WorldWideScience

Sample records for system leakage accident

  1. Early localization of containment leakage during an accident

    International Nuclear Information System (INIS)

    Pepin, P.; Chauliac, C.; Libmann, M.; Martinez, J.M.

    1990-01-01

    In case of an accident in a nuclear plant, checking the containment leaktightness would be a fundamental step for the diagnosis and prognosis of the radiological consequences. Significant help in this task can be provided by softwares. For that purpose, the French Atomic Energy Commission (CEA) is developing an expert system which can provide early in the accident a classification of the possible leakage paths and help understanding the necessary corrections which have to be undertaken by the utility. This software will be used at the Emergency Technical Center of the CEA. Its basic principles are described in this paper

  2. Effects of secondary containment air cleanup system leakage on the accident offsite dose as determined during preop tests of the Sequoyah Nuclear Plant

    International Nuclear Information System (INIS)

    Klaes, L.J.; Nass, S.A.; Proctor, L.D.

    1981-01-01

    The Sequoyah Nuclear Plant has two secondary containments. One is the annular region between the primary containment and the shield building surrounding the primary containment. The second is the auxiliary building secondary containment enclosure which is potentially subject to direct airborne radioactivity. Two air cleanup systems are provided to serve these areas. The emergency gas treatment system (EGTS) serves the annulus between the primary containment and the shield building, and the auxiliary building gas treatment system (ABGTS) serves the area inside of the auxiliary building secondary containment enclosure. The major function served by these air cleanup systems is that of controlling and processing airborne contamination released in these areas during any accident up to a design basis accident. This is accomplished by (1) creating a negative pressure in the areas served to ensure that no unprocessed air is released to the atmosphere, (2) providing filtration units to process all air exhausted from the secondary containment spaces, and (3) providing a low-leakage enclosure to limit exhaust flows. Offsite dose effects due to secondary containment release rates, bypass leakage, and duct and damper leakages are presented and parameter variations are considered. For the EGTS, a recirculation system, the most important parameter is the total inleakage of the system which causes an increase in both whole body (gamma) and thyroid (iodine) doses. For the ABGTS, a once-through system, the most important paramter is the inleakage which bypasses the filters resulting in an increase in the thyroid dose only. Actual preoperational test data are utilized. Problems encountered during the preop test are summarized. Solutions incorporated to bring the EGTS and ABGTS air cleanup systems within the test acceptance criteria required to meet offsite dose limitations are discussed and the resultant calculated offsite dose is presented

  3. Leakage potential through mechanical penetrations in a severe accident environment

    International Nuclear Information System (INIS)

    Koenig, L.N.

    1986-01-01

    This paper reviews the findings of an ongoing program, Integrity of Containment Penetrations Under Severe Accident Loads. The program is concerned with the leakage modes as well as the magnitude of leakage through mechanical penetrations in a containment building subject to a severe accident. Seal and gasket tests are used to evaluate the effect of radiation aging, thermal aging, seal geometry, and seal squeeze on seals and gaskets subjected to a hypothesized severe accident. The effects on leakage of the structural response of equipment hatches, personnel airlocks, and drywell heads subjected to severe accident pressures are studied by experiments and analyses. The data gathered during this program will be used to develop methodologies for predicting leakage

  4. Leakage resilient password systems

    CERN Document Server

    Li, Yingjiu; Deng, Robert H

    2015-01-01

    This book investigates tradeoff between security and usability in designing leakage resilient password systems (LRP) and introduces two practical LRP systems named Cover Pad and ShadowKey. It demonstrates that existing LRP systems are subject to both brute force attacks and statistical attacks and that these attacks cannot be effectively mitigated without sacrificing the usability of LRP systems. Quantitative analysis proves that a secure LRP system in practical settings imposes a considerable amount of cognitive workload unless certain secure channels are involved. The book introduces a secur

  5. New system technologies implemented at Kozloduy 3 and 4 (WWER 440-230) for containment leakage and H2 control in severe accident situations - Design, qualification, installation, commissioning

    International Nuclear Information System (INIS)

    Feuerbach, R.; Eckardt, B.; Kastner, B.

    2005-01-01

    In order to reduce the residual risk associated with hypothetical severe nuclear accidents, systems and components for filtered containment venting and H 2 reduction were developed. During severe accident scenarios large quantities of hydrogen and radioactive material may be released into the containment atmosphere within a short period of time. In the event of internal over pressurization due to hypothetical severe accident sequences a pressure barrier system has to be created to confine the activity in the containment. Unavoidable releases of activity to the environment have to be minimized to a great extent as possible. Research into the hypothetical event of core melt accidents has continued and new accident mitigation technologies have been developed. Decisions have been taken to implement these new mitigation measures in operating nuclear power plants to mitigate severe accidents consequences. In order to prevent loss of containment integrity as a result of over pressurization, nuclear power plants in the Federal Republic of Germany as well as in most other European countries have been or will be back-fitted with systems for filtered venting of the containment atmosphere and systems for H 2 -control. Similar technologies for containment venting system and H 2 control have been now implemented in the first WWER 440-230 units of Kozloduy 3 and 4. Following OECD recommendations sever accident situations were analyzed and a design of countermeasures have been performed. Main goal of the developed countermeasures was to overcome the WWER 440-230 containment design specifics like, leakage rate behavior, limited available containment volume combined with the feature of high availability of electrical supply at multiple plant sites. Further more the design of counter measures considers the common use for Kozloduy unit 3 and 4. The analysis of postulated severs accident situation - without countermeasures - showed significant increase of H 2 /O 2 concentration in the

  6. Radioactivity leakage monitoring system

    International Nuclear Information System (INIS)

    Nakajima, Takuichiro; Noguchi, Noboru.

    1982-01-01

    Purpose: To obtain a device for detecting the leakage ratio of a primary coolant by utilizing the variation in the radioactivity concentration in a reactor container when the coolant is leaked. Constitution: A measurement signal is produced from a radioactivity measuring instrument, and is continuously input to a malfunction discriminator. The discriminator inputs a measurement signal to a concentration variation discriminator when the malfunction is recognized and simultaneously inputs a measurement starting time from the inputting time to a concentration measuring instrument. On the other hand, reactor water radioactivity concentration data obtained by sampling the primary coolant is input to a concentration variation computing device. A comparator obtains the ratio of the measurement signal from the measuring instrument and the computed data signal from the computing device at the same time and hence the leakage rate, indicates the average leakage rate by averaging the leakage rate signals and also indicates the total leakage amount. (Yoshihara, H.)

  7. Support calculations for management of PRISE leakage accidents

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  8. Support calculations for management of PRISE leakage accidents

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  9. Investigation of primary-to-secondary leakage accident on the PSB-VVER integral test facility

    International Nuclear Information System (INIS)

    Lipatov, I.A.; Dremin, G.I.; Galtchanskaya, S.A.; Chmal, I.I.; Moloshnikov, A.S.; Gorbunov, Y.S.; Antonova, A.I.; Elkin, I.V.

    2001-01-01

    The full text follows. The paper presents the main results from the test on primary-to-secondary leakage of 100 mm in equivalent diameter. The test was performed on the PSB-VVER integral test facility. PSB-VVER is a 4-loops scaled down model of primary system of NPP with VVER-1000 Russian type reactor. Volume - power scale is about 1/300 while elevation scale is 1/1. All components of the primary system of the reference NPP are modeled on PSB-VVER. Both passive (accumulators) and active (high and low pressure) ECCSs, pressurizer spray and relief circuits, feed water system and atmospheric dumping system (ADS) as well as the primary circuit gas remove emergency system are also simulated. The primary-to-secondary leakage was simulated using an external break line which connects the upper part of the hot header to SG water volume. The break line included a break nozzle (a cylindrical channel d = 5.8 mm, l/d = 10 with sharp inlet edge), quick-acting valve and two-phase mass flow rate measurement system. In addition loss of off-site power at the moment when a scram-signal is generated was assumed in the experiment. Thus the accident is to be considered as a beyond-design-basic one. The loss of off-site power results in the following: -main circulation pump shutdown; -pressurizer heaters switching off; -HPIS water cooling flow rate and number of points of water injection are reduced The study focuses on the adequacy of the associated accident management (AM) procedure developed by EDO ''GIDROPRESS'' as a General Designer of VVER-type reactors. The AM-procedure was adopted to the PSB-VVER test facility conditions using CATHARE (France) and DINAMIKA (Russia) codes analysis. The AM-procedure in PSB-VVER is as follows: after about 30 min of the onset of the accident, when the accident type and the localization of the SG affected become evident for the operator, he closes all the main steam isolation valves, inhibits the ADS actuation in the affected SG and begins to remove

  10. Investigation of primary-to-secondary leakage accident on the PSB-VVER integral test facility

    Energy Technology Data Exchange (ETDEWEB)

    Lipatov, I.A.; Dremin, G.I.; Galtchanskaya, S.A.; Chmal, I.I.; Moloshnikov, A.S.; Gorbunov, Y.S.; Antonova, A.I. [Electrogorsk Research and Engineering Center, EREC, Moscow (Russian Federation); Elkin, I.V. [RRC ' ' Kurchatov Institute, Moscow (Russian Federation)

    2001-07-01

    The full text follows. The paper presents the main results from the test on primary-to-secondary leakage of 100 mm in equivalent diameter. The test was performed on the PSB-VVER integral test facility. PSB-VVER is a 4-loops scaled down model of primary system of NPP with VVER-1000 Russian type reactor. Volume - power scale is about 1/300 while elevation scale is 1/1. All components of the primary system of the reference NPP are modeled on PSB-VVER. Both passive (accumulators) and active (high and low pressure) ECCSs, pressurizer spray and relief circuits, feed water system and atmospheric dumping system (ADS) as well as the primary circuit gas remove emergency system are also simulated. The primary-to-secondary leakage was simulated using an external break line which connects the upper part of the hot header to SG water volume. The break line included a break nozzle (a cylindrical channel d = 5.8 mm, l/d = 10 with sharp inlet edge), quick-acting valve and two-phase mass flow rate measurement system. In addition loss of off-site power at the moment when a scram-signal is generated was assumed in the experiment. Thus the accident is to be considered as a beyond-design-basic one. The loss of off-site power results in the following: -main circulation pump shutdown; -pressurizer heaters switching off; -HPIS water cooling flow rate and number of points of water injection are reduced The study focuses on the adequacy of the associated accident management (AM) procedure developed by EDO ''GIDROPRESS'' as a General Designer of VVER-type reactors. The AM-procedure was adopted to the PSB-VVER test facility conditions using CATHARE (France) and DINAMIKA (Russia) codes analysis. The AM-procedure in PSB-VVER is as follows: after about 30 min of the onset of the accident, when the accident type and the localization of the SG affected become evident for the operator, he closes all the main steam isolation valves, inhibits the ADS actuation in the affected SG

  11. Assessment of the potential consequences of a large primary to secondary leakage accident. Final report

    International Nuclear Information System (INIS)

    D'Auria, F.S.; Sartmadjiev, A.; Spalj, S.; Macek, J.; Kantee, H.; Elter, J.; Kostka, P.; Bukin, N.; Alexandrov, A.G.; Kristof, M.; Kvizda, B.; Matejovic, P.; Makihara, Y.

    2006-01-01

    The present paper discusses one of the IAEA's Coordinated Research Projects (CRPs). The CRP was started in 2003 to evaluate complex phenomena of primary to secondary leakage (PRISE) accidents for WWER-440 reactors. The first Research Coordination Meeting (RCM), held in March 2003, identified the possible consequences of PRISE accidents (radioactive release to the atmosphere, pressurized thermal shock, boron dilution, loss of integrity of secondary systems and severe accidents) and designated six task groups to evaluate these, as well as uncertainties associated with PRISE analyses. The second RCM, held in March 2004, discussed the preliminary results of each task group and addressed the main safety concerns related to PRISE phenomena as well as providing recommendations on modelling for PRISE analyses and on operator actions. The third RCM, held in March 2005, discussed the results of the work performed in 2004. The CRP was concluded in 2005. Publication of the final results of the CRP is planned as an IAEA TECDOC. The paper provides a review of the final results of the project. (author)

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

    International Nuclear Information System (INIS)

    1981-01-01

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

  13. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  14. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  15. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

    The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

  16. Leakage detection system in nuclear reactor container

    International Nuclear Information System (INIS)

    Kurosawa, Masahiko.

    1993-01-01

    The present invention comprises an injection means for adding radioactive materials to coolants in a container cooler, a gamma ray amplitude analyzer connected to coolant pipelines and a means for recording/transferring the data of the result of the measurement, a gamma ray amplitude analyzer connected to a drain water sump and a means for recording/transferring the data of the result of the measurement, a gamma ray amplitude analyzer connected to various kinds of pipelines and a means for recording/transferring the data of the result of the measurement, and a data processing means for comparing and analyzing the measured data of each of the gamma ray amplitude analyzers inputted from each of date recording/transferring means. The gamma ray amplitude analysis for each of the pipelines and drain water sump are conducted at an appropriate frequency, and the measured data are compared and analyzed, to improve the detection accuracy for a trace amount of leakage from each of the pressure pipelines and the container cooler coolant pipelines, thereby enabling to specify the pipeline having leakage. Maintenance efficiency is improved, and severe rupture accident in each of pressure pipelines is prevented previously. (N.H.)

  17. Testing to determine the leakage behavior of inflatable seals subject to severe accident loadings

    International Nuclear Information System (INIS)

    Parks, M.B.

    1988-01-01

    Under the sponsorship of the United States Nuclear Regulatory Commission, Sandia National Laboratories is currently developing test validated methods to predict the pressure capacity, at elevated temperatures, of light water reactor (LWR) nuclear containment vessels subject to loads well beyond their design basis - the so-called severe accident. Scale model tests of containments with the major penetrations represented have been carried to functional failure by internal pressurization. Also, combined pressure and elevated temperature tests of typical compression seals and gaskets, a full size personnel airlock, and of typical electrical penetration assemblies (EPAs), have been conducted in order to better understand the leakage behavior of containment penetrations. Because inflatable seals are also a part of the pressure boundary of some containments, it is important to understand their leakage behavior as well. This paper discusses the results of tests that were performed to better define the leakage behavior of inflatable seals when subjected to loads well beyond their design basis

  18. Evaluation of the leakage behavior of inflatable seals subject to severe accident conditions

    International Nuclear Information System (INIS)

    Parks, M.B.

    1989-11-01

    Sandia National Laboratories, under the sponsorship of the United States Nuclear Regulatory Commission, is currently developing test validated methods to predict the pressure capacity of light water reactor containment buildings when subjected to postulated severe accident conditions. These conditions are well beyond the design basis. Scale model tests of steel and reinforced concrete containments have been conducted as well as tests of typical containment penetrations. As a part of this effort, a series of tests was recently conducted to determine the leakage behavior of inflatable seals. These seals are used to prevent leakage around personnel and escape lock doors of some containments. The results of the inflatable seals tests are the subject of this report. Inflatable seals were tested at both room temperature and at elevated temperatures representative of postulated severe accident conditions. Both aged (radiation and thermal) and unaged seals were included in the test program. The internal seal pressure at the beginning of each test was varied to cover the range of seal pressures actually used in containments. For each seal pressure level, the external (containment) pressure was increased until significant leakage past the seals was observed. Parameters that were monitored and recorded during the tests were the internal seal pressure, chamber pressure, leakage past the seals, and temperature of the test chamber and fixture to which the seals were attached. 8 refs., 34 figs., 7 tabs

  19. Criticality accident alarm system

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1991-01-01

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

  20. Effects of sodium fires on structures and materials. Practical experience with sodium leakage accidents

    International Nuclear Information System (INIS)

    Freudenstein, K.F.

    1989-01-01

    A few sodium leakage, incidents happened in SNR 300 nuclear power plant during pre-nuclear operation which were of minor importance with respect to sodium fires. The most important sodium fire accident in the past happened in the Almeria Solar platform in Spain during the attempt to repair a valve while leaving accidentally the circuit under 4 bar overpressure. Considerable damage to pipes, valves, its insulation and its support structures was observed in the influence zone of the fire. Post accident analysis gave a leaked mass of about 14 m 3 , at a sodium temperature of 225 deg. C, the leakage lasting approximately half an hour, and burning under convective heat exchange with the external air in a section of 40 m 2 up to a height of 6 m down to the catch pans. Some local temperatures were determined by metallurgical means, integral support temperatures estimated from mechanical deformation observed. From these temperatures it was concluded that a massive spray type fire must have happened. The results fall in the interpretation range of sodium-spray fire test results. (author)

  1. Secret rate - Privacy leakage in biometric systems

    NARCIS (Netherlands)

    Ignatenko, T.; Willems, F.M.J.

    2009-01-01

    Ahlswede and Csiszár [1993] introduced the concept of secret sharing. In their source model two terminals observe two correlated sequences. It is the objective of the terminals to form a common secret by interchanging a public message (helper data) in such a way that the secrecy leakage is

  2. A review on leakage rate tests for containment isolation systems

    International Nuclear Information System (INIS)

    Kim, In Goo; Kim, Hho Jung

    1992-01-01

    Wide experiences in operating containment isolation systems have been accumulated in Korea since 1978. Hence, it becomes necessary to review the operating data in order to confirm the integrity of containments with about 50 reactor-years of experience and to establish the future direction to the containment test program. The objectives of present work are to collect, consolidate and assess the leakage rate data, and then to find out dominant leakage paths and factors affecting integrated leakage rate test. General trends of overall leakage show that more careful surveillance during pre-operational test can reduce the containment leakage. Dominant leakage paths are found to be through air locks and large-sized valves, such as butterfly valves of purge lines, so that weighted surveillance and inspection on these dominant leakage paths can considerably reduce the containment leakage. The atmosphere stabilization are found to be the most important to obtain the reliable result. In order to get well stabilized atmosphere, temperature and flow rate of compressed air should be kept constant and it is preferable not to operate fan cooler during pressurizing the containment for test

  3. Leakage Testing for Different Adhesive Systems and Composites to ...

    African Journals Online (AJOL)

    2015-11-16

    Nov 16, 2015 ... resin composite, the fifth group – two‑stage SE adhesive applied and cavities filled with ... KEYWORDS: Adhesives, composite, evaluation, leakage ... the glass ionomers. ... systems are realized in one or two clinical step(s).[5].

  4. Smart integrated containment leakage rate test system using wireless communication

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Hwan; Lee, Sang Yong; Kim, Jung Sun; Kim, Gun Soo; Kim, Jong Myeong; Ahn, Jong Han [Research and Development Center, Ulsan (Korea, Republic of)

    2012-10-15

    Integrated Leakage Rate Test (ILRT) is the important test the confidentiality and integrity of the containment building, which is the last barrier when Design basis accidents (DBA) of Nuclear Power plant occur. Since the result of this test is the basis to guarantee the safety of nuclear power plants, the test process, test procedure, and the test equipment are required to have high reliability. The test devices previously used have been products of VOLUMERTRICS and GRAFTEL of USA. These devices have been inconvenient to calibrate and use. Thus improved devices needed to be developed to remove the inconveniences, to verify the safety of Korean nuclear power plants with Korea's own technology, and to secure core technology. A new leak test system was developed by domestic technology for that purpose and needed to be verified. In this paper, technical details of the newly developed easy to use and highly reliable measuring test device, which is in operation at the nuclear power plant sites, will be introduced. State of art technology was applied to the device to address the shortcomings of previous US made devices and the difficulties to use on site.

  5. The application of FLUENT in simulating outcomes from chlorine leakage accidents in a typical chemical factory.

    Science.gov (United States)

    Li, Jianfeng; Zhang, Bin; Tang, Sichuang; Tong, Ruipeng

    2016-05-01

    For improvements in market competitiveness, old brand chemical enterprises did some expansion and reconstruction on the base of original equipment. Because it is the reconstruction on the basis of the existing production equipment, it is bound to raise problems of reutilization existing in pipelines and equipment. A simplified typical chemical factory was established referring the actual workshop layout. Further, trustable accident scenarios were conducted to reveal the diffusion process. In a larger leakage rate, the chlorine leak-affected area in the downwind became larger a bit, also in a relatively shorter time, lethal scope will become larger quickly, resulting in more threats to the lives and properties in the vicinity of the factories. Further, it is not possible that the heavier-than-air effect of the chlorine will inevitably result in a higher concentration for a lower surface than that of higher surface. Actually at a certain height, a relatively higher monitoring surface has a larger diffusion range and a larger concentration than a relatively lower surface. It can be inferred that within a certain height, chlorine diffusion rate closer to the ground would be slower due to existence of turbulence or the relative resistance on the ground. © The Author(s) 2014.

  6. Leakage warning system for flexible underwater pipelines

    Energy Technology Data Exchange (ETDEWEB)

    Moeller, E; Bernstein, L

    1985-08-01

    Underwater pipelines for unloading oil tankers, e.g. in 30 km distance from the harbour site, are required to be flexible and require supervision. This is done by implementation of oil sensitive sensors between the inner rubber tube and the following impregnated textile layer. The generated sensor signals, influenced by leak oil, have to be wireless transmitted from 150 meters under water to the supervisory station at the coast. Sensor configurations are described, to derive the point of the leakage from the topologized warning signals.

  7. N13 - based reactor coolant pressure boundary leakage system

    International Nuclear Information System (INIS)

    Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

    1980-05-01

    A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

  8. Benchmarking leakage from water reticulation systems in South Africa

    African Journals Online (AJOL)

    driniev

    Abstract. A project to assess the levels of leakage in 30 water utilities throughout South Africa was initiated by the Water ... average operating pressure, systems input volume and the compo- ... simple, user-friendly model that is based on an excel spreadsheet ..... One can see from the four graphs presented here that various.

  9. Development of Wireless System for Containment Integrated Leakage Rate Test

    International Nuclear Information System (INIS)

    Lee, Kwang-Dae; Oh, Eung-Se; Yang, Seung-Ok

    2006-01-01

    The containment system leakage rate should be estimated periodically with reliable test equipment. In light-water reactor nuclear power plants, ANSI/ANS- 56.8 is a basis for determining leakage rates. Two types of data acquisition system, centralized type and networked type, has been used. In centralized type, all sensors are connected directly from sensors in the containment to the measuring equipment outside the building. The other hand, the networked type has several branch chains which connect one group of the network-sensors together. To test leakage rate, more than 20 temperature sensors and 6 humidity sensors, which are different for each plant, should be installed on a specific level in the containment. A wireless technology gives the benefits such as reducing installation efforts, making pretest easy, so it is widely used more and more in the plant monitoring. As the containment system has many kinds of complex barriers to the radio frequency, the radio power and frequency band for better transmission rate as well as the interference by the radio frequency should be considered. The overview of the wireless sensor system for the containment leakage rate test is described here and the test results on Yonggwang unit 4 PWR plant is presented

  10. 49 CFR 192.723 - Distribution systems: Leakage surveys.

    Science.gov (United States)

    2010-10-01

    ... following minimum requirements: (1) A leakage survey with leak detector equipment must be conducted in business districts, including tests of the atmosphere in gas, electric, telephone, sewer, and water system... survey with leak detector equipment must be conducted outside business districts as frequently as...

  11. Optimized quantization in Zero Leakage Helper data systems

    NARCIS (Netherlands)

    Stanko, T.; Andini, F.N.; Skoric, B.

