper is a summary of a report describing the predicted response of Unit 1 at the Tennessee Valley Authority (TVA) Browns Ferry Nuclear Plant to a hypothetical small break loss of coolant accident (SBLOCA) outside of containment. The accident studied would be initiated by a break in the scram discharge volume (SDV) piping when it is pressurized to full reactor vessel pressure as a normal consequence of a reactor scram. If the scram could be reset, the scram outlet valves would close to isolate the SDV and the piping break from the reactor vessel. However, reset is possible only if the conditions that caused the scram have cleared; it has been assumed in this study that the scram signal remains in effect over a long period of time.

  10. 49 CFR 178.345-8 - Accident damage protection.

    Science.gov (United States)

    2010-10-01

    ... or more than 2 inches from the point of attachment to the tank must have accident damage protection... accident damage protection device; or (iii) Attached to the cargo tank in accordance with the requirements... damaged in an accident could result in a loss of lading from the cargo tank must be protected by...

  11. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  12. MARINE ACCIDENTS RESEARCHED THROUGH HUMAN FACTOR PRISMA

    Directory of Open Access Journals (Sweden)

    Branislav M Ćorović

    2013-07-01

    Full Text Available We are aware of a large number of marine accidents that result in numerous casualties and even deaths and substantial negative environmental effects. The objective of this paper is to indicate factors that contribute to human errors which is identified as the most frequent cause to marine accidents. Despite rapid technological development and safety legislation, this paper identifies the human factor as the waekest link in maritime safety system. This analysis could lead to decrease of vessel accidents. In addition, starting from the European Maritime Safety Agency data and by linear regression model application, we have obtained the trend of number of ships involved in marine accidents as well as the trend of lives lost in marine accidents  in and around European Union waters.

  13. CFD Analyses of Air-Ingress Accident for VHTRs

    Science.gov (United States)

    Ham, Tae Kyu

    -ingress mechanism and to utilize the CFD simulation in the analysis of the phenomenon. Previous air-ingress studies simulated the depressurization process using simple assumptions or 1-D system code results. However, recent studies found flow oscillations near the end of the depressurization which could influence the next stage of the air-ingress accident. Therefore, CFD simulations were performed to examine the air-ingress mechanisms from the depressurization through the establishment of local natural circulation initiate. In addition to the double-guillotine break scenario, there are other scenarios that can lead to an air-ingress event such as a partial break were in the cross vessel with various break locations, orientations, and shapes. These additional situations were also investigated. The simulation results for the OSU test facility showed that the discharged helium coolant from a reactor vessel during the depressurization process will be mixed with the air in the containment. This process makes the density of the gas mixture in the containment lower and the density-driven air-ingress flow slower because the density-driven flow is established by the density difference of the gas species between the reactor vessel and the containment. In addition, for the simulations with various initial and boundary conditions, the simulation results showed that the total accumulated air in the containment collapsed within 10% standard deviation by: 1. multiplying the density ratio and viscosity ratio of the gas species between the containment and the reactor vessel and 2. multiplying the ratio of the air mole fraction and gas temperature to the reference value. By replacing the gas mixture in the reactor cavity with a gas heavier than the air, the air-ingress speed slowed down. Based on the understanding of the air-ingress phenomena for the GT-MHR air-ingress scenario, several mitigation measures of air-ingress accident are proposed. The CFD results are utilized to plan experimental

  14. Numerical investigation on thermal striping conditions for a tee junction of LMFBR coolant pipes. 3. Investigation on diameter ratio between the coolant pipes

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1999-04-01

    This report presents numerical results on thermal striping characteristics at a tee junction of LMFBR coolant pipe, carried out using a direct numerical simulation code DINUS-3. In the numerical investigations, it was considered a tee junction system consisted of a main pipe (1.33 cm{sup I.D.}) with a 90deg elbow and a branch pipe having various inner diameters, and five diameter ratio conditions between both the pipes, i.e., (D{sub main}/D{sub branch}) = 1.0, 2.0, 3.0, 5.0 and 10.0. From the numerical investigations, the following characteristics were obtained: (1) Maximum sodium temperature fluctuation amplitude in the downstream region of the tee junction were decreased with increasing of the diameter ratio (decreasing of the branch pipe diameter). One of the main reasons for this behavior was considered to be that the affects of the branch pipe jet for the main pipe flows was decreased with decreasing of the branch pipe diameter. (2) Auto-power spectral density levels were decreased by the increasing of the diameter ratio. It was indicated that coolant mixing characteristics in the downstream region were controlled by locally random turbulence processes. (3) To suppress sodium temperature fluctuations in the downstream region, it is a suitable combination for larger velocity ratio and larger diameter ratio. (author)

  15. Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2015-01-01

    Full Text Available The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs. Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel. The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

  16. Effect of hypoiodous acid volatility on the iodine source term in reactor accidents

    Energy Technology Data Exchange (ETDEWEB)

    Routamo, T. [Imatran Voima Oy, Vantaa (Finland)

    1996-12-01

    A FORTRAN code ACT WATCH has been developed to establish an improved understanding of essential radionuclide behaviour mechanisms, especially related to iodine chemistry, in reactor accidents. The accident scenarios calculated in this paper are based on the Loss of Coolant accident at the Loviisa Nuclear Power Plant. The effect of different airborne species, especially HIO, on the iodine source term has been studied. The main cause of the high HIO release in the system modelled is the increase of I{sub 2} hydrolysis rate along with the temperature increase, which accelerates HIO production. Due to the high radiation level near the reactor core, I{sub 2} is produced from I{sup -}very rapidly. High temperature in the reactor coolant causes I{sub 2} to be transformed into HIO and through the boiling of the coolant volatile I{sub 2} and HIO are transferred efficiently into the gas phase. High filtration efficiency for particulate iodine causes I{sup -} release to be much lower than those of I{sub 2} and HIO. (author) 15 figs., 1 tab., refs.

  17. Hybrid method for numerical modelling of LWR coolant chemistry

    Science.gov (United States)

    Swiatla-Wojcik, Dorota

    2016-10-01

    A comprehensive approach is proposed to model radiation chemistry of the cooling water under exposure to neutron and gamma radiation at 300 °C. It covers diffusion-kinetic processes in radiation tracks and secondary reactions in the bulk coolant. Steady-state concentrations of the radiolytic products have been assessed based on the simulated time dependent concentration profiles. The principal reactions contributing to the formation of H2, O2 and H2O2 were indicated. Simulation was carried out depending on the amount of extra hydrogen dissolved in the coolant to reduce concentration of corrosive agents. High sensitivity to the rate of reaction H+H2O=OH+H2 is shown and discussed.

  18. Communication and industrial accidents

    NARCIS (Netherlands)

    As, Sicco van

    2001-01-01

    This paper deals with the influence of organizational communication on safety. Accidents are actually caused by individual mistakes. However the underlying causes of accidents are often organizational. As a link between these two levels - the organizational failures and mistakes - I suggest the conc

  19. Accident investigation and analysis

    NARCIS (Netherlands)

    Kampen, J. van; Drupsteen, L.

    2013-01-01

    Many organisations and companies take extensive proactive measures to identify, evaluate and reduce occupational risks. However, despite these efforts things still go wrong and unintended events occur. After a major incident or accident, conducting an accident investigation is generally the next ste

  20. Effect of coolant inhibitors on AZ91D

    Institute of Scientific and Technical Information of China (English)

    I.M. Baghni; WU Yinshun; ZHANG Wei; LI Jiuqing

    2004-01-01

    The inhibition effects of sodium vanadate along with inorganic coolant inhibitors were examined on corrosion of AZ91D in ASTM D1384-80 corrosive water by polarization measurements. The galvanic corrosion of AZ91D coupled to 3003, 6063, and 356 Al alloys were also tested. An effective combination of inhibitors containing (but not limited to) sodium vanadate, silicate, and nitrate was proposed for inhibition of AZ91D and prevention of galvanic corrosion.

  1. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  2. Design of Reactor Coolant Pump Seal Online Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Ah, Sang Ha; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of); Lee, Song Kyu [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2008-05-15

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation.

  3. Coupling a system code with computational fluid dynamics for the simulation of complex coolant reactivity effects

    Energy Technology Data Exchange (ETDEWEB)

    Bertolotto, D.

    2011-11-15

    The current doctoral research is focused on the development and validation of a coupled computational tool, to combine the advantages of computational fluid dynamics (CFD) in analyzing complex flow fields and of state-of-the-art system codes employed for nuclear power plant (NPP) simulations. Such a tool can considerably enhance the analysis of NPP transient behavior, e.g. in the case of pressurized water reactor (PWR) accident scenarios such as Main Steam Line Break (MSLB) and boron dilution, in which strong coolant flow asymmetries and multi-dimensional mixing effects strongly influence the reactivity of the reactor core, as described in Chap. 1. To start with, a literature review on code coupling is presented in Chap. 2, together with the corresponding ongoing projects in the international community. Special reference is made to the framework in which this research has been carried out, i.e. the Paul Scherrer Institute's (PSI) project STARS (Steady-state and Transient Analysis Research for the Swiss reactors). In particular, the codes chosen for the coupling, i.e. the CFD code ANSYS CFX V11.0 and the system code US-NRC TRACE V5.0, are part of the STARS codes system. Their main features are also described in Chap. 2. The development of the coupled tool, named CFX/TRACE from the names of the two constitutive codes, has proven to be a complex and broad-based task, and therefore constraints had to be put on the target requirements, while keeping in mind a certain modularity to allow future extensions to be made with minimal efforts. After careful consideration, the coupling was defined to be on-line, parallel and with non-overlapping domains connected by an interface, which was developed through the Parallel Virtual Machines (PVM) software, as described in Chap. 3. Moreover, two numerical coupling schemes were implemented and tested: a sequential explicit scheme and a sequential semi-implicit scheme. Finally, it was decided that the coupling would be single

  4. Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

    Directory of Open Access Journals (Sweden)

    Bo Cheng

    2016-02-01

    Full Text Available In severe loss of coolant accidents (LOCA, similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in 1,200–1,500°C steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstrated corrosion resistance. As these composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Mo alloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are

  5. Developing techniques for cause-responsibility analysis of occupational accidents.

    Science.gov (United States)

    Jabbari, Mousa; Ghorbani, Roghayeh

    2016-11-01

    The aim of this study was to specify the causes of occupational accidents, determine social responsibility and the role of groups involved in work-related accidents. This study develops occupational accidents causes tree, occupational accidents responsibility tree, and occupational accidents component-responsibility analysis worksheet; based on these methods, it develops cause-responsibility analysis (CRA) techniques, and for testing them, analyzes 100 fatal/disabling occupational accidents in the construction setting that were randomly selected from all the work-related accidents in Tehran, Iran, over a 5-year period (2010-2014). The main result of this study involves two techniques for CRA: occupational accidents tree analysis (OATA) and occupational accidents components analysis (OACA), used in parallel for determination of responsible groups and responsibilities rate. From the results, we find that the management group of construction projects has 74.65% responsibility of work-related accidents. The developed techniques are purposeful for occupational accidents investigation/analysis, especially for the determination of detailed list of tasks, responsibilities, and their rates. Therefore, it is useful for preventing work-related accidents by focusing on the responsible group's duties.

  6. Severe Accident Scoping Simulations of Accident Tolerant Fuel Concepts for BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations [1]. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramic microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs) [2]. Researchers have investigated the effects of thermal conductivity on design basis accidents [3], investigated silicon carbide (SiC) cladding [4], as well as the effects of ATF concepts on the late stage accident progression [5]. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 [6] and the South Korean OPR-1000 [7] pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts [8]. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl) [9]. With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, Fe

  7. Analysis of Fukushima Daiichi Accident Using HFACS

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Saeed Almheiri [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    The shadow of Fukushima Daiichi nuclear power plant (NPP) accident is still too big and will last long. On the other hand, it could still teach us lots of lessons to better design and operate nuclear power plants. In this paper, we will be focusing on the Fukushima Daiichi accident, especially on human organizational factors. We will analyze the accident using Human Factors Analysis and Classification System (HFACS) in order to better understand the organizational climate of TEPCO{sup 1} and NISA{sup 2} that led to Fukushima Daiichi Accident. HFACS was developed for the U. S. aviation industry and has been used at many industries like the rail and mining industries. We found that the HFACS to be greatly beneficial in investigating the latent and organizational causes for the accident. The application results show that the causes of Fukushima Daiichi accident were spread out from sharp end (i.e. Unsafe Act) to blunt end (i. e. Organizational Influences). This means that the corresponding countermeasures should cover from front line staff to management. Thus, we managed to develop a better understanding on how to prevent similar errors or violations. The incident and near-miss have a lot of helpful information because it may show the actual and latent deficiencies of complex systems. We applied the HFACS into Fukushima Daiichi accident to better locate the causes related to both sharp and blunt ends of operation of NPP. In order to derive useful lessons from the accident analysis, the analyst should try to find the similarities not differences from the incident. It is imperative that whatever accident/incident analysis systems we use, we should fully utilize the disastrous Fukushima accident.

  8. Geometric and Road Environmental Effects against Total Number of Traffic Accidents in Kendari

    Science.gov (United States)

    Kurdin, M. Akbar; Welendo, La; Annisa, Nur

    2017-05-01

    From the large number of traffic accidents that occurred, the carrying of Kendari as the biggest contributor to accidents in the Southeast. The number of accidents in Kendari row since 2011 was recorded at 18 accidents due to the influence of geometric road, in 2012 registered at 13 accident and in 2013 amounted to 6 accidents, with accident data because of the influence Geometric recorded for 3 consecutive years the biggest contributor to accidents because of the influence of geometric is Abeli districts. This study aimed to determine the road which common point of accident-prone (Black spot) in Kecamatan Abeli as accident-prone areas in Kendari, analyze the influence of geometric and road environment against accidents on roads in Kecamatan Abeli, provide alternative treatment based on the causes of accidents on the location of the accident-prone points (blackspot) to reduce the rate of traffic accidents. From the results of a study of 6 curve the accident-prone locations, that the curve I, II, and VI is the “Black Spot” influenced by the amount and condition of traffic accidents, while at the curve II, a traffic accident that occurred also be caused by unsafe geometric where the type of geometric should be changed from Spiral-Spiral type to Spiral-Circle-Spiral type. This indicates geometric effect on the number of accidents.

  9. Comparison of CATHARE results with the experimental results of cold leg intermediate break LOCA obtained during ROSA-2/LSTF test 7

    Directory of Open Access Journals (Sweden)

    Mazgaj Piotr

    2016-01-01

    Full Text Available Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant design and operation. In the field of Loss Of Coolant Accident, evolutions of the regulations are discussed in various countries taking into account the very unlikely character of a double-ended guillotine break and questioning the necessity to study such an event with Design Basis Conditions assumptions. As a consequence, the consideration of intermediate size piping rupture becomes more and more important. The paper presents the modeling of the Test Facility ROSA-2/LSTF in the calculation code CATHARE 2.V2.5. OECD/NEA ROSA-2 Project Test 7 was conducted with the Large Scale Test Facility on June 14, 2012. The experiment simulated the thermal-hydraulic responses during a PWR 13% cold leg Intermediate Break Loss Of Coolant Accident (IBLOCA. The break was simulated by a cold leg upwardly mounted long break nozzle. The facility and the experiment conditions are modeled in CATHARE. The vessel is modeled by using a 3D module. A thermal-hydraulic analysis is conducted and the obtained results are subsequently compared with the experimental results from ROSA-2/LSTF Test 7. Evaluation of the differences between experimental and calculated results is discussed.

  10. Modeling secondary accidents identified by traffic shock waves.

    Science.gov (United States)

    Junhua, Wang; Boya, Liu; Lanfang, Zhang; Ragland, David R

    2016-02-01

    The high potential for occurrence and the negative consequences of secondary accidents make them an issue of great concern affecting freeway safety. Using accident records from a three-year period together with California interstate freeway loop data, a dynamic method for more accurate classification based on the traffic shock wave detecting method was used to identify secondary accidents. Spatio-temporal gaps between the primary and secondary accident were proven be fit via a mixture of Weibull and normal distribution. A logistic regression model was developed to investigate major factors contributing to secondary accident occurrence. Traffic shock wave speed and volume at the occurrence of a primary accident were explicitly considered in the model, as a secondary accident is defined as an accident that occurs within the spatio-temporal impact scope of the primary accident. Results show that the shock waves originating in the wake of a primary accident have a more significant impact on the likelihood of a secondary accident occurrence than the effects of traffic volume. Primary accidents with long durations can significantly increase the possibility of secondary accidents. Unsafe speed and weather are other factors contributing to secondary crash occurrence. It is strongly suggested that when police or rescue personnel arrive at the scene of an accident, they should not suddenly block, decrease, or unblock the traffic flow, but instead endeavor to control traffic in a smooth and controlled manner. Also it is important to reduce accident processing time to reduce the risk of secondary accident. Copyright © 2015 Elsevier Ltd. All rights reserved.

  11. Estimation of the contribution of gamma-emission of incorporated cesium radioisotopes in interpretation of the results of the public survey to assess the thyroidal iodine content following a radiation accident at the nuclear power plant

    Directory of Open Access Journals (Sweden)

    Shinkarev S.M.