    2017-01-01

    Helper data systems are a cryptographic primitive that allows for the reproducible extraction of secrets from noisy measurements. Redundancy data called helper data makes it possible to do error correction while leaking little or nothing (Zero Leakage) about the extracted secret string. We study the

  12. Noble gas control room accident filtration system for severe accident conditions N-CRAFT. System design

    International Nuclear Information System (INIS)

    Hill, Axel

    2014-01-01

    Severe accidents might cause the release of airborne radioactive substances to the environment of the NPP. This can either be due to leakages of the containment or due to a filtered containment venting in order to ensure the overall integrity of the containment. During the containment venting process aerosols and iodine can be retained by the FCVS which prevents long term ground contamination. Noble gases are not retainable by the FCVS. From this it follows that a large amount of radioactive noble gases (e.g. xenon, krypton) might be present in the nearby environment of the plant dominating the activity release, depending on the venting procedure and the weather conditions. Accident management measures are necessary in case of severe accidents and the prolonged stay of staff inside the main control room (MCR) or emergency response center (ERC) is essential. Therefore, the in leakage and contamination of the MRC and ERC with airborne activity has to be prevented. The radiation exposure of the crises team needs to be minimized. The entrance of noble gases cannot be sufficiently prevented by the conventional air filtration systems such as HEPA filters and iodine absorbers. With the objective to prevent an unacceptable contamination of the MCR/ERC atmosphere by noble gases AREVA GmbH has developed a noble gas retention system. The noble gas control room accident filtration system CRAFT is designed for this case and provides supply of fresh air to the MCR/ERC without time limitation. The retention process of the system is based on the dynamic adsorption of noble gases on activated carbon. The system consists of delay lines (carbon columns) which are operated by a continuous and simultaneous adsorption and desorption process. These cycles ensure a periodic load and flushing of the delay lines retaining the noble gases from entering the MCR. CRAFT allows a minimization of the dose rate inside MCR/ERC and ensures a low radiation exposure to the staff on shift maintaining

  13. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1997-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  14. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1998-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  15. A decision support system for on-line leakage localization

    OpenAIRE

    Meseguer, Jordi; Mirats-Tur, Josep M.; Cembrano, Gabriela; Puig, Vicenç; Quevedo, Joseba; Pérez, Ramon; Sanz, Gerard; Ibarra, David

    2014-01-01

    This paper describes a model-driven decision-support system (software tool) implementing a model-based methodology for on-line leakage detection and localization which is useful for a large class of water distribution networks. Since these methods present a certain degree of complexity which limits their use to experts, the proposed software tool focuses on the integration of a method emphasizing its use by water network managers as a decision support system. The proposed software tool integr...

  16. Radiation leakage dose from Elekta electron collimation system.

    Science.gov (United States)

    Pitcher, Garrett M; Hogstrom, Kenneth R; Carver, Robert L

    2016-09-08

    This study provided baseline data required for a greater project, whose objective was to design a new Elekta electron collimation system having significantly lighter electron applicators with equally low out-of field leakage dose. Specifically, off-axis dose profiles for the electron collimation system of our uniquely configured Elekta Infinity accelerator with the MLCi2 treatment head were measured and calculated for two primary purposes: 1) to evaluate and document the out-of-field leakage dose in the patient plane and 2) to validate the dose distributions calculated using a BEAMnrc Monte Carlo (MC) model for out-of-field dose profiles. Off-axis dose profiles were measured in a water phantom at 100 cm SSD for 1 and 2 cm depths along the in-plane, cross-plane, and both diagonal axes using a cylindrical ionization chamber with the 10 × 10 and 20 × 20 cm2 applicators and 7, 13, and 20 MeV beams. Dose distributions were calculated using a previously developed BEAMnrc MC model of the Elekta Infinity accelerator for the same beam energies and applicator sizes and compared with measurements. Measured results showed that the in-field beam flatness met our acceptance criteria (± 3% on major and ±4% on diagonal axes) and that out-of-field mean and maximum percent leakage doses in the patient plane met acceptance criteria as specified by the International Electrotechnical Commission (IEC). Cross-plane out-of-field dose profiles showed greater leakage dose than in-plane profiles, attributed to the curved edges of the upper X-ray jaws and multileaf collimator. Mean leakage doses increased with beam energy, being 0.93% and 0.85% of maximum central axis dose for the 10 × 10 and 20 × 20 cm2 applicators, respectively, at 20 MeV. MC calculations predicted the measured dose to within 0.1% in most profiles outside the radiation field; however, excluding model-ing of nontrimmer applicator components led to calculations exceeding measured data by as much as 0.2% for some regions

  17. Modern diagnostic systems for loose parts, vibration and leakage monitoring

    International Nuclear Information System (INIS)

    Kunze, U.

    1997-01-01

    The modern diagnostic systems for loose parts, vibration and leakage monitoring of Siemens marked improvements in signal detection, ease of operation, and the display of information. The paper gives an overview on: Loose parts monitoring system KUeS '95 - a computer-based system. The knowledge and experience about loose parts detection incorporated into this system can be characterized as ''intelligence''. Vibration monitoring system SUeS '95 - a fully automated system for early detection of changes in the vibration patterns of the reactor coolant system components and reactor pressure vessel internals. Leak detection system FLUeS - a system that detects even small leaks in steam-carrying components and very accurately determines their location. Leaks are detected on the moisture distribution in a sample air column into which the escaping steam locally diffuses. All systems described represent the latest state of technology. Nevertheless a considerable amount of operational experience can be reported. (author). 5 refs, 10 figs

  18. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    Energy Technology Data Exchange (ETDEWEB)

    Bakalov, Ivan [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Berlin (Germany); Sonnenkalb, Martin [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Koeln (Germany)

    2018-02-15

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

  19. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    International Nuclear Information System (INIS)

    Bakalov, Ivan; Sonnenkalb, Martin

    2018-01-01

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

  20. Distributed fiber optic system for oil pipeline leakage detection

    Science.gov (United States)

    Paranjape, R.; Liu, N.; Rumple, C.; Hara, Elmer H.

    2003-02-01

    We present a novel approach for the detection of leakage in oil pipelines using methods of fiber optic distributed sensors, a presence-of-oil based actuator, and Optical Time Domain Reflectometry (OTDR). While the basic concepts of our approach are well understood, the integration of the components into a complete system is a real world engineering design problem. Our focus has been on the development of the actuator design and testing using installed dark fiber. Initial results are promising, however environmental studies into the long term effects of exposure to the environment are still pending.

  1. A Labview Based Leakage Current Monitoring System For HV Insulators

    Directory of Open Access Journals (Sweden)

    N. Mavrikakis

    2015-10-01

    Full Text Available In this paper, a Labview based leakage current monitoring system for High Voltage insulators is described. The system uses a general purpose DAQ system with the addition of different current sensors. The DAQ system consists of a chassis and hot-swappable modules. Through the proper design of current sensors, low cost modules operating with a suitable input range can be employed. Fully customizable software can be developed using Labview, allowing on-demand changes and incorporation of upgrades. Such a system provides a low cost alternative to specially designed equipment with the added advantage of maximum flexibility. Further, it can be modified to satisfy the specifications (technical and economical set under different scenarios. In fact, the system described in this paper has already been installed in the HV Lab of the TEI of Crete whereas a variation of it is currently in use in TALOS High Voltage Test Station.

  2. Aerospace Accident - Injury Autopsy Data System -

    Data.gov (United States)

    Department of Transportation — The Aerospace Accident Injury Autopsy Database System will provide the Civil Aerospace Medical Institute (CAMI) Aerospace Medical Research Team (AMRT) the ability to...

  3. Noble gas control room accident filtration system for severe accident conditions (N-CRAFT)

    International Nuclear Information System (INIS)

    Hill, Axel; Stiepani, Cristoph; Drechsler, Michael

    2015-01-01

    Severe accidents might cause the release of airborne radioactive substances to the environment of the NPP either due to containment leakages or due to intentional filtered containment venting. In the latter case aerosols and iodine are retained, however noble gases are not retainable by the FCVS or by conventional air filtration systems like HEPA filters and iodine absorbers. Radioactive noble gases nevertheless dominate the activity release depending on the venting procedure and the weather conditions. To prevent unacceptable contamination of the control room atmosphere by noble gases, AREVA GmbH has developed a noble gas control room accident filtration system (CRAFT) which can supply purified fresh air to the control room without time limitation. The retention process is based on dynamic adsorption of noble gases on activated carbon. The system consists of delay lines (carbon columns) which are operated by a continuous and simultaneous adsorption and desorption process. CRAFT allows minimization of the dose rate inside the control room and ensures low radiation exposure to the staff by maintaining the control room environment suitable for prolonged occupancy throughout the duration of the accident. CRAFT consists of a proven modular design either transportable or permanently installed. (author)

  4. SETI VIA LEAKAGE FROM LIGHT SAILS IN EXOPLANETARY SYSTEMS

    International Nuclear Information System (INIS)

    Guillochon, James; Loeb, Abraham

    2015-01-01

    The primary challenge of rocket propulsion is the burden of needing to accelerate the spacecraft’s own fuel, resulting in only a logarithmic gain in maximum speed as propellant is added to the spacecraft. Light sails offer an attractive alternative in which fuel is not carried by the spacecraft, with acceleration being provided by an external source of light. By artificially illuminating the spacecraft with beamed radiation, speeds are only limited by the area of the sail, heat resistance of its material, and power use of the accelerating apparatus. In this paper, we show that leakage from a light sail propulsion apparatus in operation around a solar system analogue would be detectable. To demonstrate this, we model the launch and arrival of a microwave beam-driven light sail constructed for transit between planets in orbit around a single star, and find an optimal beam frequency on the order of tens of GHz. Leakage from these beams yields transients with flux densities of Jy and durations of tens of seconds at 100 pc. Because most travel within a planetary system would be conducted between the habitable worlds within that system, multiply transiting exoplanetary systems offer the greatest chance of detection, especially when the planets are in projected conjunction as viewed from Earth. If interplanetary travel via beam-driven light sails is commonly employed in our galaxy, this activity could be revealed by radio follow-up of nearby transiting exoplanetary systems. The expected signal properties define a new strategy in the search for extraterrestrial intelligence (SETI)

  5. SETI via Leakage from Light Sails in Exoplanetary Systems

    Science.gov (United States)

    Guillochon, James; Loeb, Abraham

    2015-10-01

    The primary challenge of rocket propulsion is the burden of needing to accelerate the spacecraft’s own fuel, resulting in only a logarithmic gain in maximum speed as propellant is added to the spacecraft. Light sails offer an attractive alternative in which fuel is not carried by the spacecraft, with acceleration being provided by an external source of light. By artificially illuminating the spacecraft with beamed radiation, speeds are only limited by the area of the sail, heat resistance of its material, and power use of the accelerating apparatus. In this paper, we show that leakage from a light sail propulsion apparatus in operation around a solar system analogue would be detectable. To demonstrate this, we model the launch and arrival of a microwave beam-driven light sail constructed for transit between planets in orbit around a single star, and find an optimal beam frequency on the order of tens of GHz. Leakage from these beams yields transients with flux densities of Jy and durations of tens of seconds at 100 pc. Because most travel within a planetary system would be conducted between the habitable worlds within that system, multiply transiting exoplanetary systems offer the greatest chance of detection, especially when the planets are in projected conjunction as viewed from Earth. If interplanetary travel via beam-driven light sails is commonly employed in our galaxy, this activity could be revealed by radio follow-up of nearby transiting exoplanetary systems. The expected signal properties define a new strategy in the search for extraterrestrial intelligence (SETI).

  6. SETI VIA LEAKAGE FROM LIGHT SAILS IN EXOPLANETARY SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    Guillochon, James; Loeb, Abraham, E-mail: jguillochon@cfa.harvard.edu, E-mail: aloeb@cfa.harvard.edu [Harvard-Smithsonian Center for Astrophysics, The Institute for Theory and Computation, 60 Garden Street, Cambridge, MA 02138 (United States)

    2015-10-01

    The primary challenge of rocket propulsion is the burden of needing to accelerate the spacecraft’s own fuel, resulting in only a logarithmic gain in maximum speed as propellant is added to the spacecraft. Light sails offer an attractive alternative in which fuel is not carried by the spacecraft, with acceleration being provided by an external source of light. By artificially illuminating the spacecraft with beamed radiation, speeds are only limited by the area of the sail, heat resistance of its material, and power use of the accelerating apparatus. In this paper, we show that leakage from a light sail propulsion apparatus in operation around a solar system analogue would be detectable. To demonstrate this, we model the launch and arrival of a microwave beam-driven light sail constructed for transit between planets in orbit around a single star, and find an optimal beam frequency on the order of tens of GHz. Leakage from these beams yields transients with flux densities of Jy and durations of tens of seconds at 100 pc. Because most travel within a planetary system would be conducted between the habitable worlds within that system, multiply transiting exoplanetary systems offer the greatest chance of detection, especially when the planets are in projected conjunction as viewed from Earth. If interplanetary travel via beam-driven light sails is commonly employed in our galaxy, this activity could be revealed by radio follow-up of nearby transiting exoplanetary systems. The expected signal properties define a new strategy in the search for extraterrestrial intelligence (SETI)

  7. Computerized accident management support system: development for severe accident management

    International Nuclear Information System (INIS)

    Garcia, V.; Saiz, J.; Gomez, C.

    1998-01-01

    The activities involved in the international Halden Reactor Project (HRP), sponsored by the OECD, include the development of a Computerized Accident Management Support System (CAMS). The system was initially designed for its operation under normal conditions, operational transients and non severe accidents. Its purpose is to detect the plant status, analyzing the future evolution of the sequence (initially using the APROS simulation code) and the possible recovery and mitigation actions in case of an accident occurs. In order to widen the scope of CAMS to severe accident management issues, the integration of the MAAP code in the system has been proposed, as the contribution of the Spanish Electrical Sector to the project (with the coordination of DTN). To include this new capacity in CAMS is necessary to modify the system structure, including two new modules (Diagnosis and Adjustment). These modules are being developed currently for Pressurized Water Reactors and Boiling Water REactors, by the engineering of UNION FENOSA and IBERDROLA companies (respectively). This motion presents the characteristics of the new structure of the CAMS, as well as the general characteristics of the modules, developed by these companies in the framework of the Halden Reactor Project. (Author)

  8. Safety analysis of accident localization system

    International Nuclear Information System (INIS)

    1999-01-01

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

  9. Single phase cascaded H5 inverter with leakage current elimination for transformerless photovoltaic system

    DEFF Research Database (Denmark)

    Guo, Xiaoqiang; Jia, X.; Lu, Z.

    2016-01-01

    Leakage current reduction is one of the important issues for the transformelress PV systems. In this paper, the transformerless single-phase cascaded H-bridge PV inverter is investigated. The common mode model for the cascaded H4 inverter is analyzed. And the reason why the conventional cascade H4...... inverter fails to reduce the leakage current is clarified. In order to solve the problem, a new cascaded H5 inverter is proposed to solve the leakage current issue. Finally, the experimental results are presented to verify the effectiveness of the proposed topology with the leakage current reduction...... for the single-phase transformerless PV systems....

  10. Managing the Cost of Plant Piping System Leakage

    International Nuclear Information System (INIS)

    Jenco, John M.; Keck, Donna R.; Johnson, Gary L.

    2002-01-01

    Recent studies have shown that the average annual cost impact of external piping system leakage on commercial nuclear plant operations and maintenance can easily range into the millions of dollars for each reactor unit. Evidence suggests that this significant O and M cost reduction opportunity has largely been overlooked, due to the number of diverse line items and budget areas affected. Results released last year from an EPRI pilot study of more than a dozen reactor units at seven plant sites operated by multiple utilities found that the average annual cost impact was indeed around $1.6 million per year per unit. Subsequent field experience has also demonstrated that an effective fluid leak management program can substantially reduce these costs within the first three years of implementation. This paper presents the general cost impact research results from various studies, outlines key elements of an effective plant fluid leak management program, discusses important implementation issues, and presents results from case studies covering different utility approaches to developing and implementing an effective fluid leak management program. Actual cost data will be included where appropriate. (authors)

  11. Leakage Current Elimination of Four-Leg Inverter for Transformerless Three-Phase PV Systems

    DEFF Research Database (Denmark)

    Guo, Xiaoqiang; He, Ran; Jian, Jiamin

    2016-01-01

    Eliminating the leakage current is one of the most important issues for transformerless three phase photovoltaic (PV) systems. In this paper, the leakage current elimination of a three-phase four-leg PV inverter is investigated. With the common mode loop model established, the generation mechanism...... of the leakage current is clearly identified. Different typical carrier-based modulation methods and their corresponding common mode voltages are discussed. A new modulation strategy with Boolean logic function is proposed to achieve the constant common mode voltage for the leakage current reduction. Finally...

  12. Accident on the gas transfer system

    International Nuclear Information System (INIS)

    Heugel, J.

    1991-10-01

    An accident has happened on the Vivitron gas transfer system on the 7 th August 1991. This report presents the context, facts and inquiries, analyses the reasons and explains also how the repairing has been effected

  13. Leakage radiation spectroscopy of organic/dielectric/metal systems

    DEFF Research Database (Denmark)

    Fiutowski, Jacek; Kawalec, Tomasz; Kostiučenko, Oksana

    2014-01-01

    side of a hemisphere fused silica prism with an index matching liquid was illuminated under normal incidence by a He-Cd 325 nm laser. Two orthogonal linear polarizations were used both parallel and perpendicular to the detection plane. Spectrally resolved leakage radiation was observed on the opposite......Leakage radiation spectroscopy of organic para-Hexaphenylene (p-6P) molecules has been performed in the spectral range 420-675 nm which overlaps with the p-6P photoluminescence band. The p-6P was deposited on 40 nm silver (Ag) films on BK7 glass, covered with SiO2 layers. The SiO2 layer thickness...

  14. Review of the Common European R + D programme on the development and propagation of leakage accidents in sodium-heated steam generators

    International Nuclear Information System (INIS)

    Foerster, K.; Currie, R.; Maupre, J.P.

    1990-01-01

    Sodium-water reactions caused by water-into-sodium leakages are an essential topic in safety considerations for LMFBR steam generators. Research work in this field has been performed worldwide for some decades. Nevertheless, not all questions have been answered yet. So investigations are still going on, nowadays preferably directed to special problem areas and/or to certain design requirements of actual reactor projects. For the European Fast Reactor (EFR) a comprehensive R + D programme is being performed covering all important aspects of leakage accidents (development and consequences of leakages, detection methods, counter measures, codes development). The work is shared between France (CEA), Great Britain (AEA) and Germany (Interatom), cooperating as partners in the development of the EFR. The individual tasks are performed in a great variety of test installations in a well coordinated manner. Based on the results the design basis accident (DBA) for the EFR steam generators will be established and computer codes will be provided for design and licensing purposes. (author). 3 refs, 4 figs, 1 tab

  15. Estimation of Leakage Ratio Using Principal Component Analysis and Artificial Neural Network in Water Distribution Systems

    Directory of Open Access Journals (Sweden)

    Dongwoo Jang

    2018-03-01

    Full Text Available Leaks in a water distribution network (WDS constitute losses of water supply caused by pipeline failure, operational loss, and physical factors. This has raised the need for studies on the factors affecting the leakage ratio and estimation of leakage volume in a water supply system. In this study, principal component analysis (PCA and artificial neural network (ANN were used to estimate the volume of water leakage in a WDS. For the study, six main effective parameters were selected and standardized data obtained through the Z-score method. The PCA-ANN model was devised and the leakage ratio was estimated. An accuracy assessment was performed to compare the measured leakage ratio to that of the simulated model. The results showed that the PCA-ANN method was more accurate for estimating the leakage ratio than a single ANN simulation. In addition, the estimation results differed according to the number of neurons in the ANN model’s hidden layers. In this study, an ANN with multiple hidden layers was found to be the best method for estimating the leakage ratio with 12–12 neurons. This suggested approaches to improve the accuracy of leakage ratio estimation, as well as a scientific approach toward the sustainable management of water distribution systems.

  16. Value-impact analysis of regulatory options for resolution of Generic Issue C-8: MSIV [Main Steam Isolation Valve] leakage and LCS [Leakage Control System] failure

    International Nuclear Information System (INIS)

    Jamison, J.D.; Vo, T.V.; Tabatabai, A.S.

    1990-05-01

    This report describes the analysis conducted to establish the basis for answering two remaining regulatory questions facing the NRC staff regarding the resolution of Generic Issue C-8, specifically:(1) What action should the NRC take concerning plants that currently have a leakage control system (LCS)? and, (2) What action should the NRC take concerning plants that do not have an LCS? Using individual MSIV leak test data, the performance of a system of eight such valves in a standard BWR con-figuration was modeled. The performance model was used along with estimates of core damage accident frequency and calculated dose consequences to determine the public risk associated with each of the alternatives. The occupational exposure implications of each alternative were calculated using estimates of labor hours in radiation zones that would be incurred or avoided. The costs to industry of implementing each alternative were estimated using standard cost formulae and NRC staff estimates. The cost to the NRC were estimated based on the effort incurred or avoided for reviews or other staff actions engendered by the selection of or avoided for reviews or other staff actions engendered by the selection of a particular alternative. The cost and risks thus calculated suggest that no regulatory action can be justified on the basis of risk reduction or cost savings. 12 refs., 1 tab

  17. Development of an Accident Reproduction Simulator System Using a Hemodialysis Extracorporeal Circulation System

    Science.gov (United States)

    Nishite, Yoshiaki; Takesawa, Shingo

    2016-01-01

    Background: Accidents that occur during dialysis treatment are notified to the medical staff via alarms raised by the dialysis apparatus. Similar to such real accidents, apparatus activation or accidents can be reproduced by simulating a treatment situation. An alarm that corresponds to such accidents can be utilized in the simulation model. Objectives: The aim of this study was to create an extracorporeal circulation system (hereinafter, the circulation system) for dialysis machines so that it sets off five types of alarms for: 1) decreased arterial pressure, 2) increased arterial pressure, 3) decreased venous pressure, 4) increased venous pressure, and 5) blood leakage, according to the five types of accidents chosen based on their frequency of occurrence and the degree of severity. Materials and Methods: In order to verify the alarm from the dialysis apparatus connected to the circulation system and the accident corresponding to it, an evaluation of the alarm for its reproducibility of an accident was performed under normal treatment circumstances. The method involved testing whether the dialysis apparatus raised the desired alarm from the moment of control of the circulation system, and measuring the time it took until the desired alarm was activated. This was tested on five main models from four dialyzer manufacturers that are currently used in Japan. Results: The results of the tests demonstrated successful activation of the alarms by the dialysis apparatus, which were appropriate for each of the five types of accidents. The time between the control of the circulatory system to the alarm signal was as follows, 1) venous pressure lower limit alarm: 7 seconds; 2) venous pressure lower limit: 8 seconds; 3) venous pressure upper limit: 7 seconds; 4) venous pressure lower limit alarm: 2 seconds; and 5) blood leakage alarm: 19 seconds. All alarms were set off in under 20 seconds. Conclusions: Thus, we can conclude that a simulator system using an extracorporeal

  18. A new on-line leakage current monitoring system of ZnO surge arresters

    International Nuclear Information System (INIS)

    Lee, Bok-Hee; Kang, Sung-Man

    2005-01-01

    This paper presents a new on-line leakage current monitoring system of zinc oxide (ZnO) surge arresters. To effectively diagnose the deterioration of ZnO surge arresters, a new algorithm and on-line leakage current detection device, which uses the time-delay addition method, for discriminating the resistive and capacitive currents was developed to use in the aging test and durability evaluation for ZnO arrester blocks. A computer-based measurement system of the resistive leakage current, the on-line monitoring device can detect accurately the leakage currents flowing through ZnO surge arresters for power frequency ac applied voltages. The proposed on-line leakage current monitoring device of ZnO surge arresters is more highly sensitive and gives more linear response than the existing devices using the detection method of the third harmonic leakage currents. Therefore, the proposed leakage current monitoring device can be useful for predicting the defects and performance deterioration of ZnO surge arresters in power system applications

  19. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    Kawai, Hiroshi

    2002-01-01

    In the Chemical Volume Control System (CVCS) reactor primary coolant leakage incident, which occurred in Tsuruga-2 (4-loop PWR, 3,423 MWt, 1,160 MWe) on July 12, 1999, it took about 14 hours before the leakage isolation. The delayed leakage isolation and a large amount of leakage have become a social concern. Effective procedure modification was studied. Three betterments were proposed based on a qualitative analysis to reduce the pressure and temperature of the primary loop as fast as possible by the current plant facilities while maintaining enough subcooling of the primary loop. I analyzed the incident with RETRAN code in order to quantitatively evaluate the leakage reduction when these betterments are adopted. This paper is very new because it created a typical analysis method for PWR plant behavior during plant shutdown procedure which conventional RETRAN transient analyses rarely dealt with. Also the event time is very long. To carry out this analysis successfully, I devised new models such as an Residual Heat Removal System (RHR) model etc. and simplified parts of the conventional model. Based on the analysis results, I confirmed that leakage can be reduced by about 30% by adopting these betterments. Then the Japan Atomic Power Company (JAPC) modified the operational procedure for reactor primary coolant leakage events adopting these betterments. (author)

  20. Pressure control for minimizing leakage in water distribution systems

    OpenAIRE

    Nourhan Samir; Rawya Kansoh; Walid Elbarki; Amr Fleifle

    2017-01-01

    In the last decades water resources availability has been a major issue on the international agenda. In a situation of worsening scarcity of water resources and the rapidly increasing of water demands, the state of water losses management is part of manâs survival on earth. Leakage in water supply networks makes up a significant amount, sometimes more than 70% of the total water losses. The best practices suggest that pressure management is one of the most effective way to reduce the amount o...