    2014-12-01

    Full Text Available Aim. A detail consideration has been done to assess an importance of the contribution of gamma-emission of incorporated cesium radioisotopes to the exposure rate measured near the thyroid by the public survey for following the Chernobyl accident. Empirical ratios have been derived to take into account that contribution under interpretation of the results of survey meter monitoring of the public. Materials and methods. Model calculations for typical radionuclide intake by the residents living in contaminated territories after the Chernobyl accident have been carried out in order to assess the contribution of gamma-emission of incorporated cesium radioisotopes to the exposure rate measured near the thyroid by the survey. Under such calculations two the most important modes of intake have been considered: 1 inhalation and 2 ingestion with cow milk. Results. According to the estimates received the contribution of gamma-emission of incorporated cesium radioisotopes to the exposure rate measured near the thyroid during the first 20 days does not exceed 20% for the residents of southern areas of Gomel region and 30% for the residents of Mogil'yov region. During 60 days following the accident that contribution is estimated to be within (50-80 % for the residents of southern areas of Gomel region and (80-95 % for the residents of Mogil'yov region. Conclusion. For the period of intensive thyroid measuring in the southern areas of Gomel region (the second part of May account of the contribution of gamma-emission of incorporated cesium radioisotopes is relatively unimportant, but for Mogil'yov region (the end of May — it is important to account for. For the thyroid measurements conducted in June of 1986 it is important for all residents living in Belarus to take into account the contribution of gamma-emission of incorporated cesium radioisotopes.

  12. Inclusion of models to describe severe accident conditions in the fuel simulation code DIONISIO

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Daverio, Hernando [Gerencia Reactores y Centrales Nucleares, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Denis, Alicia [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina)

    2017-04-15

    The simulation of fuel rod behavior is a complex task that demands not only accurate models to describe the numerous phenomena occurring in the pellet, cladding and internal rod atmosphere but also an adequate interconnection between them. In the last years several models have been incorporated to the DIONISIO code with the purpose of increasing its precision and reliability. After the regrettable events at Fukushima, the need for codes capable of simulating nuclear fuels under accident conditions has come forth. Heat removal occurs in a quite different way than during normal operation and this fact determines a completely new set of conditions for the fuel materials. A detailed description of the different regimes the coolant may exhibit in such a wide variety of scenarios requires a thermal-hydraulic formulation not suitable to be included in a fuel performance code. Moreover, there exist a number of reliable and famous codes that perform this task. Nevertheless, and keeping in mind the purpose of building a code focused on the fuel behavior, a subroutine was developed for the DIONISIO code that performs a simplified analysis of the coolant in a PWR, restricted to the more representative situations and provides to the fuel simulation the boundary conditions necessary to reproduce accidental situations. In the present work this subroutine is described and the results of different comparisons with experimental data and with thermal-hydraulic codes are offered. It is verified that, in spite of its comparative simplicity, the predictions of this module of DIONISIO do not differ significantly from those of the specific, complex codes.

  13. Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

    Energy Technology Data Exchange (ETDEWEB)

    Coryell, E.W.; Siefken, L.J.; Harvego, E.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1997-07-01

    The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.

  14. Persistence of airline accidents.

    Science.gov (United States)

    Barros, Carlos Pestana; Faria, Joao Ricardo; Gil-Alana, Luis Alberiko

    2010-10-01

    This paper expands on air travel accident research by examining the relationship between air travel accidents and airline traffic or volume in the period from 1927-2006. The theoretical model is based on a representative airline company that aims to maximise its profits, and it utilises a fractional integration approach in order to determine whether there is a persistent pattern over time with respect to air accidents and air traffic. Furthermore, the paper analyses how airline accidents are related to traffic using a fractional cointegration approach. It finds that airline accidents are persistent and that a (non-stationary) fractional cointegration relationship exists between total airline accidents and airline passengers, airline miles and airline revenues, with shocks that affect the long-run equilibrium disappearing in the very long term. Moreover, this relation is negative, which might be due to the fact that air travel is becoming safer and there is greater competition in the airline industry. Policy implications are derived for countering accident events, based on competition and regulation. © 2010 The Author(s). Journal compilation © Overseas Development Institute, 2010.

  15. Accidents with sulfuric acid

    Directory of Open Access Journals (Sweden)

    Rajković Miloš B.

    2006-01-01

    Full Text Available Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eighteen years (from 1988 till the beginning of 2006 are analyzed in this paper. It is very alarming data that, according to all the recorded accidents, over 1.6 million tons of sulfuric acid were exuded. Although water transport is the safest (only 16.38% of the total amount of accidents in that way 98.88% of the total amount of sulfuric acid was exuded into the environment. Human factor was the common factor in all the accidents, whether there was enough control of the production process, of reservoirs or transportation tanks or the transport was done by inadequate (old tanks, or the accidents arose from human factor (inadequate speed, lock of caution etc. The fact is that huge energy, sacrifice and courage were involved in the recovery from accidents where rescue teams and fire brigades showed great courage to prevent real environmental catastrophes and very often they lost their lives during the events. So, the phrase that sulfuric acid is a real "environmental bomb" has become clearer.

  16. Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5; Simulacion de un escenario de perdida de refrigerante grande (LBLOCA), sin actuacion de los sistemas de inyeccion de emergencia (ECCS) para un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Lopez M, R., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm{sup 2} and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

  17. Effect of Coolant Temperature and Mass Flow on Film Cooling of Turbine Blades

    Science.gov (United States)

    Garg, Vijay K.; Gaugler, Raymond E.

    1997-01-01

    A three-dimensional Navier Stokes code has been used to study the effect of coolant temperature, and coolant to mainstream mass flow ratio on the adiabatic effectiveness of a film-cooled turbine blade. The blade chosen is the VKI rotor with six rows of cooling holes including three rows on the shower head. The mainstream is akin to that under real engine conditions with stagnation temperature = 1900 K and stagnation pressure = 3 MPa. Generally, the adiabatic effectiveness is lower for a higher coolant temperature due to nonlinear effects via the compressibility of air. However, over the suction side of shower-head holes, the effectiveness is higher for a higher coolant temperature than that for a lower coolant temperature when the coolant to mainstream mass flow ratio is 5% or more. For a fixed coolant temperature, the effectiveness passes through a minima on the suction side of shower-head holes as the coolant to mainstream mass flow, ratio increases, while on the pressure side of shower-head holes, the effectiveness decreases with increase in coolant mass flow due to coolant jet lift-off. In all cases, the adiabatic effectiveness is highly three-dimensional.

  18. Bilateral Shoulder Dislocations Following a Motocross Accident

    National Research Council Canada - National Science Library

    H J Schroepfer; B M Martin; P J Millett

    2017-01-01

    .... A professional motocross racer experienced a high speed traumatic accident which resulted in bilateral shoulder dislocations with a rare associated complete anterosuperior RC tear due to a posterior dislocation...

  19. Road accidents and business cycles in Spain.

    Science.gov (United States)

    Rodríguez-López, Jesús; Marrero, Gustavo A; González, Rosa Marina; Leal-Linares, Teresa

    2016-11-01

    This paper explores the causes behind the downturn in road accidents in Spain across the last decade. Possible causes are grouped into three categories: Institutional factors (a Penalty Point System, PPS, dating from 2006), technological factors (active safety and passive safety of vehicles), and macroeconomic factors (the Great recession starting in 2008, and an increase in fuel prices during the spring of 2008). The PPS has been blessed by incumbent authorities as responsible for the decline of road fatalities in Spain. Using cointegration techniques, the GDP growth rate, the fuel price, the PPS, and technological items embedded in motor vehicles appear to be statistically significantly related with accidents. Importantly, PPS is found to be significant in reducing fatal accidents. However, PPS is not significant for non-fatal accidents. In view of these results, we conclude that road accidents in Spain are very sensitive to the business cycle, and that the PPS influenced the severity (fatality) rather than the quantity of accidents in Spain. Importantly, technological items help explain a sizable fraction in accidents downturn, their effects dating back from the end of the nineties.

  20. Investigation of relationship between mental workload and information flow rate of accident diagnosis tasks in NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Chang Hoon

    2005-02-15

    The objective of this study is to investigate experimentally the relationship between an operator's mental workload and the information flow rate of accident diagnosis tasks and further to propose the information flow rate as an analytic method for measuring the mental workload. There are two types of mental workload in the advanced MCR of NPPs: the information processing workload, which is the processing that the human operator must actually perform in order to complete the diagnosis task, and emotional stress workload experienced by the operator. In this study, the focus is on the former. Three kinds of methods are used to measure the operator's workload: information flow rate, subjective methods, and physiological measures. Information flows for eight accident diagnosis tasks are modeled qualitatively using a stage model and are quantified using Conant's model. The eight accident cases are considered here are: Loss Of Coolant Accident (LOCA), Steam Generator Tube Rupture (SGTR), Steam Line Break (SLB), Feedwater Line Break (FLB), Pressurizer (PZR) spray and heater failure, Reactor Coolant Pump (RCP) trip, Main Steam Isolation Valve (MSIV) failure, and PZR spray failure. The information flow rate is obtained for each diagnosis task by imposing time limit restrictions for the tasks. Subjective methods require the operators to respond to questionnaires to rate their level of mental effort. NASA-TLX and MCH scale are selected as subjective methods. NASA-TLX is a subjective method used in the various fields including the aviation, automobile, and nuclear industries. It has a multi-dimensional rating technique and provides an overall workload score based on a weighted average on six subscales using pair-wise comparison tests. MCH, on the other hand, is one-dimensional and uses a 10- point rating technique. As with NASA-TLX, the higher the score is, the higher the subjective workload is. For the physiological measurements, an eye tracking system analyzes

  1. Radionuclides release possibility analysis of MSR at various accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Wie; Kim, Hee Reyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    There are some accidents which go beyond our expectation such as Fukushima Daiichi nuclear disaster and amounts of radionuclides release to environment, so more effort and research are conducted to prevent it. MSR (Molten Salt Reactor) is one of GEN-IV reactor types, and its coolant and fuel are mixtures of molten salt. MSR has a schematic like figure 1 and it has different features with the solid fuel reactor, but most important and interesting feature of MSR is its many safety systems. For example, MSR has a large negative void coefficient. Even though power increases, the reactor slows down soon. Radionuclides release possibility of MSR was analyzed at various accident conditions including Chernobyl and Fukushima ones. The MSR was understood to prevent the severe accident by the negative reactivity coefficient and the absence of explosive material such as water at the Chernobyl disaster condition. It was expected to contain fuel salts in the reactor building and not to release radionuclides into environment even if the primary system could be ruptured or broken and fuel salts would be leaked at the Fukushima Daiichi nuclear disaster condition of earthquake and tsunami. The MSR, which would not lead to the severe accident and therefore prevents the fuel release to the environment at many expected scenarios, was thought to have priority in the aspect of accidents. A quantitative analysis and a further research are needed to evaluate the possibility of radionuclide release to the environment at the various accident conditions based on the simple comparison of the safety feature between MSR and solid fuel reactor.

  2. Boating 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...

  3. Accident resistant transport container

    Science.gov (United States)

    Andersen, John A.; Cole, James K.

    1980-01-01

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  4. The Fukushima accident; Accident nucleaire a Fukushima

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, D.

    2012-02-15

    The Fukushima accident is characterized by a sequence of natural disasters: earthquake and tsunamis that deprived simultaneously 3 reactors from cooling and electrical power for quite a long time. A series of hydrogen explosion has added to the mess. Experts agree to say that certainly nuclear fuel has melt to form corium in all 3 reactors. The accident has contaminated tens of thousand acres of land around the plant and has jeopardized local coastal fishery. The human toll is unexpectedly low: no direct casualty in the population but several suicides among the people that was forced to leave their home. 5 people from the plant staff died certainly from the consequences of the tsunami. (A.C.)

  5. Core burnup calculation and accidents analyses of a pressurized water reactor partially loaded with rock-like oxide fuel

    Science.gov (United States)

    Akie, H.; Sugo, Y.; Okawa, R.

    2003-06-01

    A rock-like oxide (ROX) fuel - light water reactor (LWR) burning system has been studied for efficient plutonium transmutation. For the improvement of small negative reactivity coefficients and severe transient behaviors of ROX fueled LWRs, a partial loading core of ROX fuel assemblies with conventional UO 2 assemblies was considered. As a result, although the reactivity coefficients could be improved, the power peaking tends to be large in this heterogeneous core configuration. The reactivity initiated accident (RIA) and loss of coolant accident (LOCA) behaviors were not sufficiently improved. In order to reduce the power peaking, the fuel composition and the assembly design of the ROX fuel were modified. Firstly, erbium burnable poison was added as Er 2O 3 in the ROX fuel to reduce the burnup reactivity swing. Then pin-by-pin Pu enrichment and Er content distributions within the ROX fuel assembly were considered. In addition, the Er content distribution was also considered in the axial direction of the ROX fuel pin. With these modifications, a power peaking factor even lower than the one in a conventional UO 2 fueled core can be obtained. The RIA and LOCA analyses of the modified core have also shown the comparable transient behaviors of ROX partial loading core to those of the UO 2 core.

  6. Development of Innovative Accident Tolerant High Thermal Conductivity UO2-Diamond Composite Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    Tulenko, James [Univ. of Florida, Gainesville, FL (United States); Subhash, Ghatu [Univ. of Florida, Gainesville, FL (United States)

    2016-01-01

    The University of Florida (UF) evaluated a composite fuel consisting of UO2 powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO2-Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO2 – SiC and UO2 – Carbon Nanotube fuel pins. UF is proving with the current research results that the addition of diamond micro particles to UO2 may greatly enhanced the thermal conductivity of the UO2 pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.

  7. Accidents with sulfuric acid

    OpenAIRE

    Rajković Miloš B.

    2006-01-01

    Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eigh...

  8. GIRAFFE test results summary

    Energy Technology Data Exchange (ETDEWEB)

    Yokobori, S.; Arai, K.; Oikawa, H. [Toshiba Corporation, Kawasaki (Japan)

    1996-03-01

    A passive system can provide engineered safety features enhancing safety system reliability and plant simplicity. Toshiba has conducted the test Program to demonstrate the feasibility of the SBWR passive safety system using a full-height, integral system test facility GIRAFFE. The test facility GIRAFFE models the SBWR in full height to correctly present the gravity driving head forces with a 1/400 volume scale. The GIRAFFE test Program includes the certification tests of the passive containment cooling system (PCCS) to remove the post-accident decay heat and the gravity driven cooling system (GDCS) to replenish the reactor coolant inventory during a LOCA. The test results have confirmed the PCCS and GDCS design and in addition, have demonstrated the operation of the pCCS with the presence of a lighter-than-steam noncondensable as well as with the presence of a heavier-than-steam, noncondensable. The GIRAFFE test Program has also provided the database to qualify a best estimate thermal-hydraulic computer code TRAC. The post test analysis results have shown that TRAC can accurately predict the PCCS heat removal Performance and the containment pressure response to a LOCA. This paper summarizes the GIRAFFE test results to investigate post-LOCA PCCS heat removal performance and post-test analysis using TRAC.

  9. Benchmarking of Decay Heat Measured Values of ITER Materials Induced by 14 MeV Neutron Activation with Calculated Results by ACAB Activation Code

    Energy Technology Data Exchange (ETDEWEB)

    Tore, C.; Ortego, P.; Rodriguez Rivada, A.

    2014-07-01

    The aim of this paper is the comparison between the calculated and measured decay heat of material samples which were irradiated at the Fusion Neutron Source of JAERI in Japan with D-T production of 14MeV neutrons. In the International Thermonuclear Experimental Reactor (ITER) neutron activation of the structural material will result in a source of heat after shutdown of the reactor. The estimation of decay heat value with qualified codes and nuclear data is an important parameter for the safety analyses of fusion reactors against lost of coolant accidents. When a loss of coolant and/or flow accident happen plasma facing components are heated up by decay heat. If the temperature of the components exceeds the allowable temperature, the accident would expand to loose the integrity of ITER. Uncertainties associated with decay prediction less than 15% are strongly requested by the ITER designers. Additionally, accurate decay heat prediction is required for making reasonable shutdown scenarios of ITER. (Author)

  10. Bayes classifiers for imbalanced traffic accidents datasets.

    Science.gov (United States)

    Mujalli, Randa Oqab; López, Griselda; Garach, Laura

    2016-03-01

    Traffic accidents data sets are usually imbalanced, where the number of instances classified under the killed or severe injuries class (minority) is much lower than those classified under the slight injuries class (majority). This, however, supposes a challenging problem for classification algorithms and may cause obtaining a model that well cover the slight injuries instances whereas the killed or severe injuries instances are misclassified frequently. Based on traffic accidents data collected on urban and suburban roads in Jordan for three years (2009-2011); three different data balancing techniques were used: under-sampling which removes some instances of the majority class, oversampling which creates new instances of the minority class and a mix technique that combines both. In addition, different Bayes classifiers were compared for the different imbalanced and balanced data sets: Averaged One-Dependence Estimators, Weightily Average One-Dependence Estimators, and Bayesian networks in order to identify factors that affect the severity of an accident. The results indicated that using the balanced data sets, especially those created using oversampling techniques, with Bayesian networks improved classifying a traffic accident according to its severity and reduced the misclassification of killed and severe injuries instances. On the other hand, the following variables were found to contribute to the occurrence of a killed causality or a severe injury in a traffic accident: number of vehicles involved, accident pattern, number of directions, accident type, lighting, surface condition, and speed limit. This work, to the knowledge of the authors, is the first that aims at analyzing historical data records for traffic accidents occurring in Jordan and the first to apply balancing techniques to analyze injury severity of traffic accidents. Copyright © 2015 Elsevier Ltd. All rights reserved.