  1. SU-F-J-144: Scatter and Leakage Survey of An Integrated MR-Linac System

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J; Bosco, G; Darenbourg, B; Ibbott, G [UT MD Anderson Cancer Center, Houston, TX (United States)

    2016-06-15

    Purpose: To assess the scatter and leakage radiation of an integrated 1.5T MRI-Linac system. Methods: A 150cc chamber (model 96020C, Inovision) was used in all the scatter and leakage measurements, after being recalibrated for MV energy by the Accredited Dosimetry Calibration Laboratory at MD Anderson. The scatter radiation was measured by placing a 25 cm stack of solid-water materials at iso-center on the patient couch to simulate patient scatter. Gantry angles were positioned at 0 degree (beam pointing downward) and 270 (beam pointing laterally). Scatter radiation was measured at selective locations inside the RF room. Beam stopper leakage was measured at the exterior panel of the gantry. The head leakage was measured at 1 meter away from the Linac head in the direction which was determined to be the area of maximum leakage by wrapped films test. All measurements were repeated with the 1.5T magnetic field turned off to study the effect of magnetic field. Results: When the magnet was on (B=1.5T), the maximum head leakage at 1 meter was 191.6mR/1000MU. The scatter radiation at 1 meter from the iso-center was 1.091R/1000MU when the radiation beam was pointing downward, 1.296R/1000MU when the beam pointed laterally. The beam stopper leakage was measured as 299.4 mR/1000MU at the exterior panel of the gantry. When magnet was off (B=0), the head leakage was measured as 198.6mR/1000MU. The scatter radiation at 1 meter was 1.153R/1000MU when beam pointed downward, 1.287R/1000MU when beam pointed laterally. The beam stopper leakage was measured as 309.4 mR/1000MU at the exterior panel of the gantry. Conclusion: The measurements indicate that the scatter and leakage radiation from the integrated MR-Linac system are in-line with the expected values. The beam stopper leakage is approximately 300 mR/1000MU. The leakage and scatter difference with the magnetic field ON and OFF was within 5%. The authors received a corporate sponsored grant from Elekta which is the vendor of

  2. Chernobyl - system accident or human error?

    International Nuclear Information System (INIS)

    Stang, E.

    1996-01-01

    Did human error cause the Chernobyl disaster? The standard point of view is that operator error was the root cause of the disaster. This was also the view of the Soviet Accident Commission. The paper analyses the operator errors at Chernobyl in a system context. The reactor operators committed errors that depended upon a lot of other failures that made up a complex accident scenario. The analysis is based on Charles Perrow's analysis of technological disasters. Failure possibility is an inherent property of high-risk industrial installations. The Chernobyl accident consisted of a chain of events that were both extremely improbable and difficult to predict. It is not reasonable to put the blame for the disaster on the operators. (author)

  3. Leakage localisation method in a water distribution system based on sensitivity matrix: methodology and real test

    OpenAIRE

    Pascual Pañach, Josep

    2010-01-01

    Leaks are present in all water distribution systems. In this paper a method for leakage detection and localisation is presented. It uses pressure measurements and simulation models. Leakage localisation methodology is based on pressure sensitivity matrix. Sensitivity is normalised and binarised using a common threshold for all nodes, so a signatures matrix is obtained. A pressure sensor optimal distribution methodology is developed too, but it is not used in the real test. To validate this...

  4. Regulatory analysis for the resolution of generic issue C---8, main steam isolation valve leakage and LCS [leakage control system] failure

    International Nuclear Information System (INIS)

    Graves, C.C.

    1990-06-01

    Generic Issue C-8 deals with staff concerns about public risk because of the incidence of leak test failures reported for main steam isolation valves (MSIVs) at boiling water reactors and the limitations of the leakage control systems (LCSs) for mitigating the consequences of leakage from these valves. If the MSIV leakage is greatly in excess of the allowable value in the technical specifications, the LCS would be unavailable because of design limitations. The issue was initiated in 1983 to assess (1) the causes of MSIV leakage failures, (2) the effectiveness of the LCS and alternative mitigation paths, and (3) the need for additional regulatory action to reduce public risk. This report presents the regulatory analysis for Generic Issue C-8 and concludes that no new regulatory requirements are warranted

  5. Ship cabin leakage alarm based on ARM SCM

    Science.gov (United States)

    Qu, Liyan

    2018-03-01

    If there is a leakage in the cabin of a sailing ship, it is a major accident that threatens the personnel and property of the ship. If we can’t take timely measures, there will be a devastating disaster. In order to judge the leakage of the cabin, it is necessary to set up a leakage alarm system, so as to achieve the purpose of detecting and alarming the leakage of the cabin, and avoid the occurrence of accidents. This paper discusses the design of ship cabin leakage alarm system based on ARM SCM. In order to ensure the stability and precision of the product, the hardware design of the alarm system is carried out, such as circuit design, software design, the programming of SCM, the software programming of upper computer, etc. It is hoped that it can be of reference value to interested readers.

  6. Development of Traffic Accidents Control System

    Directory of Open Access Journals (Sweden)

    Andrey Borisovich Nikolaev

    2015-05-01

    Full Text Available Proposed a structure of traffic accidents control system included three main parts: pre-processing, decision support and monitoring. For decision support systems we propose a method that allows to make decisions on the basis of fuzzy situational management. The advantage of the method: it allows to formalize a set of typical traffic situations, using the theory of fuzzy sets and to carry out selection of the desired management action.

  7. Description of leakage monitoring system at Angra 2 nuclear power plant primary circuit

    International Nuclear Information System (INIS)

    Costa, Lilian Rose Sobral da; Mendes, Jorge Eduardo de Souza

    1999-01-01

    This paper describes the Leakage Monitoring System installed in Angra 2 NPP. This system has the task of detecting, localizing and quantifying leaks in systems for which rupture preclusion is cited. These systems include the reactor coolant pressure boundary, the main steam and feedwater lines within the containment, and the main steam safety and relief valve station in the valve annex. (author)

  8. Limitations of systemic accident analysis methods

    Directory of Open Access Journals (Sweden)

    Casandra Venera BALAN

    2016-12-01

    Full Text Available In terms of system theory, the description of complex accidents is not limited to the analysis of the sequence of events / individual conditions, but highlights nonlinear functional characteristics and frames human or technical performance in relation to normal functioning of the system, in safety conditions. Thus, the research of the system entities as a whole is no longer an abstraction of a concrete situation, but an exceeding of the theoretical limits set by analysis based on linear methods. Despite the issues outlined above, the hypothesis that there isn’t a complete method for accident analysis is supported by the nonlinearity of the considered function or restrictions, imposing a broad vision of the elements introduced in the analysis, so it can identify elements corresponding to nominal parameters or trigger factors.

  9. Development of a totally integrated severe accident training system

    International Nuclear Information System (INIS)

    Kim, Ko Ryu; Park, Sun Hee; Choi, Young; Kim, Dong Ha

    2006-01-01

    Recently KAERI has developed the severe accident management guidance to establish the Korea standard severe accident management system. On the other hand the PC-based severe accident training simulator SATS has been developed, which uses the MELCOR code as the simulation engine. The simulator SATS graphically displays and simulates the severe accidents with interactive user commands. Especially the control capability of SATS could make a severe accident training course more interesting and effective. In this paper we will describe the development and functions of the electrical guidance module, HyperKAMG, and the SATS-HyperKAMG linkage system designed for a totally integrated and automated severe accident training. (author)

  10. Aspects of severe accidents in transmutation systems

    International Nuclear Information System (INIS)

    Wider, H.U.; Karlson, J.; Jones, A.V.

    2001-01-01

    The different types of transmutation systems under investigation include accelerator driven (ADS) and critical systems. To switch off an accelerator in case of an accident initiation is quite important for all accidents. For a fast ADS the grace times available for doing so depend strongly on the total heat capacity and the natural circulation capability of the primary coolant. Cooling with heavy metal Pb-Bi has considerable advantages in this regard compared to gas cooling. Moreover it allows passive ex-vessel cooling with natural air or water circulation. In the remote likelihood of fuel melting, oxide fuel appears to mix with the Pb-Bi coolant. Fast critical systems that are cooled by Pb-Bi will automatically shut off if the flow or heat sink is lost. Reactivity accidents can be limited by a low total control rod worth. High temperature reactors can achieve only incomplete burning of actinides. If an accelerator is added to increase burn-up, a fast spectrum region is needed, which has a low heat capacity. (author)

  11. Leakage Performance of the GM + CCL Liner System for the MSW Landfill

    Directory of Open Access Journals (Sweden)

    Fan Jingjing

    2014-01-01

    Full Text Available The contaminants in the landfill leachate press pose a grave threat to environment of the soil and the groundwater beneath the landfill. Despite there being strict requirements in relevant provisions of both domestic and foreign countries for the design of the bottom liner system. Pollution of the soil and the groundwater still took place in a number of landfills because of the leakage. To investigate the leakage rate of the liner systems, the minimum design requirements of the liner systems are summarized according to the provisions of four countries, including China, USA, Germany, and Japan. Comparative analyses using one-dimensional transport model are conducted to study the leakage performance of these liner systems composed of geomembrance (GM and compacted clay layer (CCL meeting the relevant minimum design requirements. Then parametric analyses are conducted to study the effects of the hydraulic head, the thickness of GM, the hydraulic conductivity of CCL, and so forth on the leakage performance of the liner system. It is concluded that the liner system designed according to the minimum design requirements of Germany provide the best antileakage performance, while that of Japan performs the lowest. The key parameters affecting the failure time of the liner system are summarized. Finally, some suggestions for the design of the liner systems are made according to the analyses.

  12. Monitoring and operation system for severe accidents

    International Nuclear Information System (INIS)

    Fukui, Toshiki; Niida, Shinji; Kato, Yumeto

    2017-01-01

    Monitoring and operation system for Severe Accidents (SA-MOS) is a compact Instrumentation and Control (I and C) system developed by Mitsubishi Heavy Industries (MHI) and certificated by the Japanese Nuclear Regulatory Agency (NRA) as a design application for Japanese existing PWR nuclear power plants. The system is tailored to provide monitoring and operation for Severe Accident (SA) conditions, and consists of digitalized I and C System, Human Systems Interface (HSI) system and Power Supply (PS) system as further improvement of reliability and safety. This design plans to be applied to the next Japanese PWR plants. In accordance with the new regulatory standards that NRA has established corresponding to the Fukushima accident, a long-term Station Black Out (SBO) scenario and 24-hours power supply by the storage battery in case of SA has been required. In order to address 24-hours power supply requirement in SA condition, the storage battery volume shall be increased. However, it may be difficult to introduce additional batteries to the existing plant site because of room space constraints, etc. Therefore, power distributions for the facilities which are only used for Design Basis Accident (DBA), are shut down in order to secure 24-hours operations of facilities for SA conditions including SA-MOS. That enables efficient battery resource operations as well as optimizes room space factors shared by battery cabinets. Another benefit is to introduce dedicate HSI system for SA condition and operators shift their operations using that dedicated HSI system to cope with SA events. That can reduce operator workload which forces operators to verify or choose which controllers and indicators are available in SA conditions. Furthermore, application of SA-MOS, secures the independence of the layers (DBA⇔SA) as well as secures the plant data transfer for SA conditions outside of plant. Those plant data assets can be shared by plant operation supporting personnel and

  13. Monitoring device for radioactive leakage from steam system in nuclear power plant

    International Nuclear Information System (INIS)

    Ogawa, Tateo; Sato, Kohei

    1988-01-01

    Purpose: To improve the reliability for the monitor of radio-active leakage by accumulating small quantity of radioactivities each lower than the detectable level and increasing their dose rate. Constitution: Even if the steam system radiation monitor in the nuclear power plant is disposed for the detection of in-leak radioactivity, radioactive leakage can be monitored at high reliability by increasing the small quantity of radioactivities in the drains to a detectable sensitivity range of the monitor upon detection. In view of the above, in the present invention, radioactive material catching medium is incorporated to a radio-activity monitor spool piece for accumulating small quantity of radioactivities. Specifically as the catching medium, an ion exchange resin is used for the leakage of ionic radioactive material, while an ion exchange resin increased with the mixing ratio of a cationic resin or hollow thread membrane filter is used for crud-like radioactive material leakage. These catching media are incorporated into the spool piece, thereby enabling to catch even small quantity of radioactive leakage lower than the detectable sensitivity of the radiation monitor, if should occur, in the spool piece and enabling radioactive detection for the accumulated dose rate. (Horiuchi, T.)

  14. Passive Decay Heat Removal Strategy of Integrated Passive Safety System (IPSS) for SBO-combined Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ho; Chang, Soon Heung; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    The weak points of nuclear safety would be in outmoded nuclear power plants like the Fukushima reactors. One of the systems for the safety enhancement is integrated passive safety system (IPSS) proposed after the Fukushima accidents. It has the five functions for the prevention and mitigation of a severe accident. Passive decay heat removal (PDHR) strategy using IPSS is proposed for coping with SBO-combined accidents in this paper. The two systems for removing decay heat before core-melt were applied in the strategy. The accidents were simulated by MARS code. The reference reactor was OPR1000, specifically Ulchin-3 and 4. The accidents included loss-of-coolant accidents (LOCA) because the coolant losses could be occurred in the SBO condition. The examples were the stuck open of PSV, the abnormal open of SDV and the leakage of RCP seal water. Also, as LOCAs with the failure of active safety injection systems were considered, various LOCAs were simulated in SBO. Based on the thermal hydraulic analysis, the probabilistic safety analysis was carried out for the PDHR strategy to estimate the safety enhancement in terms of the variation of core damage frequency. AIMS-PSA developed by KAERI was used for calculating CDF of the plant. The IPSS was applied in the PDHR strategy which was developed in order to cope with the SBO-combined accidents. The estimation for initiating SGGI or PSIS was based on the pressure in RCS. The simulations for accidents showed that the decay heat could be removed for the safety duration time in SBO. The increase of safety duration time from the strategy provides the increase of time for the restoration of AC power.

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

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1976-01-01

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

  16. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    Kawai, H.

    2001-01-01

    JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

  17. Effect of vadose zone on the steady-state leakage rates from landfill barrier systems

    International Nuclear Information System (INIS)

    Celik, B.; Rowe, R.K.; Unlue, K.

    2009-01-01

    Leakage rates are evaluated for a landfill barrier system having a compacted clay liner (CCL) underlain by a vadose zone of variable thickness. A numerical unsaturated flow model SEEP/W is used to simulate the moisture flow regime and steady-state leakage rates for the cases of unsaturated zones with different soil types and thicknesses. The results of the simulations demonstrate that harmonic mean hydraulic conductivity of coarse textured vadose zones is 3-4 orders of magnitude less than saturated hydraulic conductivity; whereas, the difference is only one order of magnitude for fine textured vadose zones. For both coarse and fine textured vadose zones, the effective hydraulic conductivity of the barrier system and the leakage rate to an underlying aquifer increases with increasing thickness of the vadose zone and ultimately reaches an asymptotic value for a coarse textured vadose zone thickness of about 10 m and a fine textured vadose zone thickness of about 5 m. Therefore, the fine and coarse textured vadose zones thicker than about 5 m and 10 m, respectively, act as an effective part of the barrier systems examined. Although the thickness of vadose zone affects the effective hydraulic conductivity of the overall barrier system, the results demonstrated that the hydraulic conductivity of the CCL is the dominant factor controlling the steady-state leakage rates through barrier systems having single low permeability clay layers

  18. Surveillance systems (PWR) - loose parts monitoring - vibration monitoring - leakage detection

    International Nuclear Information System (INIS)

    Schuette, A.; Blaesig, H.

    1982-01-01

    The contribution is engaged in the task and the results of the loose parts monitoring and the vibration monitoring following from the practice at the PWR of Biblis. First a description of both systems - location and type of the sensors used, the treatment of the measurements and the indications - is given. The results of the analysis of some events picked up by the surveillance systems are presented showing applicabilty and benefit of such systems. (orig.)

  19. Modular telerobot control system for accident response

    Science.gov (United States)

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

    1999-08-01

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

  20. Review of nuclear power reactor coolant system leakage events and leak detection requirements

    International Nuclear Information System (INIS)

    Chokshi, N.C.; Srinivasan, M.; Kupperman, D.S.; Krishnaswamy, P.

    2005-01-01

    In response to the vessel head event at the Davis-Besse reactor, the U.S. Nuclear Regulatory Commission (NRC) formed a Lessons Learned Task Force (LLTF). Four action plans were formulated to respond to the recommendations of the LLTF. The action plans involved efforts on barrier integrity, stress corrosion cracking (SCC), operating experience, and inspection and program management. One part of the action plan on barrier integrity was an assessment to identify potential safety benefits from changes in requirements pertaining to leakage in the reactor coolant system (RCS). In this effort, experiments and models were reviewed to identify correlations between crack size, crack-tip-opening displacement (CTOD), and leak rate in the RCS. Sensitivity studies using the Seepage Quantification of Upsets In Reactor Tubes (SQUIRT) code were carried out to correlate crack parameters, such as crack size, with leak rate for various types of crack configurations in RCS components. A database that identifies the leakage source, leakage rate, and resulting actions from RCS leaks discovered in U.S. light water reactors was developed. Humidity monitoring systems for detecting leakage and acoustic emission crack monitoring systems for the detection of crack initiation and growth before a leak occurs were also considered. New approaches to the detection of a leak in the reactor head region by monitoring boric-acid aerosols were also considered. (authors)

  1. Reactor accident diagnostic expert system: DISKET

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Yokobayashi, Masao

    1989-11-01

    A reactor accident diagnostic system DISKET has been developed to identify the cause and the type of an abnormal transient of a nuclear power plant. The system is based on the knowledge engineering and consists of an inference engine IERIAS and a knowledge base. The main features of DISKET are the following: Time-varying characteristics of transient can be treated and knowledge base can be divided into several knowledge units to handle a lot of rules effectively. This report has been provided for the convenience of DISKET's users and consists of three parts. The first part is the description of the whole system, the details of the knowledge base of DISKET are described in the second part, and how to use the DISKET system is explained in the third part. (author)

  2. ATR confinement leakage determination

    International Nuclear Information System (INIS)

    Kuan, P.; Buescher, B.J.

    1998-01-01

    The air leakage rate from the Advanced Test Reactor (ATR) confinement is an important parameter in estimating hypothesized accidental releases of radiation to the environment. The leakage rate must be determined periodically to assure that the confinement has not degraded with time and such determination is one of the technical safety requirements of ATR operation. This paper reviews the methods of confinement leakage determination and presents an analysis of leakage determination under windy conditions, which can complicate the interpretation of the determined leakage rates. The paper also presents results of analyses of building air exchange under windy conditions. High wind can enhance air exchange and this could increase the release rates of radioisotopes following an accident

  3. Improving Reliability of Information Leakage Detection and Prevention Systems

    Directory of Open Access Journals (Sweden)

    A. V. Mamaev

    2011-03-01

    Full Text Available The problem of protection from deliberate leaks of information is one of the most difficult. Integrated systems of information protection against insider have a serious drawback. Using this disadvantage the offender receives the possibility of unauthorized theft of information from working machine.

  4. Secret-key rates and privacy leakage in biometric systems

    NARCIS (Netherlands)

    Ignatenko, T.

    2009-01-01

    In this thesis both the generation of secret keys from biometric data and the binding of secret keys to biometric data are investigated. These secret keys can be used to regulate access to sensitive data, services, and environments. In a biometric secrecy system a secret key is generated or chosen

  5. Privacy-leakage codes for biometric authentication systems

    NARCIS (Netherlands)

    Ignatenko, T.; Willems, F.M.J.

    2014-01-01

    In biometric privacy-preserving authentication systems that are based on key-binding, two terminals observe two correlated biometric sequences. The first terminal selects a secret key, which is independent of the biometric data, binds this secret key to the observed biometric sequence and

  6. Specific interaction of central nervous system myelin basic protein with lipids effects of basic protein on glucose leakage from liposomes

    NARCIS (Netherlands)

    Gould, R.M.; London, Y.

    1972-01-01

    The leakage from liposomes preloaded with glucose was continuously monitored in a Perkin-Elmer Model 356 dual beam spectrophotometer using an enzyme-linked assay system. The central nervous system myelin basic protein (A1 protein) caused a 3–4-fold increase in the rate of leakage from liposomes

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  8. Low Leakage Turbine Shaft Seals for Advanced Combined Cycle Systems.

    Science.gov (United States)

    1984-11-01

    Study of Shaft Face Seal With Self-Acting Lift Augmentation", N71- 11579, Nov. 1970 29p. Povinelli , V.P. and McKibbin, A. H., "Development of...34, N73-24086, May 1973, 28p. Povinelli , V. P. and McKibbin, A. H., "Development of Mainshaft Seals for Advanced Air Breathing Propulsion Systems... Povinelli , V. P., "Current Seal Designs and Future Requirements for Turbine Engine Seals and Bearings", Journal of Aircraft, Vol. 12, No. 4, April 1975

  9. The effect of system modeling on the Fukushima accident evolution

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L.E.; Fontanet, J.; López, C.; Fernández, E.

    2015-07-01

    The Fukushima accident is becoming both a unique opportunity and a huge challenge for severe accident analysis. The OECD-BSAF project has articulated a good part of the modeling efforts conducted so far. Inside this project, CIEMAT has conducted forensic analyses of the Fukushima accident in units 1 through 3 with MELCOR 2.1 and it has postulated a set of accident scenarios consistent with data. Beyond specific results, sensitivity analyses on safety systems performance and prevailing boundary conditions have highlighted the need of conducting uncertainty analyses when modeling NPPs severe accident scenarios. (Author)

  10. Power crowbar system coupled by a current transformer with very low leakage inductance

    International Nuclear Information System (INIS)

    Kitagawa, S.; Hirano, K.I.

    1976-01-01

    A reliable, efficient power crowbar system has been developed for fast pinch experiments. In order to reduce the effective impedance of series capacitor system, a current transformer with extremely low leakage inductance has been designed and used. Primary and secondary windings of the transformer are alternately arranged as closely as possible. As a result, the leakage inductance is reduced to 2 nH. It is demonstrated that a current of 390 kA, the rise time of which is 4.5 μsec, is sustained for 100 μsec. Much larger system is being built, which maintains a current of 1 MA over 1 msec. The life of crowbar gap switches is prolonged by the aid of a mechanically-driven metal-to-metal contact switch. Another crowbar switch system with a high coulomb rating is under consideration, in which a gap switch is used together with a saturable reactor and a current transformer

  11. [System design of open-path natural gas leakage detection based on Fresnel lens].