  11. Analysis of the Fukushima accident impact in France (mainland and overseas territories) from the results of the reinforced environmental radioactivity monitoring; Analyse de l'impact de l'accident de Fukushima en France (metropole et DROM-COM) a partir des resultats de la surveillance renforcee de la radioactivite de l'environnement

    Energy Technology Data Exchange (ETDEWEB)

    Chaptal-Gradoz, N.

    2011-07-01

    The enhanced environmental radioactivity monitoring settled in France (and the country's overseas territories) following the Fukushima Dai-ichi accident detected trace levels of the main radionuclides released into the atmosphere during the accident (iodine 131, caesium 134 and 137, and smaller quantities of tellurium 132) and monitored their evolution until May. In mainland France, the various measurement results relating to air, rainwater and terrestrial products revealed that no traces of radionuclides attributable to the Fukushima accident were detected prior to 24 March. Airborne particulate iodine 131 was detected for the first time in a sample collected on 24 March at the summit of the Puy-de-Dome mountain, and this observation was confirmed nationwide with effect from 25 March. In the other compartments of the environment, the first significant results for iodine 131 were obtained over the following days. The maximum iodine 131 levels detected were of the order of: - a few milli-becquerels per cubic metre of air (mBq/m{sup 3}); - approximately 10 becquerels per kilogram (Bq/kg) of plant matter (grass or leafy vegetables); - a few becquerels per litre (Bq/L) in rainwater or milk. Caesium 137 (together with an equivalent level of caesium 134) was also detected, generally at concentrations ten times lower than those observed for iodine 131. Tellurium 132 was occasionally detected in the air in late March. The various observations revealed that all French regions were affected to the same extent, albeit with some geographic and chronological fluctuations due to the movements of air masses. These levels are consistent with the forecasts based on modelling by IRSN in association with the French meteorological institute, Meteo-France. These concentrations were between 500 and more than 1, 000 times lower than those measured in France in early May 1986, following the Chernobyl accident. The maximum potential exposures (thyroid equivalent dose in one year

  12. Cyclical Fluctuations in Workplace Accidents

    OpenAIRE

    Boone, J.; J. C. VAN OURS

    2002-01-01

    This Paper presents a theory and an empirical investigation on cyclical fluctuations in workplace accidents. The theory is based on the idea that reporting an accident dents the reputation of a worker and raises the probability that he is fired. Therefore a country with a high or an increasing unemployment rate has a low (reported) workplace accident rate. The empirical investigation concerns workplace accidents in OECD countries. The analysis confirms that workplace accident rates are invers...

  13. MELCOR analysis of the TMI-2 accident

    Energy Technology Data Exchange (ETDEWEB)

    Boucheron, E.A.

    1990-01-01

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

  14. Development of Highly Survivable Power and Communication System for NPP Instruments under Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Seung J.; Gu, Beom W.; Nguyen, Duy T.; Choi, Bo H.; Rim, Chun T. [KAIST, Daejeon (Korea, Republic of); Lee, So I. [KHNP CRI, Daejeon (Korea, Republic of)

    2014-10-15

    According to the detail report from the Fukushima nuclear accident, the failure of conventional instruments is mainly due to the following reasons. 1) Insufficient backup battery capacity after the station black out (SBO) 2) The malfunction or damage of instruments due to the extremely harsh ambient condition after the severe accident 3) The cut-off of power and communication cable due to the physical shocks of hydrogen explosion after the severe accident Since the current equipment qualification (EQ) for the NPP instruments is based on the design basis accident such as loss of coolant accident (LOCA), conventional instruments, which are examined under EQ condition, cannot guarantee their normal operation during the severe accident. A 7m-long-distance wireless power transfer and a radio frequency (RF) communication were introduced with conventional wired system to increase a redundancy. A heat isolation box and a harness are adopted to provide a protection from the expected physical shocks such as missiles and drastic increase of ambient temperature and pressure. A detail design principle of the highly survivable power and communication system, which has 4 sub-systems of a DCRS wireless power transfer, a Zigbee wireless communication, a GFRP harness, and a passive type router with a fly back regulator, has been presented in this paper. Each sub-system has been designed to have a robust operation characteristic regardless of the estimated physical shocks after the severe accident.

  15. Practical approaches in accident analysis

    Science.gov (United States)

    Stock, M.

    An accident analysis technique based on successive application of structural response, explosion dynamics, gas cloud formation, and plant operation failure mode models is proposed. The method takes into account the nonideal explosion characteristic of a deflagration in the unconfined cloud. The resulting pressure wave differs significantly from a shock wave and the response of structures like lamp posts and walls can differ correspondingly. This gives a more realistic insight into explosion courses than a simple TNT-equivalent approach.

  16. Characterization of mortality rates resulting from traffic accidents involving cycles Caracterización de la mortalidad por accidentes del tránsito con participación de ciclos

    Directory of Open Access Journals (Sweden)

    Mavis Ortis Sagasta

    2010-08-01

    Full Text Available Background: Traffic accidents are a highly complex phenomenon due to the variety of factors involved in their occurrence. They have become more relevant as the fight against illnesses becomes more effective. Objective: To characterize the mortality rates resulting from traffic accidents involving cycles. Methods: Descriptive study on case series conducted at the General University Hospital "Dr. Gustavo Lima Aldereguía " in Cienfuegos on all deaths related to traffic accidents involving cycles from January 1st to December 31st , 2009. Variables that were analyzed: age, sex, time of the accident, predominantly affected anatomical region, direct cause of death and ingestion of alcohol. Results: There were 26 deaths (6.4 per 100 000 inhabitants, mainly in the group from 15 to 29 years old, predominantly males with 23 cases (88.4%.53, 8% of them were workers and 23% were students. There was a predominance of fatalities in the hours between 6:00 pm and 11:59 pm. The most affected anatomic region was the head (61.5%, 46, 2 % of deaths had more than 50 mg of alcohol in blood and the predominant direct cause of death was the injuries of higher nervous centers. Conclusions: Traffic accidents involving cycles are more frequent in young males . They are related to the ingestion of alcohol and usually affect the head region.Introducción: los accidentes constituyen un fenómeno sumamente complejo debido a los numerosos y variados factores que intervienen en su producción. Estos se hacen más relevantes en la medida que se hace más eficaz la lucha contra las enfermedades. Objetivo: caracterizar la mortalidad por accidentes de tránsito con participación de ciclos. Métodos: estudio descriptivo de serie de casos, realizado en el Hospital General Universitario “Dr. Gustavo Aldereguía Lima” de Cienfuegos sobre todos los

  17. Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, Joy L. [Rempe and Associates, Idaho Falls, ID (United States); Knudson, Darrell L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken by these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.

  18. Impact on the marine environment of radioactive releases resulting from the Fukushima-Daiichi nuclear accident. April 4, 2011; Impact sur le milieu marin des rejets radioactifs consecutifs a l'accident nucleaire de Fukushima-Daiichi. 4 avril 2011

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    After having outlined that measurements taken over several days in the sea water in the vicinity of the power station have revealed severe contamination of the marine environment by various radionuclides released as a result of the accident at the Fukushima-Daiichi nuclear power station, this report describes the origins of the contamination of the marine environment (release of liquid effluents directly into the sea in the vicinity of the damaged reactors, atmospheric fallout onto the surface of the sea, conveyance of radioactive pollution by rainout of contaminated ground). Then, while proposing several maps, the authors analyse and comment the dispersion in the sea of radioactive pollutants by addressing the following issues: topography of the sea bed and sea currents off the Japanese coast, immediate or short term dispersion (a few days), mid-term dispersion (weeks and months), long term and large scale future of the radioactive pollutants. Finally, the report briefly discusses the impact of radioactive pollution on living species

  19. Design of coolant distribution system (CDS) for ITER PF AC/DC converter

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Bin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Zhiquan, E-mail: zhquansong@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Fu, Peng; Xu, Xuesong; Li, Chuan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wang, Min; Dong, Lin [China International Nuclear Fusion Energy Program Execution Center, Beijing 100862 (China)

    2016-10-15

    Highlights: • System process and arrangement has been proposed to meet the multiple requirements from the converter system. • Thermal hydraulic analysis model has been developed to size and predict the system operation behavior. • Prototype test has been performed to validate the proposed design methodology. - Abstract: The Poloidal Field (PF) converter unit, playing an essential role in the plasma shape and position control in vertical and horizontal direction, which is an important part of ITER power supply system. As an important subsystem of the converter unit, the coolant distribution system has the function to distribute the cooling water from ITER component cooling water system (CCWS) to its main components at the required flow rate, pressure and temperature. This paper presents the thermal hydraulic design of coolant distribution system for the ITER PF converter unit. Different operational requirements of the PF converter unit regarding flow rate, temperature and pressure have been analyzed to design the system process and arrangement. A thermal-hydraulic analysis model has been built to size the system and predict the flow rate and temperature distribution of the system under the normal operation. Based on the system thermal-hydraulic analysis results, the system pressure profile has been plotted to evaluate the pressure behavior along each client flow path. A CDS prototype for the ITER PF converter has been constructed and some experiments have been performed on it. A good agreement of the flow distribution and temperature behavior between the simulated and test results validate the proposed design methodology.