    Science.gov (United States)

    Xia, Hui; Liu, Wen-Qing; Zhang, Yu-Jun; Kan, Rui-Feng; Cui, Yi-Ben; Wang, Min; He, Ying; Cui, Xiao-Juan; Ruan, Jun; Geng, Hui

    2009-03-01

    Based on the technology of tunable diode laser absorption spectroscopy (TDLAS) in conjunction with second harmonic wave detection, a long open-path TDLAS system using a 1.65 microm InGaAsP distributed feedback laser was developed, which is used for detecting pipeline leakage. In this system, a high cost performance Fresnel lens is used as the receiving optical system, which receives the laser-beam reflected by a solid corner cube reflector, and focuses the receiving laser-beam to the InGaAs detector. At the same time, the influences of the concentration to the fluctuation of light intensity were taken into account in the process of measurement, and were eliminated by the method of normalized light intensity. As a result, the measurement error caused by the fluctuation of light intensity was made less than 1%. The experiment of natural gas leakage detection was simulated, and the detection sensitivity is 0.1 x 10(-6) (ratio by volume) with a total path of 320 m. According to the receiving light efficiency of the optical system and the detectable minimum light intensity of the detector, the detectable maximal optical path of the system was counted to be 2 000 m. The results of experiment show that it is a feasible design to use the Fresnel lens as the receiving optical system and can satisfy the demand of the leakage detection of natural gas.

  12. Phosphodiesterase-4 inhibition as a therapeutic approach to treat capillary leakage in systemic inflammation.

    Science.gov (United States)

    Schick, Martin Alexander; Wunder, Christian; Wollborn, Jakob; Roewer, Norbert; Waschke, Jens; Germer, Christoph-Thomas; Schlegel, Nicolas

    2012-06-01

    In sepsis and systemic inflammation, increased microvascular permeability and consecutive breakdown of microcirculatory flow significantly contribute to organ failure and death. Evidence points to a critical role of cAMP levels in endothelial cells to maintain capillary endothelial barrier properties in acute inflammation. However, approaches to verify this observation in systemic models are rare. Therefore we tested here whether systemic application of the phosphodiesterase-4-inhibitors (PD-4-Is) rolipram or roflumilast to increase endothelial cAMP was effective to attenuate capillary leakage and breakdown of microcirculatory flow in severe lipopolysaccharide (LPS)-induced systemic inflammation in rats. Measurements of cAMP in mesenteric microvessels demonstrated significant LPS-induced loss of cAMP levels which was blocked by application of rolipram. Increased endothelial cAMP by application of either PD-4-I rolipram or roflumilast led to stabilization of endothelial barrier properties as revealed by measurements of extravasated FITC-albumin in postcapillary mesenteric venules. Accordingly, microcirculatory flow in mesenteric venules was significantly increased following PD-4-I treatment and blood gas analyses indicated improved metabolism. Furthermore application of PD-4-I after manifestation of LPS-induced systemic inflammation and capillary leakage therapeutically stabilized endothelial barrier properties as revealed by significantly reduced volume resuscitation for haemodynamic stabilization. Accordingly microcirculation was significantly improved following treatment with PD-4-Is. Our results demonstrate that inflammation-derived loss of endothelial cAMP contributes to capillary leakage which was blocked by systemic PD-4-I treatment. Therefore these data suggest a highly clinically relevant and applicable approach to stabilize capillary leakage in sepsis and systemic inflammation.

  13. Design and application of leakage monitor for reactor and control rod driving system

    International Nuclear Information System (INIS)

    Li, Dongyu; Zou, Yimin; Ling, Qiu; Guo, Lanying

    2009-04-01

    By measuring the number of γ photons produced by the annihilation of the β + particles of 13 N's decay product in the sample air, the nuclide density of 13 N can be obtained, comparing with its density in the reactor coolant, we can get the leakage information of the reactor vessel and control rod driving system, the article describes the cause of improvement in monitoring for leakage of reactor vessel and control rod driving system of Qinshan Second Nuclear Power Plant (PWR reactor), also the determination of monitoring method and system configuration, as well as the main technical index and function. Furthermore, the main parts and its function of the monitor are introduced. After operation for more than four years, it is proved that both the stability and MTBF index of the monitor meet the design, even more, thanks to the improvement of the algorithm, the Compton Effect caused by other nuclide became neglectable, the MDA of the monitor was lowered also. (authors)

  14. A transformerless single-phase symmetrical Z-source HERIC inverter with reduced leakage currents for PV systems

    DEFF Research Database (Denmark)

    Li, Kerui; Shen, Yanfeng; Yang, Yongheng

    2018-01-01

    and thus low leakage currents in PV applications. The symmetric Z-source HERIC inverter requires two extra active switches. Nevertheless, the operation frequency of the two switches is the line frequency, leading to negligible losses. More importantly, the performance in terms of low leakage currents...... and harmonics is improved. Experimental tests are performed to validate the analysis and performance of the proposed system....

  15. Strategy generator in computerized accident management support system

    International Nuclear Information System (INIS)

    Sirola, M.

    1994-02-01

    An increased interest for research in the field of accident management of nuclear power plants can be noted. Several international programmes have been started in order to be able to understand the basic physical and chemical phenomena in accident conditions. A feasibility study has shown that it would be possible to design and develop a computerized support system for plant staff in accident situations. To achieve this goal the Halden Project has initiated a research programme on Computerized Accident Management Support (CAMS project). The aim is to utilize the capabilities of computerized tools to support the plant staff during the various accident stages. The system will include identification of the accident state, assessment of the future development of the accident and planning of accident mitigation strategies. A prototype is developed to support operators and the Technical Support Centre in decision making during serious accidents in nuclear power plants. A rule based system has been built to take care of the strategy generation. This system assists plant personnel in planning control proposals and mitigation strategies from normal operation to severe accident conditions. The idea of a safety objective tree and knowledge from the emergency procedures have been used. Future prediction requires good state identification of the plant status and some knowledge about the history of some critical variables. The information needs to be validated as well. Accurate calculations in simulators and a large database including all important information from the plant will help the strategy planning. (orig.). (40 refs., 20 figs.)

  16. Design of a DCS Based Model for Continuous Leakage Monitoring System of Rotary Air Preheater of a Thermal Power Plant

    Directory of Open Access Journals (Sweden)

    Madan BHOWMICK

    2011-01-01

    Full Text Available The leakage in rotary air preheater makes a considerable contribution to the reduced overall efficiency of fossil-fuel-fired thermal power plants and increase the effect on environment. Since it is normal phenomenon, continuous monitoring of leakage is generally omitted in most power plants. But for accurate analysis of the operation of the thermal power plant, this leakage monitoring plays a vital role. In the present paper, design of a DCS based model for continuous leakages monitoring of rotary air preheater has been described. In the proposed model, the existing DCS based instrumentation system has been modified and online leakage monitoring system has been developed. This model has been installed in a captive power plant with high capacity boilers and very much satisfactory operation of this system has been observed. The observed online data along with their analysis results are presented in this paper.

  17. Development of radiotracer technology and nucleonic control systems - Development of leakage monitoring system using radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Choon; Lee, Doon Sung [Seoil College, Seoul (Korea); Cho, Yong Suk [Konyang University, Nonsan (Korea); Shin, Seung Kwon [Sungkyunkwan University, Seoul (Korea)

    1999-04-01

    This is the equipment to find out the leakage of an oil pipeline and a water pipe reclaimed from the underground, using a Radioisotopes. For this purpose, small sized, light weight and water proofed Detection Pig in high voltage is to be designed and produced as to develope a radiation detector, high voltage power supply circuit, measurement circuit, data storage circuit, and a remote on/off function, etc. In this study, high-voltage producing circuit, an amplifier, count rate circuit, processing circuit, and radiosignal treatment part will be used as a basic material to develope a radiotracer, a radiation measuring and using equipment. And it will promote the localization of the radiation applied measuring system. As the results, domestic equipment industry will be improved a great deal. 22 refs., 22 figs., 8 tabs. (Author)

  18. Chemical dosimetry system for criticality accidents.

    Science.gov (United States)

    Miljanić, Saveta; Ilijas, Boris

    2004-01-01

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

  19. Detection of underground water distribution piping system and leakages using ground penetrating radar (GPR)

    Science.gov (United States)

    Amran, Tengku Sarah Tengku; Ismail, Mohamad Pauzi; Ahmad, Mohamad Ridzuan; Amin, Mohamad Syafiq Mohd; Sani, Suhairy; Masenwat, Noor Azreen; Ismail, Mohd Azmi; Hamid, Shu-Hazri Abdul

    2017-01-01

    A water pipe is any pipe or tubes designed to transport and deliver water or treated drinking with appropriate quality, quantity and pressure to consumers. The varieties include large diameter main pipes, which supply entire towns, smaller branch lines that supply a street or group of buildings or small diameter pipes located within individual buildings. This distribution system (underground) is used to describe collectively the facilities used to supply water from its source to the point of usage. Therefore, a leaking in the underground water distribution piping system increases the likelihood of safe water leaving the source or treatment facility becoming contaminated before reaching the consumer. Most importantly, leaking can result in wastage of water which is precious natural resources. Furthermore, they create substantial damage to the transportation system and structure within urban and suburban environments. This paper presents a study on the possibility of using ground penetrating radar (GPR) with frequency of 1GHz to detect pipes and leakages in underground water distribution piping system. Series of laboratory experiment was designed to investigate the capability and efficiency of GPR in detecting underground pipes (metal and PVC) and water leakages. The data was divided into two parts: 1. detecting/locating underground water pipe, 2. detecting leakage of underground water pipe. Despite its simplicity, the attained data is proved to generate a satisfactory result indicating GPR is capable and efficient, in which it is able to detect the underground pipe and presence of leak of the underground pipe.

  20. Design of modified annulus air sampling system for the detection of leakage in waste transfer line

    International Nuclear Information System (INIS)

    Deokar, U.V; Khot, A.R.; Mathew, P.; Ganesh, G.; Tripathi, R.M.; Srivastava, Srishti

    2018-01-01

    Various liquid waste streams are generated during the operation of reprocessing plant. The High Level (HL), Intermediate Level (IL) and Low Level (LL) liquid wastes generated, are transferred from reprocessing plant to Waste Management Facility. These respective waste streams are transferred through pipe-in-pipe lines along the shielded concrete trench. For detection of radioactive leakage from primary waste transfer line into secondary line, sampling of the annulus air between the two pipes is carried out. The currently installed pressurized annulus air sampling system did not have online leakage detection provision. Hence, there are chances of personal exposure and airborne activity in the working area. To overcome these design flaws, free air flow modified online annulus air sampling system with more safety features is designed

  1. Determination of limits for smallest detectable and largest subcritical leakage cracks in piping systems

    International Nuclear Information System (INIS)

    Bieselt, R.; Wolf, M.

    1995-01-01

    Nuclear power plant piping systems - those still in their original as-built condition as well as upgraded designs - are subject to safety analysis. In order to limit the consequences of postulated piping failures, the basic safety concept incorporating rupture preclusion criteria is applied to specific high-energy piping systems. Leak-before-break analyses are also conducted within the framework of this concept. These analyses serve to determine the potential consequences of jet and reaction forces due to maximum subcritical leakage cracks while also establishing the minimum crack sizes that would be reliably detectable by the leakage rates resulting from these cracks. The boundary conditions for these analyses are not clearly defined. Using various examples as a basis, this paper presents and discusses how the leak-before-break concept can be applied. (orig.)

  2. Modeling PWR systems for monitoring primary-to-secondary leakage using tritium tracer

    International Nuclear Information System (INIS)

    Peiffer, D.G.

    1992-01-01

    This paper discusses several techniques available for monitoring primary to secondary leakage, focusing on the advantages and disadvantages of each. A mathematical model of Millstone 2 describes the behavior of tritium activity in the secondary plant water when a leak exists. Real data from Millstone 2 illustrate the accuracy and reliability of the model and use of the model to measure the mass of water in the secondary system

  3. Radiation leakage monitoring method and device from primary to secondary coolant systems in nuclear reactor

    International Nuclear Information System (INIS)

    Tajiri, Yoshiaki; Umehara, Toshihiro; Yamada, Masataka.

    1993-01-01

    The present invention monitors radiation leaked from any one of primary cooling systems to secondary cooling systems in a plurality of steam generators. That is, radiation monitoring means each corresponding to steam each generators are disposed to the upstream of a position where main steam pipes are joined. With such a constitution, since the detection object of each of radiation monitoring means is secondary coolants before mixing with secondary coolants of other secondary loops or dilution, lowering of detection accuracy can be avoided. Except for the abnormal case, that is, a case neither of radiation leakage nor of background change, the device is adapted as a convenient measuring system only with calculation performance. Once abnormality occurs, a loop having a value exceeding a standard value is identified by a single channel analyzer function. The amount of radiation leakage from the steam generator belonging to the specified loop is monitored quantitatively by a multichannel analyzer function. According to the method of the present invention, since specific spectrum analysis is conducted upon occurrence of abnormality, presence of radiation leakage and the scale thereof can be judged rapidly. (I.S.)

  4. Early Leakage Protection System of LPG (Liquefied Petroleum Gas) Based on ATMega 16 Microcontroller

    Science.gov (United States)

    Sriwati; Ikhsan Ilahi, Nur; Musrawati; Baco, Syarifuddin; Suyuti'Andani Achmad, Ansar; Umrianah, Ejah

    2018-04-01

    LPG (Liquefied Petroleum Gas). LPG is a hydrocarbon gas production from refineries and gas refinery with the major components of propane gas (C3H8) and butane (C4H10). Limit flame (Flammable Range) or also called gas with air. Value Lower Explosive Limit (LEL) is the minimum limit of the concentration of fuel vapor in the air which if there is no source of fire, the gas will be burned. While the value of the Upper Explosive Limit (UEL), which limits the maximum concentration of fuel vapor in the air, which if no source of fire, the gas will be burned. Protection system is a defend mechanism of human, equipment, and buildings around the protected area. Goals to be achieved in this research are to design a protection system against the consequences caused by the leakage of LPG gas based on ATmega16 microcontroller. The method used in this research is to reduce the levels of leaked LPG and turned off the power source when the leakage of LPG is on the verge of explosive limit. The design of this protection system works accurately between 200 ppm up to 10000 ppm, which is still below the threshold of explosive. Thus protecting the early result of that will result in the leakage of LPG gas.

  5. Detection device for off-gas system accidents

    International Nuclear Information System (INIS)

    Kubota, Ryuji; Tsuruoka, Ryozo; Yamanari, Shozo.

    1984-01-01

    Purpose: To rapidly isolate the off-gas system by detecting the off-gas system failure accident in a short time. Constitution: Radiation monitors are disposed to ducts connecting an exhaust gas area and an air conditioning system as a portion of a turbine building. The ducts are disposed independently such that they ventilate only the atmosphere in the exhaust gas area and do not mix the atmosphere in the turbine building. Since radioactivity issued upon off-gas accidents to the exhaust gas area is sucked to the duct, it can be detected by radiation detection monitors in a short time after the accident. Further, since the operator judges it as the off-gas system accident, the off-gas system can be isolated in a short time after the accident. (Moriyama, K.)

  6. Design and Development of a Severe Accident Training System

    International Nuclear Information System (INIS)

    Kim, Ko Ryu; Park, Sun Hee; Kim, Dong Ha

    2005-01-01

    The nuclear plants' severe accidents have two big characteristics. One is that they are very rare accidents, and the other is that they bring extreme conditions such as the high pressure and temperature in their process. It is, therefore, very hard to get the severe accident data, without inquiring that the data should be real or experimental. In fact, most of severe accident analyses rely on the simulation codes where almost all severe accident knowledge is contained. These codes are, however, programmed by the Fortran language, so that their output are typical text files which are very complicated. To avoid this kind of difficulty in understanding the code output data, several kinds of graphic user interface (GUI) programs could be developed. In this paper, we will introduce a GUI system for severe accident management and training, partly developed and partly in design stage

  7. Development of an accident management expert system for containment assessment

    International Nuclear Information System (INIS)

    Nelson, W.R.; Sebo, D.E.; Haney, L.N.

    1987-01-01

    The United States Nuclear Regulatory Commission (USNRSC) is sponsoring a program at the Idaho National Engineering Laboratory (INEL) to develop an accident management expert system. The intended users of the system are the personnel of the NRC Operations Center in Washington, D.C. The expert system will be used to help NRC personnel monitor and evaluate the status and management of the containment during a severe reactor accident. The knowledge base will include severe accident knowledge regarding the maintenance of the critical safety functions, especially containment integrity, during an accident. This paper summarizes the concepts that have been developed for the accident management expert system, and the plans that have been developed for its implementation

  8. System 80+ design features for severe accident prevention and mitigation

    International Nuclear Information System (INIS)

    Jacob, M.C.; Schneider, R.E.; Finnicum, D.J.

    1993-01-01

    ABB-CE, in cooperation with the US Department of Energy, is working to develop and certify the System 80+ design, which is ABB-CE's standardized evolutionary Advanced Light Water Reactor (ALWR) design. It incorporates design enhancements based on Probabilistic Risk Assessment (PRA) insights, guidance from the EPRI's Utility Requirements Document, and US NRC's Severe Accident Policy. Major severe accident prevention and mitigation design features of the system is discussed along with its conformance to EPRI URD guidance, as applicable. Computer simulation of a best estimate severe accident scenario is presented to illustrate the acceptable containment performance of the design. It is concluded that by considering severe accident prevention and mitigation early in the design process, the System 80+ design represents a robust plant design that has low core damage frequencies, low containment conditional failure probabilities, and acceptable deterministic containment performance under severe accident conditions

  9. Central nervous system affecting drugs and road traffic accidents ...

    African Journals Online (AJOL)

    Central nervous system affecting drugs and road traffic accidents among commercial motorcyclists. ... including driving under the influence of drugs that affect the central nervous system (CNS). ... Keywords: Brain, influence, riders, substances ...

  10. Development of systems for detection, early warning, and control of pipeline leakage in drinking water distribution: a case study.

    Science.gov (United States)

    Li, Weifeng; Ling, Wencui; Liu, Suoxiang; Zhao, Jing; Liu, Ruiping; Chen, Qiuwen; Qiang, Zhimin; Qu, Jiuhui

    2011-01-01

    Water leakage in drinking water distribution systems is a serious problem for many cities and a huge challenge for water utilities. An integrated system for the detection, early warning, and control of pipeline leakage has been developed and successfully used to manage the pipeline networks in selected areas of Beijing. A method based on the geographic information system has been proposed to quickly and automatically optimize the layout of the instruments which detect leaks. Methods are also proposed to estimate the probability of each pipe segment leaking (on the basis of historic leakage data), and to assist in locating the leakage points (based on leakage signals). The district metering area (DMA) strategy is used. Guidelines and a flowchart for establishing a DMA to manage the large-scale looped networks in Beijing are proposed. These different functions have been implemented into a central software system to simplify the day-to-day use of the system. In 2007 the system detected 102 non-obvious leakages (i.e., 14.2% of the total detected in Beijing) in the selected areas, which was estimated to save a total volume of 2,385,000 m3 of water. These results indicate the feasibility, efficiency and wider applicability of this system.

  11. Impact of the filtered venting system design upon the total radioactive release in case of a severe accident and a comparison of European requirements

    International Nuclear Information System (INIS)

    Cederqvist, H.; Elisson, K.; Loewenhielm, G.; Appelgren, E.

    1991-01-01

    Filtered containment venting systems have been introduced in several nuclear power plants in Europe. The objective is to relieve the containment overpressure in a controlled way during a severe accident involving core-melt. The release of fission products when operating the venting system has been compared to that resulting from diffuse leakage from the containment. The conclusion is that the diffuse leakage of gaseous and particulate species can not be neglected in comparison to that resulting from operating the filtered containment venting system. Representative European requirements related to filtered containment venting have been analyzed and compared

  12. Bubble-vacuum system of accident localization of reference nuclear power plant with two WWER's

    International Nuclear Information System (INIS)

    Sykora, D.; Sykorova, I.

    1988-01-01

    Higher efficiency of the safety system for removing the consequences of project design accidents and higher radiation safety of a nuclear power plant with two WWER-440 units is the subject of Czechoslovak patent document 243961. The principle consists in interconnecting air chambers which are the end parts of safety systems for the two units. The air chamber is separated from the other parts of the safety system by double swing-check valves or closures. The connecting pipes of the two air chambers do not in any way reduce the reliability of the safety system thanks to their high technical safety and totally passive function. The benefits of the interconnection of the air chambers are given by the fact that it reduces maximum accident overpressure both in the air chambers and in the airtight zones. The reduction of the overpressure reduces the total leakage of radioactive substances and the radiation burden of the environment in case of a nuclear power plant accident. (Z.M.). 2 figs

  13. Improvements in the nuclear accident response system in Brazil

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Prigent, R.

    1991-01-01

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

  15. Leakage Reduction in Water Distribution Systems with Efficient Placement and Control of Pressure Reducing Valves Using Soft Computing Techniques

    Directory of Open Access Journals (Sweden)

    A. Gupta

    2017-04-01

    Full Text Available Reduction of leakages in a water distribution system (WDS is one of the major concerns of water industries. Leakages depend on pressure, hence installing pressure reducing valves (PRVs in the water network is a successful techniques for reducing leakages. Determining the number of valves, their locations, and optimal control setting are the challenges faced. This paper presents a new algorithm-based rule for determining the location of valves in a WDS having a variable demand pattern, which results in more favorable optimization of PRV localization than that caused by previous techniques. A multiobjective genetic algorithm (NSGA-II was used to determine the optimized control value of PRVs and to minimize the leakage rate in the WDS. Minimum required pressure was maintained at all nodes to avoid pressure deficiency at any node. Proposed methodology is applied in a benchmark WDS and after using PRVs, the average leakage rate was reduced by 6.05 l/s (20.64%, which is more favorable than the rate obtained with the existing techniques used for leakage control in the WDS. Compared with earlier studies, a lower number of PRVs was required for optimization, thus the proposed algorithm tends to provide a more cost-effective solution. In conclusion, the proposed algorithm leads to more favorable optimized localization and control of PRV with improved leakage reduction rate.

  16. Neural network-based expert system for severe accident management

    International Nuclear Information System (INIS)

    Klopp, G.T.; Silverman, E.B.