  20. Experimental Investigation of Coolant Boiling in a Half-Heated Circular Tube - Final CRADA Report

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Wenhua [Argonne National Lab. (ANL), Argonne, IL (United States); Singh, Dileep [Argonne National Lab. (ANL), Argonne, IL (United States); France, David M. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-11-01

    Coolant subcooled boiling in the cylinder head regions of heavy-duty vehicle engines is unavoidable at high thermal loads due to high metal temperatures. However, theoretical, numerical, and experimental studies of coolant subcooled flow boiling under these specific application conditions are generally lacking in the engineering literature. The objective of this project was to provide such much-needed information, including the coolant subcooled flow boiling characteristics and the corresponding heat transfer coefficients, through experimental investigations.

  1. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Bunt, R. [Southern Nuclear, Atlanta, GA (United States); Corradini, M. [Univ. of Wisconsin, Madison, WI (United States); Ellison, Paul B. [GE Power and Water, Duluth, GA (United States); Francis, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Gabor, John D. [Erin Engineering, Walnut Creek, CA (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Henry, C. [Fauske and Associates, Burr Ridge, IL (United States); Linthicum, R. [Exelon Corp., Chicago, IL (United States); Luangdilok, W. [Fauske and Associates, Burr Ridge, IL (United States); Lutz, R. [PWR Owners Group (PWROG); Paik, C. [Fauske and Associates, Burr Ridge, IL (United States); Plys, M. [Fauske and Associates, Burr Ridge, IL (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Robb, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wachowiak, R. [Electric Power Research Inst. (EPRI), Knovville, TN (United States)

    2015-01-31

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  2. Biomass accident investigations – missed opportunities for learning and accident prevention

    DEFF Research Database (Denmark)

    Hedlund, Frank Huess

    2017-01-01

    selected serious accidents involving biogas and wood pellets in Denmark and argues that such opportunities for learning were missed because accident investigations were superficial, follow-up incomplete and information sharing absent. In one particularly distressing case, a facility saw a repeat accident......, this time with fatal outcome, still without any learning taking place. The paper presents some information on other biomass accidents in Denmark, mostly involving biogas from anaerobic digestion. Details are lacking however, precisely because the accidents were insufficiently investigated and results...... not communicated. The biomass industry needs to pay more attention to safety. Utmost care should be taken to avoid so-called mediashifting i.e. that the resolution of a problem within one domain, the environmental, creates a new problem in another, the workplace safety domain....

  3. Effect of spray system on fission product distribution in containment during a severe accident in a two-loop pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dehjourian, Mehdi; Rahgoshay, Mohammad; Jahanfamia, Gholamreza [Dept. of Nuclear Engineering, Science and Research Branch, Islamic Azad University of Tehran, Tehran (Iran, Islamic Republic of); Sayareh, Reza [Faculty of Electrical and Computer Engineering, Kerman Graduate University of Technology, Kerman (Iran, Islamic Republic of); Shirani, Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-08-15

    The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

  4. Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Mehdi Dehjourian

    2016-08-01

    Full Text Available The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

  5. Investigating hydrodynamic characteristics and peculiarities of the coolant flow behind a spacer grid of a fuel rod assembly of the floating nuclear power unit

    Science.gov (United States)

    Dmitriev, S. M.; Doronkov, D. V.; Legchanov, M. A.; Pronin, A. N.; Solncev, D. N.; Sorokin, V. D.; Hrobostov, A. E.

    2016-05-01

    The results of experimental investigations of local hydrodynamics of a coolant flow in fuel rod assembly (FA) of KLT-40C reactor behind a plate spacer grid have been presented. The investigations were carried out on an aerodynamic rig using the gas-phase diffusive tracer test. An analysis of spatial distribution of absolute flow velocity projections and distribution of tracer concentration allowed specifying a coolant flow pattern behind the plate spacer grid of the FA. On the basis of obtained experimental data the recommendations were provided to specify procedures for determining the coolant flow rates for the programs of cell-wise calculation of a core zone of KLT-40C reactor. Investigation results were accepted for the practical use in JSC "OKBM Afrikantov" to assess heat engineering reliability of cores of KLT-40C reactor and were included in a database for verification of CFD programs (CFD-codes).

  6. Development of seismic sloshing analysis method of liquid coolant sodium in the KALIMER reactor vessel including several cylindrical components

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Yoo, Bong

    2000-11-01

    It is important to establish a highly accurate technique of evaluating the sloshing behavior of liquid sodium coolant during earthquake for structural integrity of KALIMER reactor vessel and internals. The analysis procedure of sloshing behaviors is established using finite element computer program ANSYS, and the effectiveness of the procedure is confirmed by comparison with theoretical and experimental results in the literature. The analysis results agree well with experimental ones. Based on the procedure, the sloshing characteristics of liquid sodium coolant in the KALIMER reactor vessel including reactor internal components are evaluated. The maximum response height of sodium free surface at the reactor vessel is about 55cm when subjected to horizontal safe shutdown earthquake (SSE) of 0.3g for seismically isolated reactor building.

  7. Station blackout with reactor coolant pump seal leakage

    Energy Technology Data Exchange (ETDEWEB)

    Evinay, A. (Southern California Edison, Irvine, CA (United States))

    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.

  8. Aqueous Nanofluid as a Two-Phase Coolant for PWR

    Directory of Open Access Journals (Sweden)

    Pavel N. Alekseev

    2012-01-01

    Full Text Available Density fluctuations in liquid water consist of two topological kinds of instant molecular clusters. The dense ones have helical hydrogen bonds and the nondense ones are tetrahedral clusters with ice-like hydrogen bonds of water molecules. Helical ordering of protons in the dense water clusters can participate in coherent vibrations. The ramified interface of such incompatible structural elements induces clustering impurities in any aqueous solution. These additives can enhance a heat transfer of water as a two-phase coolant for PWR due to natural forming of nanoparticles with a thermal conductivity higher than water. The aqueous nanofluid as a new condensed matter has a great potential for cooling applications. It is a mixture of liquid water and dispersed phase of extremely fine quasi-solid particles usually less than 50 nm in size with the high thermal conductivity. An alternative approach is the formation of gaseous (oxygen or hydrogen nanoparticles in density fluctuations of water. It is possible to obtain stable nanobubbles that can considerably exceed the molecular solubility of oxygen (hydrogen in water. Such a nanofluid can convert the liquid water in the nonstoichiometric state and change its reduction-oxidation (RedOx potential similarly to adding oxidants (or antioxidants for applying 2D water chemistry to aqueous coolant.

  9. Fitness for service assessment of coolant channels of Indian PHWRs

    Science.gov (United States)

    Sinha, R. K.; Sinha, S. K.; Madhusoodanan, K.

    2008-12-01

    A typical coolant channel assembly of pressurised heavy water reactors mainly consists of pressure tube, calandria tube, garter spring spacers, all made of zirconium alloys and end fittings made of SS 403. The pressure tube is rolled at both its ends to the end fittings and is located concentrically inside the calandria tube with the help of garter spring spacers. Pressure tube houses the fuel bundles, which are cooled by means of pressurised heavy water. It, thus, operates under the environment of high pressure and temperature (typically 10 MPa and 573 K), and fast neutron flux (typically 3 × 10 17 n/m 2 s, E > 1 MeV neutrons). Under this operating environment, the material of the pressure tube undergoes degradation over a period of time, and eventually needs to be assessed for fitness for continued operation, without jeopardising the safety of the reactor. The other components of the coolant channel assembly, which are inaccessible for any in-service inspection, are assessed for their fitness, whenever a pressure tube is removed for either surveillance purpose or any other reasons. This paper, while describing the latest developments taking place to address the issue of fitness for service of the Zr-2.5 wt% Nb pressure tubes, also dwells briefly upon the developments taken place, to address the issues of life management and extension of zircaloy-2 pressure tubes in the earlier generation of Indian pressurised heavy water reactors.