    1992-01-01

    This paper presents the results of the second phase of a three-phase Severe Accident Management expert system program underway at Commonwealth Edison Company (CECo). Phase I successfully demonstrated the feasibility of Artificial Neural Networks to support several of the objectives of severe accident management. Simulated accident scenarios were generated by the Modular Accident Analysis Program (MAAP) code currently in use by CECo as part of their Individual Plant Evaluations (IPE)/Accident Management Program. The primary objectives of the second phase were to develop and demonstrate four capabilities of neural networks with respect to nuclear power plant severe accident monitoring and prediction. The results of this work would form the foundation of a demonstration system which included expert system performance features. These capabilities included the ability to: (1) Predict the time available prior to support plate (and reactor vessel) failure; (2) Calculate the time remaining until recovery actions were too late to prevent core damage; (3) Predict future parameter values of each of the MAAP parameter variables; and (4) Detect simulated sensor failure and provide best-value estimates for further processing in the presence of a sensor failure. A variety of accident scenarios for the Zion and Dresden plants were used to train and test the neural network expert system. These included large and small break LOCAs as well as a range of transient events. 3 refs., 1 fig., 1 tab

  17. The computer aided education and training system for accident management

    International Nuclear Information System (INIS)

    Yoneyama, Mitsuru; Kubota, Ryuji; Fujiwara, Tadashi; Sakuma, Hitoshi

    1999-01-01

    The education and training system for Accident Management was developed by the Japanese BWR group and Hitachi Ltd. The education and training system is composed of two systems. One is computer aided instruction (CAI) education system and the education and training system with computer simulations. Both systems are designed to be executed on personal computers. The outlines of the CAI education system and the education and training system with simulator are reported below. These systems provides plant operators and technical support center staff with the effective education and training for accident management. (author)

  18. The computer aided education and training system for accident management

    International Nuclear Information System (INIS)

    Yoneyama, Mitsuru; Masuda, Takahiro; Kubota, Ryuji; Fujiwara, Tadashi; Sakuma, Hitoshi

    2000-01-01

    Under severe accident conditions of a nuclear power plant, plant operators and technical support center (TSC) staffs will be under a amount of stress. Therefore, those individuals responsible for managing the plant should promote their understanding about the accident management and operations. Moreover, it is also important to train in ordinary times, so that they can carry out accident management operations effectively on severe accidents. Therefore, the education and training system which works on personal computers was developed by Japanese BWR group (Tokyo Electric Power Co.,Inc., Tohoku Electric Power Co. ,Inc., Chubu Electric Power Co. ,Inc., Hokuriku Electric Power Co.,Inc., Chugoku Electric Power Co.,Inc., Japan Atomic Power Co.,Inc.), and Hitachi, Ltd. The education and training system is composed of two systems. One is computer aided instruction (CAI) education system and the other is education and training system with a computer simulation. Both systems are designed to execute on MS-Windows(R) platform of personal computers. These systems provide plant operators and technical support center staffs with an effective education and training tool for accident management. TEPCO used the simulation system for the emergency exercise assuming the occurrence of hypothetical severe accident, and have performed an effective exercise in March, 2000. (author)

  19. ADAM: An Accident Diagnostic,Analysis and Management System - Applications to Severe Accident Simulation and Management

    International Nuclear Information System (INIS)

    Zavisca, M.J.; Khatib-Rahbar, M.; Esmaili, H.; Schulz, R.

    2002-01-01

    The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM's severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric approaches for elements including (but not limited to) thermal hydraulics; heat transfer; fuel heatup, meltdown, and relocation; fission product release and transport; combustible gas generation and combustion; and core-concrete interaction. The overall model is defined by a relatively coarse spatial nodalization of the reactor coolant and containment systems and is advanced explicitly in time. The result is to enable much faster than real time (i.e., 100 to 1000 times faster than real time on a personal computer) applications to on-line investigations and/or accident management training. Other features of the simulation module include provision for activation of water injection, including the Engineered Safety Features, as well as other mechanisms for the assessment of accident management and recovery strategies and the evaluation of PSA success criteria. The accident diagnostics module of ADAM uses on-line access to selected plant parameters (as measured by plant sensors) to compute the thermodynamic state of the plant, and to predict various margins to safety (e.g., times to pressure vessel saturation and steam generator dryout). Rule-based logic is employed to classify the measured data as belonging to one of a number of likely scenarios based on symptoms, and a number of 'alarms' are generated to signal the state of the reactor and containment. This paper will address the features and limitations of ADAM with particular focus on accident simulation and management. (authors)

  20. Analysis of Boiler Operational Variables Prior to Tube Leakage Fault by Artificial Intelligent System

    Directory of Open Access Journals (Sweden)

    Al-Kayiem Hussain H.

    2014-07-01

    Full Text Available Steam boilers are considered as a core of any steam power plant. Boilers are subjected to various types of trips leading to shut down of the entire plant. The tube leakage is the worse among the common boiler faults, where the shutdown period lasts for around four to five days. This paper describes the rules of the Artificial Intelligent Systems to diagnosis the boiler variables prior to tube leakage occurrence. An Intelligent system based on Artificial Neural Network was designed and coded in MATLAB environment. The ANN was trained and validated using real site data acquired from coal fired power plant in Malaysia. Ninety three boiler operational variables were identified for the present investigation based on the plant operator experience. Various neural networks topology combinations were investigated. The results showed that the NN with two hidden layers performed better than one hidden layer using Levenberg-Maquardt training algorithm. Moreover, it was noticed that hyperbolic tangent function for input and output nodes performed better than other activation function types.

  1. Transient analysis and leakage detection algorithm using GA and HS algorithm for a pipeline system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Hyun; Yoo, Wan Suk; Oh, Kwang Jung; Hwang, In Sung; Oh, Jeong Eun [Pusan National University, Pusan (Korea, Republic of)

    2006-03-15

    The impact of leakage was incorporated into the transfer functions of the complex head and discharge. The impedance transfer functions for the various leaking pipeline systems were also derived. Hydraulic transients could be efficiently analyzed by the developed method. The simulation of normalized pressure variation using the method of characteristics and the impulse response method shows good agreement to the condition of turbulent flow. The leak calibration could be performed by incorporation of the impulse response method with Genetic Algorithm (GA) and Harmony Search (HS). The objective functions for the leakage detection can be made using the pressure-head response at the valve, or the pressure-head or the flow response at a certain point of the pipeline located upstream from the valve. The proposed method is not constrained by the Courant number to control the numerical dissipation of the method of characteristics. The limitations associated with the discreteness of the pipeline system in the inverse transient analysis can be neglected in the proposed method.

  2. Transient analysis and leakage detection algorithm using GA and HS algorithm for a pipeline system

    International Nuclear Information System (INIS)

    Kim, Sang Hyun; Yoo, Wan Suk; Oh, Kwang Jung; Hwang, In Sung; Oh, Jeong Eun

    2006-01-01

    The impact of leakage was incorporated into the transfer functions of the complex head and discharge. The impedance transfer functions for the various leaking pipeline systems were also derived. Hydraulic transients could be efficiently analyzed by the developed method. The simulation of normalized pressure variation using the method of characteristics and the impulse response method shows good agreement to the condition of turbulent flow. The leak calibration could be performed by incorporation of the impulse response method with Genetic Algorithm (GA) and Harmony Search (HS). The objective functions for the leakage detection can be made using the pressure-head response at the valve, or the pressure-head or the flow response at a certain point of the pipeline located upstream from the valve. The proposed method is not constrained by the Courant number to control the numerical dissipation of the method of characteristics. The limitations associated with the discreteness of the pipeline system in the inverse transient analysis can be neglected in the proposed method

  3. Research on sever accident emergency simulation system for CPR1000

    International Nuclear Information System (INIS)

    Yang Zhifei; Liao Yehong; Liang Manchun; Li Ke; Yang Jie; Chen Yali

    2015-01-01

    The enhanced capability to nuclear power plant (NPP) severe accident management and emergency response depends heavily on exercises. Since the exercise scene is usually monotonous and not realistic, and conduct of exercise has a high cost, the effect of enhancing the capability is limited. Thus, the development of a Sever Accident Emergency Simulation System (SAESS) is necessary. SAESS is able to connect NPP simulator, and simulates the process of severe accident management, personnel evacuation, the dispersion of radioactive plume, and emergency response of emergency organizations. The system helps to design several of exercise scenes and optimize the disposal strategy in different severe accidents. In addition, the system reduces the cost of emergency exercise by computer simulation, benefits the research of exercise, increases the efficiency of exercise and enhances the emergency decision-making capability. This paper introduces the design and application of SAESS. (author)

  4. Response of Soviet VVER-440 accident localization systems to overpressurization

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  5. Performance evaluation of control room HVAC and air cleaning systems under accident conditions

    International Nuclear Information System (INIS)

    Almerico, F.; Machiels, A.J.; Ornberg, S.C.; Lahti, G.P.

    1985-01-01

    In light water reactors, control rooms and technical support centers must be designed to provide habitable environments in accordance with the requirements specified in General Design Criterion 19 of Appendix A, 10 CFR Part 50. Therefore, the effectiveness of HVAC and air cleaning system designs with respect to plant operator protection has to be evaluated by the system designer. Guidance for performing the analysis has been previously given in ANSI/ASME N509-1980 as well as in presentations at past Air Cleaning Conferences. The previous work is extended and the methodology used in a generic, interactive computer program that performs Main Control Room and Technical Support Center (TSC) habitability analyses for LWR nuclear power plants is presented. For given accident concentrations of radionuclides or hazardous gases in the outdoor air intakes and plant spaces surrounding the Main Control Room (or TSC), the program models the performance of the HVAC and air cleaning system designs, and determines control room (or TSC) contaminant concentrations and plant operator protection factors. Calculated or actual duct leakage, air cleaning efficiency, and airborne contamination are taken into account. Flexibility of the model allows for the representation of most control rooms (or TSC) and associated HVAC and air cleaning system conceptual designs that have been used by the US architect/engineers. The program replaced tedious calculations to determine the effects of HVAC ductwork and equipment leakage and permits (1) parametric analyses of various HVAC system design options early in the conceptual phase of a project, and (2) analysis of the effects of leakage test results on contaminant room concentrations, and therefore operator doses

  6. Combination Effect of Hemostatic and Disinfecting Agents on Micro-leakage of Restorations Bonded with Different Bonding Systems

    Directory of Open Access Journals (Sweden)

    Farhadpour H

    2016-09-01

    Full Text Available Statement of Problem: Hemostatic agents may affect the micro-leakage of different adhesive systems. Also, chlorhexidine has shown positive effects on micro-leakage. However, their interaction effect has not been reported yet. Objectives: To evaluate the effect of contamination with a hemostatic agent on micro- leakage of total- and self-etching adhesive systems and the effect of chlorhexidine application after the removal of the hemostatic agent. Materials and Methods: Standardized Class V cavity was prepared on each of the sixty caries free premolars at the cemento-enamel junction, with the occlusal margin located in enamel and the gingival margin in dentin. Then, the specimens were randomly divided into 6 groups (n = 10 according to hemostatic agent (H contamination, chlorhexidine (CHX application, and the type of adhesive systems (Adper Single Bond and Clearfil SE Bond used. After filling the cavities with resin composite, the root apices were sealed with utility wax. Furthermore, all the surfaces, except for the restorations and 1mm from the margins, were covered with two layers of nail varnish. The teeth were immersed in a 0.5% basic fuschin dye for 24 hours, rinsed, blot-dried and sectioned longitudinally through the center of the restorations bucco- lingualy. The sections were examined using a stereomicroscope and the extension of dye penetration was analyzed according to a non-parametric scale from 0 to 3. Statistical analysis was performed using Kruskal-Wallis test and Mann-Whitney U-test. Results: While ASB group showed no micro-leakage in enamel, none of the groups showed complete elimination of micro-leakage from the dentin. Regarding micro- leakage at enamel, and dentin margins, there was no significant difference between groups 1 and 2, 1 and 3, and 2 and 3 (p > 0.05. A significantly lower micro-leakage at the enamel and dentin margins was observed in group 3, compared to group 6. No significant difference was observed between

  7. Severe Accident Management System On-line Network SAMSON

    International Nuclear Information System (INIS)

    Silverman, Eugene B.

    2004-01-01

    SAMSON is a computational tool used by accident managers in the Technical Support Centers (TSC) and Emergency Operations Facilities (EOF) in the event of a nuclear power plant accident. SAMSON examines over 150 status points monitored by nuclear power plant process computers during a severe accident and makes predictions about when core damage, support plate failure, and reactor vessel failure will occur. These predictions are based on the current state of the plant assuming that all safety equipment not already operating will fail. SAMSON uses expert systems, as well as neural networks trained with the back propagation learning algorithms to make predictions. Training on data from an accident analysis code (MAAP - Modular Accident Analysis Program) allows SAMSON to associate different states in the plant with different times to critical failures. The accidents currently recognized by SAMSON include steam generator tube ruptures (SGTRs), with breaks ranging from one tube to eight tubes, and loss of coolant accidents (LOCAs), with breaks ranging from 0.0014 square feet (1.30 cm 2 ) in size to breaks 3.0 square feet in size (2800 cm 2 ). (author)

  8. Proton-sensing transistor systems for detecting ion leakage from plasma membranes under chemical stimuli.

    Science.gov (United States)

    Imaizumi, Yuki; Goda, Tatsuro; Schaffhauser, Daniel F; Okada, Jun-Ichi; Matsumoto, Akira; Miyahara, Yuji

    2017-03-01

    The membrane integrity of live cells is routinely evaluated for cytotoxicity induced by chemical or physical stimuli. Recent progress in bioengineering means that high-quality toxicity validation is required. Here, we report a pH-sensitive transistor system developed for the continuous monitoring of ion leakage from cell membranes upon challenge by toxic compounds. Temporal changes in pH were generated with high reproducibility via periodic flushing of HepG2 cells on a gate insulator of a proton-sensitive field-effect transistor with isotonic buffer solutions with/without NH 4 Cl. The pH transients at the point of NH 4 Cl addition/withdrawal originated from the free permeation of NH 3 across the semi-permeable plasma membranes, and the proton sponge effect produced by the ammonia equilibrium. Irreversible attenuation of the pH transient was observed when the cells were subjected to a membrane-toxic reagent. Experiments and simulations proved that the decrease in the pH transient was proportional to the area of the ion-permeable pores on the damaged plasma membranes. The pH signal was correlated with the degree of hemolysis produced by the model reagents. The pH assay was sensitive to the formation of molecularly sized pores that were otherwise not measurable via detection of the leakage of hemoglobin, because the hydrodynamic radius of hemoglobin was greater than 3.1nm in the hemolysis assay. The pH transient was not disturbed by inherent ion-transporter activity. The ISFET assay was applied to a wide variety of cell types. The system presented here is fast, sensitive, practical and scalable, and will be useful for validating cytotoxins and nanomaterials. The plasma membrane toxicity and hemolysis are widely and routinely evaluated in biomaterials science and biomedical engineering. Despite the recent development of a variety of methods/materials for efficient gene/drug delivery systems to the cytosol, the methodologies for safety validation remain unchanged in

  9. [Violence and accidents among older and younger adults: evidence from the Surveillance System for Violence and Accidents (VIVA), Brazil].

    Science.gov (United States)

    Luz, Tatiana Chama Borges; Malta, Deborah Carvalho; Sá, Naíza Nayla Bandeira de; Silva, Marta Maria Alves da; Lima-Costa, Maria Fernanda

    2011-11-01

    Data from the Brazilian Surveillance System for Violence and Accidents (VIVA) in 2009 were used to examine socio-demographic characteristics, outcomes, and types of accidents and violence treated at 74 sentinel emergency services in 23 Brazilian State capitals and the Federal District. The analysis included 25,201 individuals aged > 20 years (10.1% > 60 years); 89.3% were victims of accidents and 11.9% victims of violence. Hospitalization was the outcome in 11.1% of cases. Compared to the general population, there were more men and non-white individuals among victims of accidents, and especially among victims of violence. As compared to younger adults (20-59 years), accidents and violence against elderly victims showed less association with alcohol, a higher proportion of domestic incidents, more falls and pedestrian accidents, and aggression by family members. Policies for the prevention of accidents and violence should consider the characteristics of these events in the older population.

  10. Space Vector Modulation Technique to Reduce Leakage Current of a Transformerless Three-Phase Four-Leg Photovoltaic System

    Directory of Open Access Journals (Sweden)

    F. Hasanzad

    2017-06-01

    Full Text Available Photovoltaic systems integrated to the grid have received considerable attention around the world. They can be connected to the electrical grid via galvanic isolation (transformer or without it (transformerless. Despite making galvanic isolation, low frequency transformer increases size, cost and losses. On the other hand, transformerless PV systems increase the leakage current (common-mode current, (CMC through the parasitic capacitors of the PV array. Inverter topology and switching technique are the most important parameters the leakage current depends on. As there is no need to extra hardware for switching scheme modification, it's an economical method for reducing leakage current. This paper evaluates the effect of different space vector modulation techniques on leakage current for a two-level three-phase four-leg inverter used in PV system. It proposes an efficient space vector modulation method which decreases the leakage current to below the quantity specified in VDE-0126-1-1 standard. furthermore, some other characteristics of the space vector modulation schemes that have not been significantly discussed for four-leg inverter, are considered, such as, modulation index, switching actions per period, common-mode voltage (CMV, and total harmonic distortion (THD. An extend software simulation using MATLAB/Simulink is performed to verify the effectiveness of the modulation technique.

  11. Leakage based precoding for multi-user MIMO-OFDM systems

    KAUST Repository

    Sadek, Mirette

    2011-08-01

    In downlink multi-user multiple-input multiple-output (MIMO) transmissions, several precoding schemes have been proposed to decrease interference among users. Notable among these precoding schemes is one that uses the signal-to-leakage-plus-noise ratio (SLNR) as an optimization criterion. In this paper, leveraging the efficiency of the SLNR optimization, we generalize this precoding scheme to MIMO orthogonal frequency division multiplexing (OFDM) multi-user systems where the OFDM is used to overcome the inter-symbol- interference (ISI) introduced by multipath channels. We also introduce a channel compensation technique that reconstructs the channel at the transmitter for every time instant given a significantly lower channel feedback rate by the receiver. © 2006 IEEE.

  12. Severe accident tests and development of domestic severe accident system codes

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

  13. Severe accident tests and development of domestic severe accident system codes

    International Nuclear Information System (INIS)

    2013-01-01

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

  14. Review of severe accidents and the results of accident consequence assessment in different energy systems (Contract research)

    International Nuclear Information System (INIS)

    Matsuki, Yoshio; Muramatsu, Ken

    2008-05-01

    The cases of severe accidents and the consequence assessments in different energy systems, Coal, Oil, Gas, Hydro and Nuclear, were collected, and then they were further analyzed. In this report, the information on the accidents in various energy systems were collected from the sources of the Paul Scherrer Institute (hereinafter, 'PSI') and the International Atomic Energy Agency (hereinafter, 'IAEA'). The information on the severe accidents of nuclear power plants were collected from the report of the US Presidential Commission on Catastrophic Nuclear Accidents and several relevant reports issued in the countries of the European Union, together with the reports of the PSI and the IAEA. To analyze the collected information, several parameters, which are numbers of fatalities, injuries, evacuees and the costs of the damages, were chosen to characterize those accidents in different energy systems. And then, upon the comparison of these characteristics of different accidents, the impacts of the accidents in nuclear and other energy systems were compared. Upon the results of the analysis, it is pointed out that the cost caused by the Chernobyl Accident, the severe accident in nuclear energy, tends to be higher than in the other energy systems. On the other hand, from the aspects of fatalities and injuries, it is not confirmed that the damages of the Chernobyl Accident are larger than in the other energy systems. However, it is also recognized, as the specific characteristics of the severe nuclear accident, that the impacts of the accident spread in a wider area, and stay for a longer period, in comparison with the ones in the other energy systems. (author)

  15. Cognitive systems engineering analysis of the JCO criticality accident

    International Nuclear Information System (INIS)

    Tanabe, Fumiya; Yamaguchi, Yukichi

    2000-01-01

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

  16. Microvascular dysfunction with increased vascular leakage response in mice systemically exposed to arsenic.

    Science.gov (United States)

    Chen, Shih-Chieh; Huang, Shin-Yin; Lu, Chi-Yu; Hsu, Ya-Hung; Wang, Dean-Chuan

    2014-09-01

    The mechanisms underlying cardiovascular disease induced by arsenic exposure are not completely understood. The objectives of this study were to investigate whether arsenic-fed mice have an increased vascular leakage response to vasoactive agents and whether enhanced type-2 protein phosphatase (PP2A) activity is involved in mustard oil-induced leakage. ICR mice were fed water or sodium arsenite (20 mg/kg) for 4 or 8 weeks. The leakage response to vasoactive agents was quantified using the Evans blue (EB) technique or vascular labeling with carbon particles. Increased EB leakage and high density of carbon-labeled microvessels were detected in arsenic-fed mice treated with mustard oil. Histamine induced significantly higher vascular leakage in arsenic-fed mice than in water-fed mice. Pretreatment with the PP2A inhibitor okadaic acid or the neurokinin 1 receptor (NK1R) blocker RP67580 significantly reduced mustard oil-induced vascular leakage in arsenic-fed mice. The protein levels of PP2Ac and NK1R were similar in both groups. PP2A activity was significantly higher in the arsenic-fed mice compared with the control group. These findings indicate that microvessels generally respond to vasoactive agents, and that the increased PP2A activity is involved in mustard oil-induced vascular leakage in arsenic-fed mice. Arsenic may initiate endothelial dysfunction, resulting in vascular leakage in response to vasoactive agents.

  17. A severity-based quantification of data leakages in database systems

    NARCIS (Netherlands)

    Vavilis, S.; Petkovic, M.; Zannone, N.

    2016-01-01

    The detection and handling of data leakages is becoming a critical issue for organizations. To this end, data leakage solutions are usually employed by organizations to monitor network traffic and the use of portable storage devices. However, these solutions often produce a large number of alerts,

  18. Development of an Accident Reproduction Simulator System Using a Hemodialysis Extracorporeal Circulation System

    OpenAIRE

    Nishite, Yoshiaki; Takesawa, Shingo

    2016-01-01

    Background Accidents that occur during dialysis treatment are notified to the medical staff via alarms raised by the dialysis apparatus. Similar to such real accidents, apparatus activation or accidents can be reproduced by simulating a treatment situation. An alarm that corresponds to such accidents can be utilized in the simulation model. Objectives The aim of this study was to create an extracorporeal circulation system (herein...

  19. Methods for air cleaning system design and accident analysis

    International Nuclear Information System (INIS)

    Gregory, W.S.; Nichols, B.D.

    1987-01-01

    This paper describes methods, in the form of a handbook and five computer codes, that can be used for nuclear facility air cleaning system design and accident analysis. Four of the codes were developed primarily at the Los Alamos National Laboratory, and one was developed in France. Tools such as these are used to design ventilation systems in the mining industry but do not seem to be commonly used in the nuclear industry. For example, the Nuclear Air Cleaning Handbook is an excellent design reference, but it fails to include information on computer codes that can be used to aid in the design process. These computer codes allow the analyst to use the handbook information to form all the elements of a complete system design. Because these analysis methods are in the form of computer codes they allow the analyst to investigate many alternative designs. In addition, the effects of many accident scenarios on the operation of the air cleaning system can be evaluated. These tools originally were intended for accident analysis, but they have been used mostly as design tools by several architect-engineering firms. The Cray, VAX, and personal computer versions of the codes, an accident analysis handbook, and the codes availability will be discussed. The application of these codes to several design operations of nuclear facilities will be illustrated, and their use to analyze the effect of several accident scenarios also will be described

  20. Analytical scaling relations to evaluate leakage and intrusion in intermittent water supply systems

    Science.gov (United States)

    Slocum, Alexander H.; Whittle, Andrew J.

    2018-01-01

    Intermittent water supplies (IWS) deliver piped water to one billion people; this water is often microbially contaminated. Contaminants that accumulate while IWS are depressurized are flushed into customers’ homes when these systems become pressurized. In addition, during the steady-state phase of IWS, contaminants from higher-pressure sources (e.g., sewers) may continue to intrude where pipe pressure is low. To guide the operation and improvement of IWS, this paper proposes an analytic model relating supply pressure, supply duration, leakage, and the volume of intruded, potentially-contaminated, fluids present during flushing and steady-state. The proposed model suggests that increasing the supply duration may improve water quality during the flushing phase, but decrease the subsequent steady-state water quality. As such, regulators and academics should take more care in reporting if water quality samples are taken during flushing or steady-state operational conditions. Pipe leakage increases with increased supply pressure and/or duration. We propose using an equivalent orifice area (EOA) to quantify pipe quality. This provides a more stable metric for regulators and utilities tracking pipe repairs. Finally, we show that the volume of intruded fluid decreases in proportion to reductions in EOA. The proposed relationships are applied to self-reported performance indicators for IWS serving 108 million people described in the IBNET database and in the Benchmarking and Data Book of Water Utilities in India. This application shows that current high-pressure, continuous water supply targets will require extensive EOA reductions. For example, in order to achieve national targets, utilities in India will need to reduce their EOA by a median of at least 90%. PMID:29775462

  1. Analytical scaling relations to evaluate leakage and intrusion in intermittent water supply systems.

    Science.gov (United States)

    Taylor, David D J; Slocum, Alexander H; Whittle, Andrew J

    2018-01-01

    Intermittent water supplies (IWS) deliver piped water to one billion people; this water is often microbially contaminated. Contaminants that accumulate while IWS are depressurized are flushed into customers' homes when these systems become pressurized. In addition, during the steady-state phase of IWS, contaminants from higher-pressure sources (e.g., sewers) may continue to intrude where pipe pressure is low. To guide the operation and improvement of IWS, this paper proposes an analytic model relating supply pressure, supply duration, leakage, and the volume of intruded, potentially-contaminated, fluids present during flushing and steady-state. The proposed model suggests that increasing the supply duration may improve water quality during the flushing phase, but decrease the subsequent steady-state water quality. As such, regulators and academics should take more care in reporting if water quality samples are taken during flushing or steady-state operational conditions. Pipe leakage increases with increased supply pressure and/or duration. We propose using an equivalent orifice area (EOA) to quantify pipe quality. This provides a more stable metric for regulators and utilities tracking pipe repairs. Finally, we show that the volume of intruded fluid decreases in proportion to reductions in EOA. The proposed relationships are applied to self-reported performance indicators for IWS serving 108 million people described in the IBNET database and in the Benchmarking and Data Book of Water Utilities in India. This application shows that current high-pressure, continuous water supply targets will require extensive EOA reductions. For example, in order to achieve national targets, utilities in India will need to reduce their EOA by a median of at least 90%.