  10. Fitness for service assessment of coolant channels of Indian PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, R.K.; Sinha, S.K. [Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Madhusoodanan, K. [Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai 400 085 (India)], E-mail: kmadhu@barc.gov.in

    2008-12-15

    A typical coolant channel assembly of pressurised heavy water reactors mainly consists of pressure tube, calandria tube, garter spring spacers, all made of zirconium alloys and end fittings made of SS 403. The pressure tube is rolled at both its ends to the end fittings and is located concentrically inside the calandria tube with the help of garter spring spacers. Pressure tube houses the fuel bundles, which are cooled by means of pressurised heavy water. It, thus, operates under the environment of high pressure and temperature (typically 10 MPa and 573 K), and fast neutron flux (typically 3 x 10{sup 17} n/m{sup 2} s, E > 1 MeV neutrons). Under this operating environment, the material of the pressure tube undergoes degradation over a period of time, and eventually needs to be assessed for fitness for continued operation, without jeopardising the safety of the reactor. The other components of the coolant channel assembly, which are inaccessible for any in-service inspection, are assessed for their fitness, whenever a pressure tube is removed for either surveillance purpose or any other reasons. This paper, while describing the latest developments taking place to address the issue of fitness for service of the Zr-2.5 wt% Nb pressure tubes, also dwells briefly upon the developments taken place, to address the issues of life management and extension of zircaloy-2 pressure tubes in the earlier generation of Indian pressurised heavy water reactors.

  11. Liquid metal reactor development -Studies on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author).

  12. Effect of internal coolant crossflow orientation on the discharge coefficient of shaped film-cooling holes

    Energy Technology Data Exchange (ETDEWEB)

    Gritsch, M.; Saumweber, C.; Schulz, A.; Wittig, S.; Sharp, E.

    2000-01-01

    Discharge coefficients of three film-cooling hole geometries are presented over a wide range of engine like conditions. The hole geometries comprise a cylindrical hole and two holes with a diffuser-shaped exit portion (a fanshaped and a laidback fanshaped hole). For all three hole geometries the hole axis was inclined 30 deg with respect to the direction of the external (hot gas) flow. The flow conditions considered were the hot gas crossflow Mach number (up to 0.6), the coolant crossflow Mach number (up to 0.6) and the pressure ratio across the hole (up to 2). The effect of internal crossflow approach direction, perpendicular or parallel to the main flow direction, is particularly addressed in the present study. Comparison is made of the results for a parallel and perpendicular orientation, showing that the coolant crossflow orientation has a strong impact on the discharge behavior of the different hole geometries. The discharge coefficients were found to strongly depend on both hole geometry and crossflow conditions. Furthermore, the effects of internal and external crossflow on the discharge coefficients were described by means of correlations used to derive a predicting scheme for discharge coefficients. A comparison between predictions and measurements reveals the capability of the method proposed.

  13. Performance of water and diluted ethylene glycol as coolants for electronic cooling

    Directory of Open Access Journals (Sweden)

    M. Gayatri,

    2015-05-01

    Full Text Available As the number of transistors increases with new generation of microprocessor chips, the power draw and heat load to dissipate during operation increases. As a result of increasing the heat loads and heat fluxes the Conventional cooling technologies such as fan, heat sinks are unable to absorb and heat transfer excess heat dissipated by these new microprocessor. So, new technologies are needed to improve the heat removal capacity. In the present work single phase liquid cooling system with mini channel is analyzed and experimentally investigated. Mini channels are chosen as to provide higher heat transfer co-efficient than conventional channel. Copper pipes of 0.36 mm diameter are taken to fabricate heat sink and heat exchanger. A pump is used to circulate the fluid through heat sink and heat exchanger. A solid heated aluminium block to simulate heat generated electronic component is used and electrical input is supplied to the heated aluminium block and cooling system is placed over the heated block. The performance of the cooling system is analyzed from the experimental data obtained. It is experimentally observed that the mini channel liquid cooling system with water as a coolant has better performance than diluted ethylene glycol as coolant at different flow rates. The surface temperature of the heated aluminium block with convective heat transfer co-efficient is observed

  14. Determination of the {sup 129}I in primary coolant of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ke Chon; Park, Yong Joon; Song, Kyu Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-02-15

    Among the radioactive wastes generated from the nuclear power plant, a radioactive nuclide such as {sup 129}I is classified as a difficult-to-measure (DTM) nuclide, owing to its low specific activity. Therefore, the establishment of an analytical procedure, including a chemical separation for {sup 129}I as a representative DTM, becomes essential. In this report, the adsorption and recovery rate were measured by adding {sup 125}I as a radio-isotopic tracer (t1/2 = 60.14 d) to the simulation sample, in order to measure the activity concentration of {sup 129}I in a pressurized-water reactor primary coolant. The optimum condition for the maximum recovery yield of iodine on the anion exchange resins (AG1 x2, 50-100 mesh, Clform) was found to be at pH 7. In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of {sup 129}I was examined, as was the effect of {sup 3}H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous {sup 3}H presence was found with activity concentrations of {sup 3}H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 {mu}g/mL.

  15. Numerical Investigation of Urea Freezing and Melting Characteristics Using Coolant Heater

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Yeop; Kim, Nam Il; Kim, Man Young [Chounbuk Nat' l Univ., Jeonju (Korea, Republic of); Park, Yun Beom [Jeju College of Technology, Jeju (Korea, Republic of)

    2013-08-15

    UREA-SCR technology is known as one of the powerful NOx reduction systems for vehicles as well as stationary applications. For its consistent and reliable operation in vehicle applications, however, the freezing and melting of the urea solution in cold environments have to be resolved. In this study, therefore, a numerical study of three-dimensional unsteady problems was analyzed to understand the urea freezing and heating phenomena and heat transfer characteristics in terms of urea liquid volume fraction, temperature profiles, and phase change behavior in urea solutions with time by using the commercial software Fluent 6.3. As a result, it was found that the freezing phenomenon proceeds with a phase change from the tank wall to the center, whereas the melting phenomenon occurs faster in the upper part of the storage tank by natural convection and in the adjacent part of the coolant pipe than in other parts. Furthermore, approximately 190s were required to obtain 1a of urea solution using a 4-coiled coolant heater under conditions of 70 .deg. C and 200 L/h.

  16. System Assessment of Carbon Dioxide Used as Gas Oxidant and Coolant in Vanadium-Extraction Converter

    Science.gov (United States)

    Du, Wei Tong; Wang, Yu; Liang, Xiao Ping

    2017-07-01

    With the aim of reducing carbon dioxide (CO2) emissions and of using waste resources in steel plants, the use of CO2 as a gas oxidant and coolant in the converter to increase productivity and energy efficiency was investigated in this study. Experiments were performed in combination with thermodynamic theory on vanadium-extraction with CO2 and oxygen (O2) mixed injections. The results indicate that the temperature of the hot metal bath decreased as the amount of CO2 introduced into O2 increased. At an injection of 85 vol.% O2 and 15 vol.% CO2, approximately 12% of additional carbon was retained in the hot metal. Moreover, the content of vanadium trioxide in the slag was higher. In addition, the O2 consumption per ton of hot metal was reduced by 8.5% and additional chemical energy was recovered by the controlled injection of CO2 into the converter. Therefore, using CO2 as a gas coolant was conducive to vanadium extraction, and O2 consumption was reduced.

  17. System Assessment of Carbon Dioxide Used as Gas Oxidant and Coolant in Vanadium-Extraction Converter

    Science.gov (United States)

    Du, Wei Tong; Wang, Yu; Liang, Xiao Ping

    2017-10-01

    With the aim of reducing carbon dioxide (CO2) emissions and of using waste resources in steel plants, the use of CO2 as a gas oxidant and coolant in the converter to increase productivity and energy efficiency was investigated in this study. Experiments were performed in combination with thermodynamic theory on vanadium-extraction with CO2 and oxygen (O2) mixed injections. The results indicate that the temperature of the hot metal bath decreased as the amount of CO2 introduced into O2 increased. At an injection of 85 vol.% O2 and 15 vol.% CO2, approximately 12% of additional carbon was retained in the hot metal. Moreover, the content of vanadium trioxide in the slag was higher. In addition, the O2 consumption per ton of hot metal was reduced by 8.5% and additional chemical energy was recovered by the controlled injection of CO2 into the converter. Therefore, using CO2 as a gas coolant was conducive to vanadium extraction, and O2 consumption was reduced.