  2. SU-E-T-774: Use of a Scintillator-Mirror-Camera System for the Measurement of MLC Leakage Radiation with the CyberKnife M6 System

    Energy Technology Data Exchange (ETDEWEB)

    Goggin, L; Kilby, W; Noll, M; Maurer, C [Accuray Inc, Sunnyvale, CA (United States)

    2015-06-15

    Purpose: A technique using a scintillator-mirror-camera system to measure MLC leakage was developed to provide an efficient alternative to film dosimetry while maintaining high spatial resolution. This work describes the technique together with measurement uncertainties. Methods: Leakage measurements were made for the InCise™ MLC using the Logos XRV-2020A device. For each measurement approximately 170 leakage and background images were acquired using optimized camera settings. Average background was subtracted from each leakage frame before filtering the integrated leakage image to replace anomalous pixels. Pixel value to dose conversion was performed using a calibration image. Mean leakage was calculated within an ROI corresponding to the primary beam, and maximum leakage was determined by binning the image into overlapping 1mm x 1mm ROIs. 48 measurements were performed using 3 cameras and multiple MLC-linac combinations in varying beam orientations, with each compared to film dosimetry. Optical and environmental influences were also investigated. Results: Measurement time with the XRV-2020A was 8 minutes vs. 50 minutes using radiochromic film, and results were available immediately. Camera radiation exposure degraded measurement accuracy. With a relatively undamaged camera, mean leakage agreed with film measurement to ≤0.02% in 92% cases, ≤0.03% in 100% (for maximum leakage the values were 88% and 96%) relative to reference open field dose. The estimated camera lifetime over which this agreement is maintained is at least 150 measurements, and can be monitored using reference field exposures. A dependency on camera temperature was identified and a reduction in sensitivity with distance from image center due to optical distortion was characterized. Conclusion: With periodic monitoring of the degree of camera radiation damage, the XRV-2020A system can be used to measure MLC leakage. This represents a significant time saving when compared to the traditional

  3. Application of the SPEEDI system to the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Chino, Masamichi; Ishikawa, Hirohiko; Yamazawa, Hiromi; Moriuchi, Shigeru

    1986-10-01

    The SPEEDI system is a computational code system to predict the radiological dose due to the plume released in a nuclear accident in Japan. This paper describes the SPEEDI's application to the Chernobyl reactor accident for the estimation of the movement of plume and the release rate of radioactive nuclides into the environment. The predicted results on the movement of plume agreed well with the monitoring data in Europe. The estimated results on the release rate showed that half of the noble gas inventory, about 5 % of the iodine inventory and about 3 % of the cesium inventory are released into the environment within 24 hours. (author)

  4. A systems approach to the management of radiation accidents

    International Nuclear Information System (INIS)

    Richter, L.L.; Berk, H.W.; Teates, C.D.; Larkham, N.E.; Friesen, E.J.; Edlich, R.F.

    1980-01-01

    Management of radiation accident patients should have a multidisciplinary approach that includes all health professionals as well as members of public safety agencies. Emergency plans for radiation accidents include detection of the ionizing radiation, patient evacuation, resuscitation, and decontamination. The resuscitated patient should be transported to a radiation control area located outside but adjacent to the emergency department. Ideally this area is accessed through an entrance separate from that used for the main flow of daily emergency department patients. The hospital staff, provided with protective clothing, dosimeters, and preprinted guidelines, continues the resuscitation and definitive care of the patient. This system approach to the management of radiation accidents may be tailored to meet the specific needs of other emergency medical systems

  5. Post-accident monitoring systems in Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Suriya Murthy, N.; Sivasailanathan, Vidhya; Ananth, Allu; Roy, Kallol

    2018-01-01

    PFBR is a 500 MW(e) MOX fueled and sodium cooled fast reactor (SFR) under advanced stage of commissioning at Kalpakkam. Currently, the main vessel is preheated and sodium has been charged into two secondary loops that are operated in recirculation mode. In order to characterize the radiation field and contamination, the workplace monitoring is undertaken using installed monitors that are commissioned and made operational. This helps to ensure radiological protection during normal operating conditions. On the other hand, radiological monitoring in emergency conditions is quite different. For undertaking the mitigative accident management, a set of specialized nuclear instruments called post-accident monitoring systems (PAMS) which include radiation monitors are stipulated. The Fukushima Daiichi accident emphasized the importance and need for reliable accident monitoring instrumentation to indicate the safety functions during the progression and aftermath of accident in NPP. In PFBR, the PAMS are integrated with other monitoring systems in design stage itself to manage the measurements and indicating the safety functions for implementing EOP and SAMG

  6. Acoustic emission monitoring of leakage through seal-plug in PHWR systems

    Energy Technology Data Exchange (ETDEWEB)

    Jha, S K; Badgujar, B P; Goswami, G L [Bhabha Atomic Research Centre, Bombay (India). Atomic Fuels Div.; Patel, R J; Bhattacharya, S; Agrawal, R G [Bhabha Atomic Research Centre, Mumbai (India). Refuelling Technology Division

    1994-12-31

    Acoustic Emission (AE) technique is being developed as an in-service inspection tool for monitoring the leakage through seal-plugs in Pressurised Heavy Water Reactors (PHWRs). Time as well as frequency domain analysis have been utilised during the experiment carried out at Bhabha Atomic Research Centre (BARC) using test set up simulating the pressure and temperature conditions. The work involved were to determine the temperature profile on end-fitting, effect of pressure and temperature on leakage etc. This paper discusses various relationships like signal-level vs. pressure, frequency spectrum of signal, signal-level vs. leakage based on the above experimental work. (author). 4 refs., 6 figs.

  7. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T.; Rollstin, J.A.; Chanin, D.I.

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs

  8. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    Chanin, D.I.; Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems

  9. MELCOR Accident Consequence Code System (MACCS)

    International Nuclear Information System (INIS)

    Rollstin, J.A.; Chanin, D.I.; Jow, H.N.

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projections, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management

  10. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T. (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA))

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.

  11. MELCOR Accident Consequence Code System (MACCS)

    Energy Technology Data Exchange (ETDEWEB)

    Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian (Sandia National Labs., Albuquerque, NM (USA))

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

  12. The Use of Low-frequency Active Channel Signals in an Information Leakage Detection and Prevention Systems

    Directory of Open Access Journals (Sweden)

    A. V. Mamaev

    2011-06-01

    Full Text Available This article describes a solution for the protection of the temporary lifting of control over the target machine, while using information leakage detection and prevention systems. It is proposed to use a specially designed channel for alert’s signals through the computer’s power supply.

  13. The Hardware and Software Implementation of Low-Frequency Active Channel Signals in an Information Leakage Detection and Prevention Systems

    Directory of Open Access Journals (Sweden)

    A. V. Mamaev

    2011-12-01

    Full Text Available This article discusses a new way of developing a special channel for the alarms, through computer’s power supply network, to solve the problem of protection from removal of the temporary control over the victim’s machine, using information leakage detection and prevention systems.

  14. [A monitoring system for work-related accidents in Piracicaba, São Paulo, Brazil].

    Science.gov (United States)

    Cordeiro, Ricardo; Vilela, Rodolfo Andrade Gouveia; de Medeiros, Maria Angélica Tavares; Gonçalves, Cláudia Giglio de Oliveira; Bragantini, Clarice Aparecida; Varolla, Antenor J; Celso, Stephan

    2005-01-01

    The authors report on the development of a work accident monitoring system in Piracicaba, São Paulo State, Brazil, with the following characteristics: information feeding the system is obtained in real time directly from accident treatment centers; the system has universal monitoring, covering all work-related accidents in Piracicaba, regardless of the nature of the worker's employment conditions, place of work, or place of residence; health surveillance and promotion of health initiatives are triggered by identification of sentinel events; spatial distribution analysis of work-related accidents is a basic tool in designing accident awareness strategies and accident prevention policies. The system was implemented in November 2003 and by October 2004 had identified 5,320 work-related accidents, or a 3.8% annual proportional incidence of work-related accidents in the municipal area. We illustrate spatial analysis of registered work-related accidents and present a detailed investigation of one example of a serious accident.

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  16. Comparison of the Standard of Air Leakage in Current Metal Duct Systems in the World

    Science.gov (United States)

    Di, Yuhui; Wang, Jiqian; Feng, Lu; Li, Xingwu; Hu, Chunlin; Shi, Junshe; Xu, Qingsong; Qiao, Leilei

    2018-01-01

    Based on the requirements of air leakage of metal ducts in Chinese design standards, technical measures and construction standards, this paper compares the development history, the classification of air pressure levels and the air tightness levels of air leakage standards of current Chinese and international metal ducts, sums up the differences, finds shortage by investigating the design and construction status and access to information, and makes recommendations, hoping to help the majority of engineering and technical personnel.

  17. Influence of rain water leakage on the hygrothermal performance of exterior insulation systems

    OpenAIRE

    Künzel, H.M.; Zirkelbach, D.

    2008-01-01

    This investigation shows that ETICS on light weight structures pose no moisture problem in cold and moderate climates when the detailing of joints and openings is well done, i.e. there is no rain water leakage. This holds for all locations investigated. However, if water leakage cannot be excluded completely and therefore the North-American Standard assumptions - penetration of 1% of the driving rain load - are applied this picture changes, making the drying potential an essential feature. Th...

  18. A case of death of the driver due to environmental asphyxia by liquid nitrogen leakage in the cabin of the car during a road accident

    Science.gov (United States)

    Raczkowska, Zuzanna; Samojłowicz, Dorota

    2013-01-01

    Nitrogen causes environmental asphyxia by displacing oxygen in the air leading to death. The study presents a case of a death of a driver death who was transporting flasks with liquid nitrogen that depressurized during an accident. The mechanism and cause of death were determined based on the result of the autopsy and histopathologic examination. The authors emphasize the relevance of accident scene inspection during establishing the cause of death in similar cases.

  19. Implementation of an Analytical Model for Leakage Neutron Equivalent Dose in a Proton Radiotherapy Planning System

    Energy Technology Data Exchange (ETDEWEB)

    Eley, John [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, 1515 Holcombe Blvd., Houston, TX 77030 (United States); Graduate School of Biomedical Sciences, The University of Texas, 6767 Bertner Ave., Houston, TX 77030 (United States); Newhauser, Wayne, E-mail: newhauser@lsu.edu [Department of Physics and Astronomy, Louisiana State University and Agricultural and Mechanical College, 202 Nicholson Hall, Tower Drive, Baton Rouge, LA 70803 (United States); Mary Bird Perkins Cancer Center, 4950 Essen Lane, Baton Rouge, LA 70809 (United States); Homann, Kenneth; Howell, Rebecca [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, 1515 Holcombe Blvd., Houston, TX 77030 (United States); Graduate School of Biomedical Sciences, The University of Texas, 6767 Bertner Ave., Houston, TX 77030 (United States); Schneider, Christopher [Department of Physics and Astronomy, Louisiana State University and Agricultural and Mechanical College, 202 Nicholson Hall, Tower Drive, Baton Rouge, LA 70803 (United States); Mary Bird Perkins Cancer Center, 4950 Essen Lane, Baton Rouge, LA 70809 (United States); Durante, Marco; Bert, Christoph [GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany)

    2015-03-11

    Equivalent dose from neutrons produced during proton radiotherapy increases the predicted risk of radiogenic late effects. However, out-of-field neutron dose is not taken into account by commercial proton radiotherapy treatment planning systems. The purpose of this study was to demonstrate the feasibility of implementing an analytical model to calculate leakage neutron equivalent dose in a treatment planning system. Passive scattering proton treatment plans were created for a water phantom and for a patient. For both the phantom and patient, the neutron equivalent doses were small but non-negligible and extended far beyond the therapeutic field. The time required for neutron equivalent dose calculation was 1.6 times longer than that required for proton dose calculation, with a total calculation time of less than 1 h on one processor for both treatment plans. Our results demonstrate that it is feasible to predict neutron equivalent dose distributions using an analytical dose algorithm for individual patients with irregular surfaces and internal tissue heterogeneities. Eventually, personalized estimates of neutron equivalent dose to organs far from the treatment field may guide clinicians to create treatment plans that reduce the risk of late effects.

  20. Systemic accident analysis: examining the gap between research and practice.

    Science.gov (United States)

    Underwood, Peter; Waterson, Patrick

    2013-06-01

    The systems approach is arguably the dominant concept within accident analysis research. Viewing accidents as a result of uncontrolled system interactions, it forms the theoretical basis of various systemic accident analysis (SAA) models and methods. Despite the proposed benefits of SAA, such as an improved description of accident causation, evidence within the scientific literature suggests that these techniques are not being used in practice and that a research-practice gap exists. The aim of this study was to explore the issues stemming from research and practice which could hinder the awareness, adoption and usage of SAA. To achieve this, semi-structured interviews were conducted with 42 safety experts from ten countries and a variety of industries, including rail, aviation and maritime. This study suggests that the research-practice gap should be closed and efforts to bridge the gap should focus on ensuring that systemic methods meet the needs of practitioners and improving the communication of SAA research. Copyright © 2013 Elsevier Ltd. All rights reserved.

  1. A System Supporting the Analysis of Motorway Traffic Accidents

    Directory of Open Access Journals (Sweden)

    Davide Anghinolfi

    2015-12-01

    Full Text Available This work presents a business intelligence tool for monitoring traffic accidents on motorways and supporting decisions relevant to road safety. The system manages information on road characteristics, traffic accidents and traffic volumes and produces reports for monitoring the evolution of key performance indicators for road safety, supporting decisions on actions for risk mitigation and safety improvements for road users. The paper illustrates the different types of analyses performed by the system. Pattern based analysis is used to evaluate safety performance indicators for the road sections matching defined patterns. Two different road segmentation algorithms, used to identify the most critical road sections according to various severity indicators, are presented and discussed. Differential analysis compares the value of selected severity indicators before and after the implementation of an intervention on a road. Finally, a graphical user interface allows the accident locations to be visualized and accidents with specific characteristics to be highlighted. The system was evaluated on the data collected between 2009 and 2011 for the A15 motorway in Italy, connecting Parma to La Spezia.

  2. URBAN TRAFFIC ACCIDENT ANALYSIS BY USING GEOGRAPHIC INFORMATION SYSTEM

    Directory of Open Access Journals (Sweden)

    Meltem SAPLIOĞLU

    2006-03-01

    Full Text Available In recent years, traffic accidents that cause more social and economic losses than that of natural disasters,have become a national problem in Turkey. To solve this problem and to reduce the casualties, road safety programs are tried to be developed. It is necessary to develop the most effective measures with low investment cost due to limited budgets allocated to such road safety programs. The most important program is to determine dangerous locations of traffic accidents and to improve these sections from the road safety view point. New Technologies are driving a cycle of continuous improvement that causes rapid changes in the traffic engineering and any engineering services within it. It is obvious that this developed services will be the potential for forward-thinking engineering studies to take a more influence role. In this study, Geographic Information System (GIS was used to identify the hazardous locations of traffic accidents in Isparta. Isparta city map was digitized by using Arcinfo 7.21. Traffic accident reports occurred between 1998-2002 were obtained from Directory of Isparta Traffic Region and had been used to form the database. Topology was set up by using Crash Diagrams and Geographic Position Reference Systems. Tables are formed according to the obtained results and interpreted.

  3. Development of Integrated Evaluation System for Severe Accident Management

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Ha; Kim, K. R.; Park, S. H.; Park, S. Y.; Park, J. H.; Song, Y. M.; Ahn, K. I.; Choi, Y

    2007-06-15

    The objective of the project is twofold. One is to develop a severe accident database (DB) for the Korean Standard Nuclear Power plant (OPR-1000) and a DB management system, and the other to develop a localized computer code, MIDAS (Multi-purpose IntegrateD Assessment code for Severe accidents). The MELCOR DB has been constructed for the typical representative sequences to support the previous MAAP DB in the previous phase. The MAAP DB has been updated using the recent version of MAAP 4.0.6. The DB management system, SARD, has been upgraded to manage the MELCOR DB in addition to the MAAP DB and the network environment has been constructed for many users to access the SARD simultaneously. The integrated MIDAS 1.0 has been validated after completion of package-wise validation. As the current version of MIDAS cannot simulate the anticipated transient without scram (ATWS) sequence, point-kinetics model has been implemented. Also the gap cooling phenomena after corium relocation into the RPV can be modeled by the user as an input parameter. In addition, the subsystems of the severe accident graphic simulator are complemented for the efficient severe accident management and the engine of the graphic simulator was replaced by the MIDAS instead of the MELCOR code. For the user's convenience, MIDAS input and output processors are upgraded by enhancing the interfacial programs.

  4. Development of Integrated Evaluation System for Severe Accident Management

    International Nuclear Information System (INIS)

    Kim, Dong Ha; Kim, K. R.; Park, S. H.; Park, S. Y.; Park, J. H.; Song, Y. M.; Ahn, K. I.; Choi, Y.

    2007-06-01

    The objective of the project is twofold. One is to develop a severe accident database (DB) for the Korean Standard Nuclear Power plant (OPR-1000) and a DB management system, and the other to develop a localized computer code, MIDAS (Multi-purpose IntegrateD Assessment code for Severe accidents). The MELCOR DB has been constructed for the typical representative sequences to support the previous MAAP DB in the previous phase. The MAAP DB has been updated using the recent version of MAAP 4.0.6. The DB management system, SARD, has been upgraded to manage the MELCOR DB in addition to the MAAP DB and the network environment has been constructed for many users to access the SARD simultaneously. The integrated MIDAS 1.0 has been validated after completion of package-wise validation. As the current version of MIDAS cannot simulate the anticipated transient without scram (ATWS) sequence, point-kinetics model has been implemented. Also the gap cooling phenomena after corium relocation into the RPV can be modeled by the user as an input parameter. In addition, the subsystems of the severe accident graphic simulator are complemented for the efficient severe accident management and the engine of the graphic simulator was replaced by the MIDAS instead of the MELCOR code. For the user's convenience, MIDAS input and output processors are upgraded by enhancing the interfacial programs

  5. Intelligent system for accident identification in NPP

    International Nuclear Information System (INIS)

    Hernandez, J.L.

    1998-01-01

    Accidental situations in NPP are great concern for operators, the facility, regulatory bodies and the environmental. This work proposes a design of intelligent system aimed to assist the operator in the process of decision making initiator events with higher relative contribution to the reactor core damage occur. The intelligent System uses the results of the pre-operational Probabilistic safety Assessment and the Thermal hydraulic Safety Analysis of the NPP Juragua as source for building its knowledge base. The nucleus of the system is presented as a design of an intelligent hybrid from the combination of the artificial intelligence techniques fuzzy logic and artificial neural networks. The system works with variables from the process of the first circuit, second circuit and the containment and it is presented as a model for the integration of safety analyses in the process of decision making by the operator when tackling with accidental situations

  6. Virtual system concept aiming at prevention of troubles and accidents

    International Nuclear Information System (INIS)

    Uchimoto, Tetsuya; Takagi, Toshiyuki

    2001-01-01

    A main impediment to optimization of the plant maintenance is the fact that we can not predict when and how troubles are introduced in a plant. Having regard to the point, the authors propose a 'virtual system' concept for prevention and prediction of accidents in plants. The virtual system is a system constructed in computers and it evaluates responses to various loads of the object system. The authors introduce the resistance to loads and the testing availability as key parameters characterizing object sub-systems and place their evaluation as the first step of construction of the virtual system. (author)

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

    National Research Council Canada - National Science Library

    Qureshi, Zahid H

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Takahashi, Fumiaki; Shigemori, Yuji; Seki, Akiyuki

    2009-07-01

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

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

    International Nuclear Information System (INIS)

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

    1990-12-01

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

  10. System of accidents notification: the ROSIS experience

    International Nuclear Information System (INIS)

    Coffey, M.; Cunningham, J.

    2009-01-01

    ROSIS is short for 'Radiation Oncology Safety Information System' and it is a voluntary web-based safety information database for Radiotherapy. The system is based on professional front-line staff in radiotherapy clinics reporting incidents and corrective actions over the Internet to a database. On a six years period, 120 health establishments registered more than 1200 events. Almost 98% of statements concern external radiotherapy. The reports can be consulted on the Internet site (www.clin.radfys.lu.se/) besides, a mini training to the risk management in the field of radiotherapy based on the Rosis data has been finalized and proposed for six years. (N.C.)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  12. Risk factors for the leakage of chemotherapeutic agents into systemic circulation after transcatheter arterial chemoembolization of hepatocellular carcinoma

    Directory of Open Access Journals (Sweden)

    Ming-Yen Hsieh

    2011-10-01

    Full Text Available This prospective study was to investigate the possible risk factors for the leakage of chemotherapeutic agent into the systemic circulation after transcatheter arterial chemoembolization (TACE of hepatocellular carcinoma (HCC. Peripheral plasma concentrations of chemotherapeutic agents were determined at 1 hour and 72 hours after TACE by high-performance liquid chromatography in 53 patients. HCC were divided into three types namely single nodule (<5 cm, multiple nodules (all <5 cm, and main nodule measuring 5 cm or more. Forty-four patients (83% showed detectable chemotherapeutic concentrations within 72 hours after TACE. Patients with single nodular-type HCC had lower incidence of detectable plasma chemotherapeutic agents after TACE than the other two groups (all p<0.05. The injected doses of lipiodol, epirubicin, and mitomycin C were lower in patients without detection than in patients with detectable chemotherapeutic agents (all p<0.05. Multivariate logistic regression showed that tumor type and injected dose of lipiodol were two independent risk factors for the leakage of mitomycin C at 1 hour after TACE (all p<0.05, and the injected dose of mitomycin C was the risk factor for the leakage of epirubicin at 1 hour after TACE (p<0.05. In conclusion, multiple nodular type and large nodule measuring 5 cm or more have a risk of leakage of mitomycin C after TACE. Injected dose of lipiodol and mitomycin C as risk factor for the leakage of mitomycin C and epirubicin respectively may be because of competition of their injected volume within the limited space of target.

  13. Development of a system of computer codes for severe accident analyses and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1991-12-15

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy.

  14. Development of a system of computer codes for severe accident analyses and its applications

    International Nuclear Information System (INIS)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan

    1991-12-01

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy

  15. Sub-assembly accident protection instrumentation systems

    International Nuclear Information System (INIS)

    Vaughan, G.J.; Lunt, A.R.W.; Evans, N.J.; Lawrence, L.A.J.

    1982-01-01

    The possibility of an incident in a sub-assembly progressing to the stage at which the whole core may be at hazard has to be guarded against. It is proposed that for CDFR specific instrumentation will be provided to protect against this incident. Three such systems are described, these are: Acoustic Boiling Noise Detection, Burst Pin Detection and Individual Sub-Assembly Thermocouple (ISAT) monitoring. In the ISAT case, multiplexers and microprocessors are employed, using novel techniques to ensure failure-to-safety. The role of these systems and the implementation of them in the reactor design are also considered. It is concluded that sufficient protection can be provided for both core and breeder sub-assemblies

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  17. A radiological accident consequence assessment system for Hong Kong

    International Nuclear Information System (INIS)

    Wong, M.C.; Lam, H.K.

    1993-01-01

    An account is given of the Hong Kong Radiological Accident Consequence Assessment System which would be used to assess the potential consequences of an emergency situation involving atmospheric release of radioactive material. The system has the capability to acquire real-time meteorological information from the Observatory's network of automatic stations, synoptic stations in the nearby region as well as forecast data from numerical prediction models. The system makes use of these data to simulate the transport and dispersion of the released radioactive material. The effectiveness of protective action on the local population is also modeled. The system serves as a powerful aid in the protective action recommendation processes

  18. A collimated detection system for assessing leakage dose from medical linear accelerators at the patient plane.

    Science.gov (United States)

    Lonski, P; Taylor, M L; Franich, R D; Kron, T

    2014-03-01

    Leakage radiation from linear accelerators can make a significant contribution to healthy tissue dose in patients undergoing radiotherapy. In this work thermoluminescent dosimeters (LiF:Mg,Cu,P TLD chips) were used in a focused lead cone loaded with TLD chips for the purpose of evaluating leakage dose at the patient plane. By placing the TLDs at one end of a stereotactic cone, a focused measurement device is created; this was tested both in and out of the primary beam of a Varian 21-iX linac using 6 MV photons. Acrylic build up material of 1.2 cm thickness was used inside the cone and measurements made with either one or three TLD chips at a given distance from the target. Comparing the readings of three dosimeters in one plane inside the cone offered information regarding the orientation of the cone relative to a radiation source. Measurements in the patient plane with the linac gantry at various angles demonstrated that leakage dose was approximately 0.01% of the primary beam out of field when the cone was pointed directly towards the target and 0.0025% elsewhere (due to scatter within the gantry). No specific 'hot spots' (e.g., insufficient shielding or gaps at abutments) were observed. Focused cone measurements facilitate leakage dose measurements from the linac head directly at the patient plane and allow one to infer the fraction of leakage due to 'direct' photons (along the ray-path from the bremsstrahlung target) and that due to scattered photons.

  19. A collimated detection system for assessing leakage dose from medical linear accelerators at the patient plane

    International Nuclear Information System (INIS)

    Lonski, P.; Kron, T.; Taylor, M.L.; Franich, R.D.

    2014-01-01

    Leakage radiation from linear accelerators can make a significant contribution to healthy tissue dose in patients undergoing radiotherapy. In this work thermoluminescent dosimeters (LiF:Mg,Cu,P TLD chips) were used in a focused lead cone loaded with TLD chips for the purpose of evaluating leakage dose at the patient plane. By placing the TLDs at one end of a stereotactic cone, a focused measurement device is created; this was tested both in and out of the primary beam of a Varian 21-iX linac using 6 MV photons. Acrylic build up material of 1.2 cm thickness was used inside the cone and measurements made with either one or three TLD chips at a given distance from the target. Comparing the readings of three dosimeters in one plane inside the cone offered information regarding the orientation of the cone relative to a radiation source. Measurements in the patient plane with the linac gantry at various angles demonstrated that leakage dose was approximately 0.01 % of the primary beam out of field when the cone was pointed directly towards the target and 0.0025 % elsewhere (due to scatter within the gantry). No specific ‘hot spots’ (e.g., insufficient shielding or gaps at abutments) were observed. Focused cone measurements facilitate leakage dose measurements from the linac head directly at the patient plane and allow one to infer the fraction of leakage due to ‘direct’ photons (along the ray-path from the bremsstrahlung target) and that due to scattered photons.

  20. Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gilles Youinou; R. Sonat Sen

    2013-09-01

    The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy – UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

  1. Cone-morse implant connection system significantly reduces bacterial leakage between implant and abutment: an in vitro study.

    Science.gov (United States)

    Baj, A; Bolzoni, A; Russillo, A; Lauritano, D; Palmieri, A; Cura, F; Silvestre, F J; Giannì, A B

    2017-01-01

    Osseointegrated implants are very popular dental treatments today in the world. In osseointegrated implants, the occlusal forces are transmitted from prosthesis through an abutment to a dental implant. The abutment is connected to the implant by mean of a screw. A screw is the most used mean for connecting an implant to an abutment. Frequently the screws break and are lost. There is an alternative to screw retained abutment systems: the cone-morse connection (CMC). The CMC, thanks to the absence of the abutment screw, guarantees no micro-gaps, no micro-movements, and a reduction of bacterial leakage between implant and abutment. As P. gingivalis and T. forsythia penetration might have clinical relevance, it was the purpose of this investigation to evaluate molecular leakage of these two bacteria in a new CMC implants systems (Leone Spa®, Florence, Italy). To identify the capability of the implant to protect the internal space from the external environment, the passage of genetically modified Escherichia coli across implant-abutment interface was evaluated. Four cone-morse Leone implants (Leone® Spa, Florence, Italy) were immerged in a bacterial culture for 24 h and bacteria amount was then measured inside implant-abutment interface with Real-time PCR. Bacteria were detected inside all studied implants, with a median percentage of 3% for P. gingivalis and 4% for T. forsythia. Cone-morse connection implant system has very low bacterial leakage percentage and is similar to one-piece implants.

  2. New developments in the surveillance and diagnostics technology for vibration, structure-borne sound and leakage monitoring systems

    International Nuclear Information System (INIS)

    Gloth, Gerrit

    2009-01-01

    Monitoring and diagnostic systems are of main importance for a safe and efficient operation of nuclear power plants. The author describes new developments with respect to vibration monitoring with a functional extension in the time domain for den secondary circuit, the development of a local system for the surveillance of rotating machines, the structure-borne sound monitoring with improvement of event analysis, esp. the loose part locating, leakage monitoring with a complete system for humidity measurement, and the development of a common platform for all monitoring and diagnostic systems, that allows an efficient access for comparison and cross references.

  3. Improvement of the following accident dose assessment system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Enn Han; Han, Moon Hee; Suh, Kyung Suk; Hwang, Won Tae; Choi, Young Gil [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1999-12-15

    The FADAS has been updates for calculating the real-time wind fields continuously at the nuclear sites in Korea. The system has been constructed to compute the wind fields using its own process for the dummy meteorological data, and dose not effect on the overall wind field module. If the radioactive materials are released into the atmosphere in real situation, the calculations of wind fields and exposure dose in the previous FADAS are performed in the case of the recognition of the above situation in the source term evaluation module. The current version of FADAS includes the program for evaluating the effect of the predicted accident and the assumed scenario together. The dose assessment module is separated into the real-time and the supposed accident respectively.

  4. Iodine removal in containment filtered venting system during nuclear accident

    International Nuclear Information System (INIS)

    Bera, Subrata; Deo, Anuj Kumar; Nagrale, D.B.; Paul, U.K.; Prasad, M.; Gaikwad, A.J.

    2015-01-01

    Post Fukushima nuclear accident, containment filtered venting system is being introduced in Indian nuclear power plant to strengthen the defense in depth safety barrier by depressurizing the containment building along with minimization of radioactivity release to environment during a severe accident. Radioactive iodine is one of the major contributors to radiation dose during early release phase of a severe accident. Physical and Chemical form of iodine and iodine bearing compounds includes particulates, elemental and organic. In the most efficient design of CFVS, wet scrubbing mechanism has been employed through use of venture scrubber. The Iodine removal process in wet scrubber involves two processes: chemical reaction in highly alkaline aqueous solution and impingement of particulates with water droplets produced in the venturi nozzle. In this paper, venturi has been modeled using the Calvert model. The variation of efficiency has been estimated for the different particle sizes. The impact of the shape parameter of log-normal distribution on the amount of scrubbed iodine has also been assessed. Release phase wise the scrubbed amount of iodine in the venturi based CFVS system has been estimated for a typical BWR. (author)

  5. Research and development of a high-temperature helium-leak detection system (joint research). Part 1 survey on leakage events and current leak detection technology

    Energy Technology Data Exchange (ETDEWEB)

    Sakaba, Nariaki; Nakazawa, Toshio; Kawasaki, Kozo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Urakami, Masao; Saisyu, Sadanori [Japan Atomic Power Co., Tokyo (Japan)

    2003-03-01

    In High Temperature Gas-cooled Reactors (HTGR), the detection of leakage of helium at an early stage is very important for the safety and stability of operations. Since helium is a colourless gas, it is generally difficult to identify the location and the amount of leakage when very little leakage has occurred. The purpose of this R and D is to develop a helium leak detection system for the high temperature environment appropriate to the HTGR. As the first step in the development, this paper describes the result of surveying leakage events at nuclear facilities inside and outside Japan and current gas leakage detection technology to adapt optical-fibre detection technology to HTGRs. (author)

  6. Roxby Downs water leakage

    International Nuclear Information System (INIS)

    1996-01-01

    The Environment, Resource and development Committee has been asked by Parliament to examine 'a massive leakage of water at Roxby Downs' and to make recommendations 'as to further action'. It has also been specifically asked to comment on 'the desirability of the Department of Mines and Energy having prime responsibility for environmental matters in relation to mining operations'. This report begins with a description of the Olympic Dam operations near Roxby Downs and with a brief overview of the regulations controlling those operations. The site of the leakage the Olympic Dam tailings retention system is then described in greater detail. Part 3 describes how the system was originally designed, modified and approved. It ends with a series of findings about the adequacy of the original design (including the monitoring systems built into it) and of the approvals process. Recommendations are then made about how future approvals should be handled. Part 4 of the report outlines how the tailings retention system was built and operated and how the massive leakage from it was detected and reported. Findings about the adequacy of the management of the system and about the initial reactions to the leakage are then made, together with recommendations designed to improve future management of the system. 25 refs., 15 figs

  7. Transcytosis Involvement in Transport System and Endothelial Permeability of Vascular Leakage during Dengue Virus Infection

    Directory of Open Access Journals (Sweden)

    Chanettee Chanthick

    2018-02-01

    Full Text Available The major role of endothelial cells is to maintain homeostasis of vascular permeability and to preserve the integrity of vascular vessels to prevent fluid leakage. Properly functioning endothelial cells promote physiological balance and stability for blood circulation and fluid components. A monolayer of endothelial cells has the ability to regulate paracellular and transcellular pathways for transport proteins, solutes, and fluid. In addition to the paracellular pathway, the transcellular pathway is another route of endothelial permeability that mediates vascular permeability under physiologic conditions. The transcellular pathway was found to be associated with an assortment of disease pathogeneses. The clinical manifestation of severe dengue infection in humans is vascular leakage and hemorrhagic diatheses. This review explores and describes the transcellular pathway, which is an alternate route of vascular permeability during dengue infection that corresponds with the pathologic finding of intact tight junction. This pathway may be the route of albumin transport that causes endothelial dysfunction during dengue virus infection.

  8. Conceptual Design of Portable Filtered Air Suction Systems For Prevention of Released Radioactive Gas under Severe Accidents of NPP

    Energy Technology Data Exchange (ETDEWEB)

    Gu, Beom W.; Choi, Su Y.; Yim, Man S.; Rim, Chun T. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    It becomes evident that severe accidents may occur by unexpected disasters such as tsunami, heavy flood, or terror. Once radioactive material is released from NPP through severe accidents, there are no ways to prevent the released radioactive gas spreading in the air. As a remedy for this problem, the idea on the portable filtered air suction system (PoFASS) for the prevention of released radioactive gas under severe accidents was proposed. In this paper, the conceptual design of a PoFASS focusing on the number of robot fingers and robot arm rods are proposed. In order to design a flexible robot suction nozzle, mathematical models for the gaps which represent the lifted heights of extensible covers for given convex shapes of pipes and for the covered areas are developed. In addition, the system requirements for the design of the robot arms of PoFASS are proposed, which determine the accessible range of leakage points of released radioactive gas. In this paper, the conceptual designs of the flexible robot suction nozzle and robot arm have been conducted. As a result, the minimum number of robot fingers and robot arm rods are defined to be four and three, respectively. For further works, extensible cover designs on the flexible robot suction nozzle and the application of the PoFASS to the inside of NPP should be studied because the radioactive gas may be released from connection pipes between the containment building and auxiliary buildings.

  9. An application of probabilistic safety assessment methods to model aircraft systems and accidents

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Guridi, G.; Hall, R.E.; Fullwood, R.R.

    1998-08-01

    A case study modeling the thrust reverser system (TRS) in the context of the fatal accident of a Boeing 767 is presented to illustrate the application of Probabilistic Safety Assessment methods. A simplified risk model consisting of an event tree with supporting fault trees was developed to represent the progression of the accident, taking into account the interaction between the TRS and the operating crew during the accident, and the findings of the accident investigation. A feasible sequence of events leading to the fatal accident was identified. Several insights about the TRS and the accident were obtained by applying PSA methods. Changes proposed for the TRS also are discussed.

  10. An Accident Precursor Analysis Process Tailored for NASA Space Systems

    Science.gov (United States)

    Groen, Frank; Stamatelatos, Michael; Dezfuli, Homayoon; Maggio, Gaspare

    2010-01-01

    Accident Precursor Analysis (APA) serves as the bridge between existing risk modeling activities, which are often based on historical or generic failure statistics, and system anomalies, which provide crucial information about the failure mechanisms that are actually operative in the system and which may differ in frequency or type from those in the various models. These discrepancies between the models (perceived risk) and the system (actual risk) provide the leading indication of an underappreciated risk. This paper presents an APA process developed specifically for NASA Earth-to-Orbit space systems. The purpose of the process is to identify and characterize potential sources of system risk as evidenced by anomalous events which, although not necessarily presenting an immediate safety impact, may indicate that an unknown or insufficiently understood risk-significant condition exists in the system. Such anomalous events are considered accident precursors because they signal the potential for severe consequences that may occur in the future, due to causes that are discernible from their occurrence today. Their early identification allows them to be integrated into the overall system risk model used to intbrm decisions relating to safety.

  11. Investigation of air cleaning system response to accident conditions

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  12. Development of Information Display System for Operator Support in Severe Accident

    International Nuclear Information System (INIS)

    Jeong, Kwang Il; Lee, Joon Ku

    2016-01-01

    When the severe accident occurs, the technical support center (TSC) performs the mitigation strategy with severe accident management guidelines (SAMG) and communicates with main control room (MCR) operators to obtain information of plant's status. In such circumstances, the importance of an information display for severe accident is increased. Therefore an information display system dedicated to severe accident conditions is required to secure the plant information, to provide the necessary information to MCR operators and TSC operators, and to support the decision using these information. We setup the design concept of severe accident information display system (SIDS) in the previous study and defined its requirements of function and performance. This paper describes the process, results of the identification of the severe accident information for MCR operator and the implementation of SIDS. Further implementation on post-accident monitoring function and data validation function for severe accidents will be accomplished in the future

  13. Development of Information Display System for Operator Support in Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwang Il; Lee, Joon Ku [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    When the severe accident occurs, the technical support center (TSC) performs the mitigation strategy with severe accident management guidelines (SAMG) and communicates with main control room (MCR) operators to obtain information of plant's status. In such circumstances, the importance of an information display for severe accident is increased. Therefore an information display system dedicated to severe accident conditions is required to secure the plant information, to provide the necessary information to MCR operators and TSC operators, and to support the decision using these information. We setup the design concept of severe accident information display system (SIDS) in the previous study and defined its requirements of function and performance. This paper describes the process, results of the identification of the severe accident information for MCR operator and the implementation of SIDS. Further implementation on post-accident monitoring function and data validation function for severe accidents will be accomplished in the future.

  14. Development of system of computer codes for severe accident analysis and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Jang, H S; Jeon, M H; Cho, N J. and others [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1992-01-15

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in nuclear power plants. This system of codes is necessary to conduct Individual Plant Examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident-resistance. Severe accident can be mitigated by the proper accident management strategies. Some operator action for mitigation can lead to more disastrous result and thus uncertain severe accident phenomena must be well recognized. There must be further research for development of severe accident management strategies utilizing existing plant resources as well as new design concepts.

  15. Development of system of computer codes for severe accident analysis and its applications

    International Nuclear Information System (INIS)

    Jang, H. S.; Jeon, M. H.; Cho, N. J. and others

    1992-01-01

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in nuclear power plants. This system of codes is necessary to conduct Individual Plant Examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident-resistance. Severe accident can be mitigated by the proper accident management strategies. Some operator action for mitigation can lead to more disastrous result and thus uncertain severe accident phenomena must be well recognized. There must be further research for development of severe accident management strategies utilizing existing plant resources as well as new design concepts

  16. Development of instrumentation systems for severe accidents. 4. New accident tolerant in-containment pressure transducer for containment pressure monitoring system

    International Nuclear Information System (INIS)

    Oba, Masato; Teruya, Kuniyuki; Yoshitsugu, Makoto; Ikeuchi, Takeshi

    2015-01-01

    The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident and are developing the Severe Accident Instrumentations and Monitoring Systems that could measure the parameters in severe accident conditions. Mitsubishi Heavy Industries, LTD (MHI) developed a new accident tolerant containment pressure monitoring system and demonstrated that the monitoring system could endure extremely harsh environmental conditions that envelop severe accident environmental conditions inside a containment such as maximum operating temperature of up to 300degC and total integrated dose (TID) of 1 MGy gamma. The new containment pressure monitoring system comprises of a strain gage type pressure transducer and a mineral insulated (MI) cable with ceramic connectors, which are located in the containment, and a strain measuring amplifier located outside the containment. Less thermal and radiation degradation is achieved because of minimizing use of organic materials for in-containment equipment such as the transducer and connectors. Several tests were performed to demonstrate the performance and capability of the in-containment equipment under severe accident environmental conditions and the major steps in this testing were run in the following test sequences: (1) the baseline functional tests (e.g., repeatability, non-linearity, hysteresis, and so on) under normal conditions, (2) accident radiation testing, (3) seismic testing, and (4) steam/temperature test exposed to simulated severe accident environmental conditions. The test results demonstrate that the new pressure transducer can endure the simulated severe accident conditions. (author)

  17. Cirrus Airframe Parachute System and Odds of a Fatal Accident in Cirrus Aircraft Crashes.

    Science.gov (United States)

    Alaziz, Mustafa; Stolfi, Adrienne; Olson, Dean M

    2017-06-01

    General aviation (GA) accidents have continued to demonstrate high fatality rates. Recently, ballistic parachute recovery systems (BPRS) have been introduced as a safety feature in some GA aircraft. This study evaluates the effectiveness and associated factors of the Cirrus Airframe Parachute System (CAPS) at reducing the odds of a fatal accident in Cirrus aircraft crashes. Publicly available Cirrus aircraft crash reports were obtained from the National Transportation Safety Board (NTSB) database for the period of January 1, 2001-December 31, 2016. Accident metrics were evaluated through univariate and multivariate analyses regarding odds of a fatal accident and use of the parachute system. Included in the study were 268 accidents. For CAPS nondeployed accidents, 82 of 211 (38.9%) were fatal as compared to 8 of 57 (14.0%) for CAPS deployed accidents. After controlling for all other factors, the adjusted odds ratio for a fatal accident when CAPS was not deployed was 13.1. The substantial increased odds of a fatal accident when CAPS was not deployed demonstrated the effectiveness of CAPS at providing protection of occupants during an accident. Injuries were shifted from fatal to serious or minor with the use of CAPS and postcrash fires were significantly reduced. These results suggest that BPRS could play a significant role in the next major advance in improving GA accident survival.Alaziz M, Stolfi A, Olson DM. Cirrus Airframe Parachute System and odds of a fatal accident in Cirrus aircraft crashes. Aerosp Med Hum Perform. 2017; 88(6):556-564.

  18. Future Integrated Systems Concept for Preventing Aircraft Loss-of-Control Accidents

    Science.gov (United States)

    Belcastro, Christine M.; Jacobson, Steven r.

    2010-01-01

    Loss of control remains one of the largest contributors to aircraft fatal accidents worldwide. Aircraft loss-of-control accidents are highly complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. Hence, there is no single intervention strategy to prevent these accidents. This paper presents future system concepts and research directions for preventing aircraft loss-of-control accidents.

  19. A Modular Telerobot Control System for Accident Response

    International Nuclear Information System (INIS)

    Anderson, Robert J.; Shirey, David L.

    1999-01-01

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

  20. Centrifugal Filtration System for Severe Accident Source Term Treatment

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Shu Chang; Yim, Man Sung [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of this paper is to present the conceptual design of a filtration system that can be used to process airborne severe accident source term. Reactor containment may lose its structural integrity due to over-pressurization during a severe accident. This can lead to uncontrolled radioactive releases to the environment. For preventing the dispersion of these uncontrolled radioactive releases to the environment, several ways to capture or mitigate these radioactive source term releases are under investigation at KAIST. Such technologies are based on concepts like a vortex-like air curtain, a chemical spray, and a suction arm. Treatment of the radioactive material captured by these systems would be required, before releasing to environment. For current filtration systems in the nuclear industry, IAEA lists sand, multi-venturi scrubber, high efficiency particulate arresting (HEPA), charcoal and combinations of the above in NS-G-1-10, 4.143. Most if not all of the requirements of the scenario for applying this technology near the containment of an NPP site and the environmental constraints were analyzed for use in the design of the centrifuge filtration system.

  1. The development of a nuclear accident risk information system

    International Nuclear Information System (INIS)

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

    2001-01-01

    The computerized system NARIS (Nuclear Accident Risk Information System) was developed in order to support the estimation of health effects and the establishment the effective risk reduction strategies. Using the system, we can analyze the distribution of health effects easily by displaying the results on the digital map of the site. Also, the thematic mapping allows the diverse analyses of the distribution of the health effects. The NARIS can be used in the emergency operation facilities in order to analyze the distribution of the health effects resulting from the severe accidents of a nuclear power plant. Also, the rapid analysis of the health effect is possible by storing the health effect results in the form of a database. Therefore, the staffs of the emergency operation facilities can establish the rapid and effective emergency response strategies. The module for the optimization of the costs and benefits and the decision making support will be added. The technical support for the establishment of the optimum and effective emergency response strategies will be possible using this system

  2. Loose parts, vibration and leakage monitoring methods and systems to increase availability, transparency and lifetime of power plants

    International Nuclear Information System (INIS)

    Streicher, V.; Jax, P.; Ruthrof, K.

    1987-01-01

    This paper deals with three stand-alone-systems as an aid to check the mechanical integrity of the primary circuit of nuclear power plants. The main goals of these systems are early detection of faults and malfunctions, the facilitation of fault clearance, the avoidance of sequential damage and reduction of inspection time and cost. Obviously the proper application of the systems as well as the measures they induce and make possible increase the availability of the plant and contribute to lifetime extension. In order to detect, identify and pinpoint the changes in component structure such as loosened connections, broken parts or components, loose or loosened particles, fatigued materials, cracks and leaks, specialized monitoring systems were developed by KWU (Kraftwerk Union AG) during the last ten years. Requirements concerning vibration, loose parts and leakage monitoring are part of German guidelines and safety standards. Therefore systems for these applications are implemented in most of the nuclear power plants in Western Germany. This paper presents newly developed, microprocessor-based systems for loose parts monitoring, vibration monitoring and leakage monitoring and also includes specific case histories for the different topics

  3. On Probability Leakage

    OpenAIRE

    Briggs, William M.