  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. Effect of complicated coolant flow behavior in the ABWR lower plenum on flow distribution to the core

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Shun, E-mail: swatanabe@edu.esys.tsukuba.ac.jp [University of Tsukuba, 1-1-1 Tennoudai, Tsukuba 305-8573, Ibaraki (Japan); Abe, Yutaka, E-mail: abe@kz.tsukuba.ac.jp [University of Tsukuba, 1-1-1 Tennoudai, Tsukuba 305-8573, Ibaraki (Japan); Kaneko, Akiko, E-mail: kaneko@kz.tsukuba.ac.jp [University of Tsukuba, 1-1-1 Tennoudai, Tsukuba 305-8573, Ibaraki (Japan); Watanabe, Fumitoshi, E-mail: Watanabe.Fumitoshi@tepco.co.jp [Tokyo Electric Power Company, 4-1, Egasaki, Tsurumi, Yokohama 230-8510, Kanagawa (Japan); Tezuka, Kenichi, E-mail: tezuka.kenichi@tepco.co.jp [Tokyo Electric Power Company, 4-1, Egasaki, Tsurumi, Yokohama 230-8510, Kanagawa (Japan); Hirao, Kazuki, E-mail: hirao.kazuki@tepco.co.jp [Tokyo Electric Power Company, 4-1, Egasaki, Tsurumi, Yokohama 230-8510, Kanagawa (Japan)

    2015-12-15

    Highlights: • A 1/10th model of the lower plenum was constructed to obtain the benchmark data. • Complicated flow behaviors were measured using LDV and PIV. • Flow distribution in the lower plenum became non-uniform. • Flow distribution became uniform at the core inlet through the side entry orifices. • The vortices were observed around the side entry orifices. - Abstract: One of the cost reduction strategies in nuclear power generation is the augmentation of power outputs by increasing the coolant flow rate. To achieve augmentation of the power outputs in an Advanced Boiling Water Reactor, it is extremely important to evaluate the coolant flow from the lower plenum to the core inlet, which dominates the coolant flow distribution over the rod bundle of the core. In the lower plenum, there are a number of structures, such as control rod driving mechanisms and guide tubes. In addition, core support beams and side entry orifices are placed at the core inlet. Therefore, it is estimated that the coolant flow becomes very complicated in these areas. To predict complicated coolant flow in the lower plenum and core inlet, simulation using a Computational Fluid Dynamics (CFD) code is helpful. However, previous experimental data are not enough to verify the simulation results for the intended flow channel in this study. Hence, this study focuses on establishing a benchmark for the CFD code using the visualization method. Based on the validation process, the functions of the complicated flow structure at the core inlet and through the lower plenum are demonstrated in detail in the coolant flow distribution to the core. A 1/10th model of a lower plenum was constructed, and the velocity profiles were measured using Particle Image Velocimetry (PIV) and Laser-Doppler Velocimetry (LDV). Each measurement was performed for a Reynolds number of 2620. From the flow velocity measurement, the vertical velocity at the center of the lower plenum is determined to be faster than

  20. FUNDING OF ACCIDENT INSURANCE IN UKRAINE

    Directory of Open Access Journals (Sweden)

    O. Gamankova

    2016-03-01

    Full Text Available The paper deals with the peculiarities of financial provision of public and private accident insurance. Analyzes the methodology of forming insurance premium rates in private accident insurance. The study examines the practice of reforming the financial security of the state social insurance against accidents. The results show need to implement scientifically proven approach to determining premium rates in the state social insurance based on mathematical statistics and actuarial calculations to ensure that such conditions on the one hand, can insure the risk, and the other - to provide the insurer the ability to perform insurance obligations' commitments. One of the promising areas of improvement Accident Insurance determines to define increasing the role of the private sector to create and attraction of investments, financing, reduce costs, and provision of insurance services on insurance against accidents at a qualitatively higher level. The results show the need to consider the usefulness of a mixed provision of services to accident insurance under state or non-state system, and the viability and effectiveness of the combination of these systems.

  1. UO2 and PuO2 utilization in high temperature engineering test reactor with helium coolant

    Science.gov (United States)

    Waris, Abdul; Aji, Indarta K.; Novitrian, Pramuditya, Syeilendra; Su'ud, Zaki

    2016-03-01

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO2 fuel. In this study, we have evaluated the use of UO2 and PuO2 in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of 235U in loaded fuel is 18.0% or above.

  2. An Analysis of an Automatic Coolant Bypass in the International Space Station Node 2 Internal Active Thermal Control System

    Science.gov (United States)

    Clanton, Stephen E.; Holt, James M.; Turner, Larry D. (Technical Monitor)

    2001-01-01

    A challenging part of International Space Station (ISS) thermal control design is the ability to incorporate design changes into an integrated system without negatively impacting performance. The challenge presents itself in that the typical ISS Internal Active Thermal Control System (IATCS) consists of an integrated hardware/software system that provides active coolant resources to a variety of users. Software algorithms control the IATCS to specific temperatures, flow rates, and pressure differentials in order to meet the user-defined requirements. What may seem to be small design changes imposed on the system may in fact result in system instability or the temporary inability to meet user requirements. The purpose of this paper is to provide a brief description of the solution process and analyses used to implement one such design change that required the incorporation of an automatic coolant bypass in the ISS Node 2 element.

  3. Assessment of Loads and Performance of a Containment in a Hypothetical Accident (ALPHA). Facility design report

    Energy Technology Data Exchange (ETDEWEB)

    Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Moriyama, Kiyofumi; Ito, Hideo; Komori, Keiichi; Sonobe, Hisao; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-06-01

    In the ALPHA (Assessment of Loads and Performance of Containment in Hypothetical Accident) program, several tests have been performed to quantitatively evaluate loads to and performance of a containment vessel during a severe accident of a light water reactor. The ALPHA program focuses on investigating leak behavior through the containment vessel, fuel-coolant interaction, molten core-concrete interaction and FP aerosol behavior, which are generally recognized as significant phenomena considered to occur in the containment. In designing the experimental facility, it was considered to simulate appropriately the phenomena mentioned above, and to cover experimental conditions not covered by previous works involving high pressure and temperature. Experiments from the viewpoint of accident management were also included in the scope. The present report describes design specifications, dimensions, instrumentation of the ALPHA facility based on the specific test objectives and procedures. (author)

  4. ASTEC V2 severe accident integral code: Fission product modelling and validation

    Energy Technology Data Exchange (ETDEWEB)

    Cantrel, L., E-mail: laurent.cantrel@irsn.fr; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

    2014-06-01

    One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology. ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase. In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R and D knowledge. This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix.

  5. The consequences of Chernobyl accident

    Directory of Open Access Journals (Sweden)

    Ion Chioșilă

    2016-12-01

    Full Text Available These days marks 30 years since the Chernobyl nuclear accident, followed by massive radioactive contamination of the environment and human in Belarus, Ukraine and Russia, and resulted in many deaths among people who intervened to decrease the effects of the nuclear disaster. The 26 April 1986 nuclear accident contaminated all European countries, but at a much lower level, without highlighted consequences on human health. In special laboratories, the main radionuclides (I-131, Cs-137, Cs-134 and Sr-90 were also analyzed in Romania from environmental samples, food, even human subjects. These radionuclides caused the population to receive a low dose of about 1 mSv in 1986 that is half of the dose of the natural background radiation (2.4 mSv per year. As in all European countries (excluding Ukraine, Belarus and Russia this dose of about 1 mSv fell rapidly by 1990, reaching levels close to ones before the accident at the nuclear tests.

  6. Medical consequences of Chernobyl accident

    Directory of Open Access Journals (Sweden)

    Galstyan I.A.

    2015-12-01

    Full Text Available Aim: to study the long-term effects of acute radiation syndrome (ARS, developed at the victims of the Chernobyl accident. Material and Methods. 237 people were exposed during the accident, 134 of them were diagnosed with ARS. Dynamic observation implies a thorough annual examination in a hospital. Results. In the first 1.5-2 years after the ARS mean group indices of peripheral blood have returned to normal. However, many patients had transient expressed moderate cytopenias. Granulocytopenia, thrombocytopenia, lymphopenia and erythropenia were the most frequently observed things during the first 5 years after the accident. After 5 years their occurences lowered. In 11 patients the radiation cataract was detected. A threshold dose for its development is a dose of 3.2 Gy Long-term effects of local radiation lesions (LRL range from mild skin figure smoothing to a distinct fibrous scarring, contrac