    2012-01-01

    The probability leakage of model M with respect to evidence E is defined. Probability leakage is a kind of model error. It occurs when M implies that events $y$, which are impossible given E, have positive probability. Leakage does not imply model falsification. Models with probability leakage cannot be calibrated empirically. Regression models, which are ubiquitous in statistical practice, often evince probability leakage.

  4. Development of supporting system for emergency response to maritime transport accidents involving radioactive material

    International Nuclear Information System (INIS)

    Odano, N.; Matsuoka, T.; Suzuki, H.

    2004-01-01

    National Maritime Research Institute has developed a supporting system for emergency response of competent authority to maritime transport accidents involving radioactive material. The supporting system for emergency response has functions of radiation shielding calculation, marine diffusion simulation, air diffusion simulation and radiological impact evaluation to grasp potential hazard of radiation. Loss of shielding performance accident and loss of sealing ability accident were postulated and impact of the accidents was evaluated based on the postulated accident scenario. Procedures for responding to emergency were examined by the present simulation results

  5. Developement of integrated evaluation system for severe accident management

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Ha; Kim, H. D.; Park, S. Y.; Kim, K. R.; Park, S. H.; Choi, Y.; Song, Y. M.; Ahn, K. I.; Park, J. H

    2005-04-01

    The scope of the project includes four activities such as construction of DB, development of data base management tool, development of severe accident analysis code system and FP studies. In the construction of DB, level-1,2 PSA results and plant damage states event trees were mainly used to select the following target initiators based on frequencies: LLOCA, MLOCA, SLOCA, station black out, LOOP, LOFW and SGTR. These scenarios occupy more than 95% of the total frequencies of the core damage sequences at KSNP. In the development of data base management tool, SARD 2.0 was developed under the PC microsoft windows environment using the visual basic 6.0 language. In the development of severe accident analysis code system, MIDAS 1.0 was developed with new features of FORTRAN-90 which makes it possible to allocate the storage dynamically and to use the user-defined data type, leading to an efficient memory treatment and an easy understanding. Also for user's convenience, the input (IEDIT) and output (IPLOT) processors were developed and implemented into the MIDAS code. For the model development of MIDAS concerning the FP behavior, the one dimensional thermophoresis model was developed and it gave much improvement to predict the amount of FP deposited on the SG U-tube. Also the source term analysis methodology was set up and applied to the KSNP and APR1400.

  6. Development of ultrasonic high temperature system for severe accidents research

    International Nuclear Information System (INIS)

    Koo, Kil Mo; Kang, Kyung Ho; Kim, Young Ro and others

    2000-07-01

    The aims of this study are to find a gap formation between corium melt and the reactor lower head vessel, to verify the principle of the gap formation and to analyze the effect of the gap formation on the thermal behavior of corium melt and the lower plenum. This report aims at suggesting development of a new high temperature measuring system using an ultrasonic method which overcomes the limitations of the present thermocouple method used for severe accident experiments. Also, this report describes the design and manufacturing method of the ultrasonic system. At that time, the sensor element is fabricated to a reflective element using 1mm diameter and 50 mm and 80 mm long tungsten alloy wires. This temperature measuring system is intended to measure up to 2800 deg C

  7. Development of the severe accident risk information database management system SARD

    International Nuclear Information System (INIS)

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

    The main purpose of this report is to introduce essential features and functions of a severe accident risk information management system, SARD (Severe Accident Risk Database Management System) version 1.0, which has been developed in Korea Atomic Energy Research Institute, and database management and data retrieval procedures through the system. The present database management system has powerful capabilities that can store automatically and manage systematically the plant-specific severe accident analysis results for core damage sequences leading to severe accidents, and search intelligently the related severe accident risk information. For that purpose, the present database system mainly takes into account the plant-specific severe accident sequences obtained from the Level 2 Probabilistic Safety Assessments (PSAs), base case analysis results for various severe accident sequences (such as code responses and summary for key-event timings), and related sensitivity analysis results for key input parameters/models employed in the severe accident codes. Accordingly, the present database system can be effectively applied in supporting the Level 2 PSA of similar plants, for fast prediction and intelligent retrieval of the required severe accident risk information for the specific plant whose information was previously stored in the database system, and development of plant-specific severe accident management strategies

  8. Development of the severe accident risk information database management system SARD

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

    The main purpose of this report is to introduce essential features and functions of a severe accident risk information management system, SARD (Severe Accident Risk Database Management System) version 1.0, which has been developed in Korea Atomic Energy Research Institute, and database management and data retrieval procedures through the system. The present database management system has powerful capabilities that can store automatically and manage systematically the plant-specific severe accident analysis results for core damage sequences leading to severe accidents, and search intelligently the related severe accident risk information. For that purpose, the present database system mainly takes into account the plant-specific severe accident sequences obtained from the Level 2 Probabilistic Safety Assessments (PSAs), base case analysis results for various severe accident sequences (such as code responses and summary for key-event timings), and related sensitivity analysis results for key input parameters/models employed in the severe accident codes. Accordingly, the present database system can be effectively applied in supporting the Level 2 PSA of similar plants, for fast prediction and intelligent retrieval of the required severe accident risk information for the specific plant whose information was previously stored in the database system, and development of plant-specific severe accident management strategies.

  9. [a Monitoring System For Work-related Accidents In Piracicaba, São Paulo, Brazil].

    OpenAIRE

    Cordeiro, Ricardo; Vilela, Rodolfo Andrade Gouveia; de Medeiros, Maria Angélica Tavares; Gonçalves, Cláudia Giglio de Oliveira; Bragantini, Clarice Aparecida; Varolla, Antenor J; Celso, Stephan

    2015-01-01

    The authors report on the development of a work accident monitoring system in Piracicaba, São Paulo State, Brazil, with the following characteristics: information feeding the system is obtained in real time directly from accident treatment centers; the system has universal monitoring, covering all work-related accidents in Piracicaba, regardless of the nature of the worker's employment conditions, place of work, or place of residence; health surveillance and promotion of health initiatives ar...

  10. Evaluation of decision support systems for nuclear accidents

    International Nuclear Information System (INIS)

    Sdouz, G.; Mueck, K.

    1998-05-01

    In order to adopt countermeasures to protect the public after an accident in a nuclear power plant in an appropriate and optimum way, decision support systems offer a valuable assistance in supporting the decision maker in choosing and optimizing protective actions. Such decision support systems may range from simple systems to accumulate relevant parameters for the evaluation of the situation over prediction models for the rapid evaluation of the dose to be expected to systems which permit the evaluation and comparison of possible countermeasures. Since the establishment of a decision support systems obviously is also required in Austria, an evaluation of systems available or in the state of development in other countries or unions was performed. The aim was to determine the availability of decision support systems in various countries and to evaluate them with regard to depth and extent of the system. The evaluation showed that in most industrialized countries the requirement for a decision support system was realized, but in only few countries actual systems are readily available and operable. Most systems are limited to early phase consequences, i.e. dispersion calculations of calculated source terms and the estimation of exposure in the vicinity of the plant. Only few systems offer the possibility to predict long-term exposures by ingestion. Few systems permit also an evaluation of potential countermeasures, in most cases, however, limited to a few short-term countermeasures. Only one system which is presently not operable allows the evaluation of a large number of agricultural countermeasures. In this report the different systems are compared. The requirements with regard to an Austrian decision support system are defined and consequences for a possible utilization of a DSS or parts thereof for the Austrian decision support system are derived. (author)

  11. Supporting system in emergency response plan for nuclear material transport accidents

    International Nuclear Information System (INIS)

    Nakagome, Y.; Aoki, S.

    1993-01-01

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

  12. Replacement of the drain system of secondary circuit at Monju

    International Nuclear Information System (INIS)

    Itoh, Kenji; Onuki, Koji; Tomobe, Katsuma; Taniyama, Sadami

    2003-01-01

    Monju is as a Japan's prototype fast breeder reactor cooled by liquid sodium. In the course of power buildup tests, the sodium leakage accident broke out on 8th December 1995. Though Monju has been already equipped with countermeasure systems against the sodium leakage accident, some additional improvements will be taken in order to reduce the damage by the leaked sodium when another leakage accident should recur. The most characteristic work is the drain system modification that leads to shorten the drain time and to reduce the quantity of leaked sodium in the event of sodium leakage. (author)

  13. Overview of main accident parameters in car-to-cyclist accidents for use in AEB-system test protocol

    NARCIS (Netherlands)

    Uittenbogaard, J.; Camp, O.M.G.C. op den; Montfort, S. van

    2016-01-01

    The number of fatalities in road traffic accidents in Europe is decreasing. Unfortunately, the number of fatalities among cyclists does not follow this trend with the same rate [1]. The au-tomotive industry is making a significant effort in the development and implementation of safety systems in

  14. A database system for the management of severe accident risk information, SARD

    International Nuclear Information System (INIS)

    Ahn, K. I.; Kim, D. H.

    2003-01-01

    The purpose of this paper is to introduce main features and functions of a PC Windows-based database management system, SARD, which has been developed at Korea Atomic Energy Research Institute for automatic management and search of the severe accident risk information. Main functions of the present database system are implemented by three closely related, but distinctive modules: (1) fixing of an initial environment for data storage and retrieval, (2) automatic loading and management of accident information, and (3) automatic search and retrieval of accident information. For this, the present database system manipulates various form of the plant-specific severe accident risk information, such as dominant severe accident sequences identified from the plant-specific Level 2 Probabilistic Safety Assessment (PSA) and accident sequence-specific information obtained from the representative severe accident codes (e.g., base case and sensitivity analysis results, and summary for key plant responses). The present database system makes it possible to implement fast prediction and intelligent retrieval of the required severe accident risk information for various accident sequences, and in turn it can be used for the support of the Level 2 PSA of similar plants and for the development of plant-specific severe accident management strategies

  15. A database system for the management of severe accident risk information, SARD

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, K. I.; Kim, D. H. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    The purpose of this paper is to introduce main features and functions of a PC Windows-based database management system, SARD, which has been developed at Korea Atomic Energy Research Institute for automatic management and search of the severe accident risk information. Main functions of the present database system are implemented by three closely related, but distinctive modules: (1) fixing of an initial environment for data storage and retrieval, (2) automatic loading and management of accident information, and (3) automatic search and retrieval of accident information. For this, the present database system manipulates various form of the plant-specific severe accident risk information, such as dominant severe accident sequences identified from the plant-specific Level 2 Probabilistic Safety Assessment (PSA) and accident sequence-specific information obtained from the representative severe accident codes (e.g., base case and sensitivity analysis results, and summary for key plant responses). The present database system makes it possible to implement fast prediction and intelligent retrieval of the required severe accident risk information for various accident sequences, and in turn it can be used for the support of the Level 2 PSA of similar plants and for the development of plant-specific severe accident management strategies.

  16. Reliability analysis of emergency decay heat removal system of nuclear ship under various accident conditions

    International Nuclear Information System (INIS)

    Matsuoka, Takeshi

    1984-01-01

    A reliability analysis is given for the emergency decay heat removal system of the Nuclear Ship ''Mutsu'' and the emergency sea water cooling system of the Nuclear Ship ''Savannah'', under ten typical nuclear ship accident conditions. Basic event probabilities under these accident conditions are estimated from literature survey. These systems of Mutsu and Savannah have almost the same reliability under the normal condition. The dispersive arrangement of a system is useful to prevent the reduction of the system reliability under the condition of an accident restricted in one room. As for the reliability of these two systems under various accident conditions, it is seen that the configuration and the environmental condition of a system are two main factors which determine the reliability of the system. Furthermore, it was found that, for the evaluation of the effectiveness of safety system of a nuclear ship, it is necessary to evaluate its reliability under various accident conditions. (author)

  17. Feasibility study on the rod ejection accident analysis with RETRAN-MASTER code system

    International Nuclear Information System (INIS)

    Kim, Y. H.; Lee, C. S.

    2003-01-01

    KEPRI has been developed the in-house methodology for non-LOCA safety analyses based on the codes and methodologies of vendors and EPRI. Using the methodology, the rod ejection accident, which is classified into the generic accident analysis category of reactivity insertion accident in primary system, has been analyzed with RETRAN-MASTER code system. And the feasibility of the coupled code system has been verified by the review of the results. Furthermore, to assess the important parameters to the accident, the sensitivity analyses have been carried out over some parameters

  18. A proposed scalable design and simulation of wireless sensor network-based long-distance water pipeline leakage monitoring system.

    Science.gov (United States)

    Almazyad, Abdulaziz S; Seddiq, Yasser M; Alotaibi, Ahmed M; Al-Nasheri, Ahmed Y; BenSaleh, Mohammed S; Obeid, Abdulfattah M; Qasim, Syed Manzoor

    2014-02-20

    Anomalies such as leakage and bursts in water pipelines have severe consequences for the environment and the economy. To ensure the reliability of water pipelines, they must be monitored effectively. Wireless Sensor Networks (WSNs) have emerged as an effective technology for monitoring critical infrastructure such as water, oil and gas pipelines. In this paper, we present a scalable design and simulation of a water pipeline leakage monitoring system using Radio Frequency IDentification (RFID) and WSN technology. The proposed design targets long-distance aboveground water pipelines that have special considerations for maintenance, energy consumption and cost. The design is based on deploying a group of mobile wireless sensor nodes inside the pipeline and allowing them to work cooperatively according to a prescheduled order. Under this mechanism, only one node is active at a time, while the other nodes are sleeping. The node whose turn is next wakes up according to one of three wakeup techniques: location-based, time-based and interrupt-driven. In this paper, mathematical models are derived for each technique to estimate the corresponding energy consumption and memory size requirements. The proposed equations are analyzed and the results are validated using simulation.

  19. Effect of diesel leakage in circulating cooling water system on preponderant bacteria diversity and bactericidal effect of biocides.

    Science.gov (United States)

    Zhong, Huiyun; Liu, Fang; Lu, Jinjin; Yang, Wei; Zhao, Chaocheng

    2015-01-01

    Petroleum products leakage results in adverse effect on the normal operation of a circulating cooling water system. However, relatively little research has been done to explore the effect of petroleum products leakage on circulating cooling water quality and biofilm preponderant bacteria diversity. Also, normal biocides application modes cannot fulfil the need for biofilm control. In this study, diesel oil was used as the experimental subject representing leaking petroleum products; the effect of diesel addition on biofilm preponderant bacteria diversity and the bactericidal effect of chlorine dioxide and tetradecyl dimethyl benzyl ammonium chloride (1427) was investigated. Bacterial community structures were examined by PCR-denaturing gradient gel electrophoresis and PCR cloning of 16S rDNA genes. Except for 100 mg/L diesel, increasing diesel concentration enhanced the biofilm detachment ratio compared with the control test. The microstructure of biofilm samples with 0, 300 and 900 mg/L diesel addition was observed. The species of preponderant bacteria in the biofilm sample with 300 mg/L diesel addition were more and the bacterial distribution was more uniform than those in the biofilm sample with 900 mg/L diesel addition. With ClO2 and 1427 addition, chemical oxygen demand increased, lipid phosphorus and bacterial count first decreased and then remained stable, and the bactericidal ratio first increased and then remained stable. Diesel addition variation has more obvious effect on ClO2 than 1427.

  20. Suppression of fast electron leakage from large openings in a plasma neutralizer for N-NB systems

    International Nuclear Information System (INIS)

    Kashiwagi, Mieko; Hanada, Masaya; Yamana, Takashi; Inoue, Takashi; Imai, Tsuyoshi; Taniguchi, Masaki; Watanabe, Kazuhiro

    2006-01-01

    To produce highly ionized plasmas at low operating pressure in a plasma neutralizer of negative ion based neutral beam (N-NB) systems, it is a critical issue to suppress leakage of fast electrons through large openings as the beam entrance/exit. The authors propose to form weak transverse magnetic fields without a significant beam deflection, called the shield field, across the large openings of the neutralizer. A numerical study showed that the shield field of only few tens of Gauss is sufficient to suppress the fast electron leakage from the openings. By measuring of an electron energy distribution function (EEDF), it was confirmed that such a weak magnetic field is enough to repel the fast electrons back into the neutralizer plasma. As the result, the plasma density increased with the shield field strength and saturated at 30 G. The plasma density reached 50% higher value than that without the shield field. Thus it was found that reflected fast electrons by the shield field of only 30 G work effectively for the plasma generation. It was also estimated that such a weak magnetic field sufficiently suppresses the deflection of a 1 MeV beam. This weak magnetic field would be applicable to the plasma neutralizer for the fusion demonstration (DEMO) plant

  1. A Proposed Scalable Design and Simulation of Wireless Sensor Network-Based Long-Distance Water Pipeline Leakage Monitoring System

    Directory of Open Access Journals (Sweden)

    Abdulaziz S. Almazyad

    2014-02-01

    Full Text Available Anomalies such as leakage and bursts in water pipelines have severe consequences for the environment and the economy. To ensure the reliability of water pipelines, they must be monitored effectively. Wireless Sensor Networks (WSNs have emerged as an effective technology for monitoring critical infrastructure such as water, oil and gas pipelines. In this paper, we present a scalable design and simulation of a water pipeline leakage monitoring system using Radio Frequency IDentification (RFID and WSN technology. The proposed design targets long-distance aboveground water pipelines that have special considerations for maintenance, energy consumption and cost. The design is based on deploying a group of mobile wireless sensor nodes inside the pipeline and allowing them to work cooperatively according to a prescheduled order. Under this mechanism, only one node is active at a time, while the other nodes are sleeping. The node whose turn is next wakes up according to one of three wakeup techniques: location-based, time-based and interrupt-driven. In this paper, mathematical models are derived for each technique to estimate the corresponding energy consumption and memory size requirements. The proposed equations are analyzed and the results are validated using simulation.

  2. Development of monitoring system using acoustic emission for detection of helium gas leakage for primary cooling system and flow-induced vibration for heat transfer tube of heat exchangers for the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Kunitomi, Kazuhiko; Furusawa, Takayuki; Shinozaki, Masayuki; Satoh, Yoshiyuki; Yanagibashi, Minoru

    1998-10-01

    The High Temperature Engineering Test Reactor (HTTR) uses helium gas for its primary coolant, whose leakage inside reactor containment vessel is considered in design of the HTTR. It is necessary to detect leakage of helium gas at an early stage so that total amount of the leakage should be as small as possible. On the other hand, heat transfer tubes of heat exchangers of the HTTR are designed not to vibrate at normal operation, but the flow-induced vibration is to be monitored to provide against an emergency. Thus monitoring system of acoustic emission for detection of primary coolant leakage and vibration of heat transfer tubes was developed and applied to the HTTR. Before the application to the HTTR, leakage detection test was performed using 1/4 scaled model of outer tube of primary concentric hot gas duct. Result of the test covers detectable minimum leakage rate and effect of difference in gas, pressure, shape of leakage path and distance from the leaking point. Detectable minimum leakage rate was about 5 Ncc/sec. The monitoring system is promising in leakage detection, though countermeasure to noise is to be needed after the HTTR starts operating. (author)

  3. Elements of a national emergency response system for nuclear accidents

    International Nuclear Information System (INIS)

    Dickerson, M.H.

    1987-01-01

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

  4. Severe accident simulation at Olkiuoto

    Energy Technology Data Exchange (ETDEWEB)

    Tirkkonen, H.; Saarenpaeae, T. [Teollisuuden Voima Oy (TVO), Olkiluoto (Finland); Cliff Po, L.C. [Micro-Simulation Technology, Montville, NJ (United States)

    1995-09-01

    A personal computer-based simulator was developed for the Olkiluoto nuclear plant in Finland for training in severe accident management. The generic software PCTRAN was expanded to model the plant-specific features of the ABB Atom designed BWR including its containment over-pressure protection and filtered vent systems. Scenarios including core heat-up, hydrogen generation, core melt and vessel penetration were developed in this work. Radiation leakage paths and dose rate distribution are presented graphically for operator use in diagnosis and mitigation of accidents. Operating on an graphically for operator use in diagnosis and mitigation of accidents. Operating on an 486 DX2-66, PCTRAN-TVO achieves a speed about 15 times faster than real-time. A convenient and user-friendly graphic interface allows full interactive control. In this paper a review of the component models and verification runs are presented.

  5. A severe accident analysis for the system-integrated modular advanced reactor

    International Nuclear Information System (INIS)

    Jung, Gunhyo; Jae, Moosung

    2015-01-01

    The System-Integrated Modular Advanced Reactor (SMART) that has been recently designed in KOREA and has acquired standard design certification from the nuclear power regulatory body (NSSC) is an integral type reactor with 330MW thermal power. It is a small sized reactor in which the core, steam generator, pressurizer, and reactor coolant pump that are in existing pressurized light water reactors are designed to be within a pressure vessel without any separate pipe connection. In addition, this reactor has much different design characteristics from existing pressurized light water reactors such as the adoption of a passive residual heat removal system and a cavity flooding system. Therefore, the safety of the SMART against severe accidents should be checked through severe accident analysis reflecting the design characteristics of the SMART. For severe accident analysis, an analysis model has been developed reflecting the design information presented in the standard design safety analysis report. The severe accident analysis model has been developed using the MELCOR code that is widely used to evaluate pressurized LWR severe accidents. The steady state accident analysis model for the SMART has been simulated. According to the analysis results, the developed model reflecting the design of the SMART is found to be appropriate. Severe accident analysis has been performed for the representative accident scenarios that lead to core damage to check the appropriateness of the severe accident management plan for the SMART. The SMART has been shown to be safe enough to prevent severe accidents by utilizing severe accident management systems such as a containment spray system, a passive hydrogen recombiner, and a cavity flooding system. In addition, the SMART is judged to have been technically improved remarkably compared to existing PWRs. The SMART has been designed to have a larger reactor coolant inventory compared to its core's thermal power, a large surface area in

  6. Modeling of pipe break accident in a district heating system using RELAP5 computer code

    International Nuclear Information System (INIS)

    Kaliatka, A.; Valinčius, M.

    2012-01-01

    Reliability of a district heat supply system is a very important factor. However, accidents are inevitable and they occur due to various reasons, therefore it is necessary to have possibility to evaluate the consequences of possible accidents. This paper demonstrated the capabilities of developed district heating network model (for RELAP5 code) to analyze dynamic processes taking place in the network. A pipe break in a water supply line accident scenario in Kaunas city (Lithuania) heating network is presented in this paper. The results of this case study were used to demonstrate a possibility of the break location identification by pressure decrease propagation in the network. -- Highlights: ► Nuclear reactor accident analysis code RELAP5 was applied for accident analysis in a district heating network. ► Pipe break accident scenario in Kaunas city (Lithuania) district heating network has been analyzed. ► An innovative method of pipe break location identification by pressure-time data is proposed.

  7. Accident Tolerant Reactor Shutdown for NTP Systems, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — In brief, USNC's accident submersion safe drums are control drums where a small amount of fuel is added opposite to the neutron absorber and the drums impinge on the...

  8. Inner volume leakage during integrated leakage rate testing

    International Nuclear Information System (INIS)

    Glover, J.P.

    1987-01-01

    During an integrated leak rate test (ILRT), the containment structure is maintained at test pressure with most penetrations isolated. Since penetrations typically employ dual isolation, the possibility exists for the inner isolation to leak while the outer holds. In this case, the ILRT instrumentation system would indicate containment out-leakage when, in fact, only the inner volume between closures is being pressurized. The problem is compounded because this false leakage is not readily observable outside of containment by standard leak inspection techniques. The inner volume leakage eventually subsides after the affected volumes reach test pressure. Depending on the magnitude of leakage and the size of the volumes, equalization could occur prior to the end of the pretest stabilization period, or significant false leakages may persist throughout the entire test. Two simple analyses were performed to quantify the effects of inside volume leakages. First, a lower bound for the equalization time was found. A second analysis was performed to find an approximate upper bound for the stabilization time. The results of both analyses are shown

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

    International Nuclear Information System (INIS)

    1993-01-01

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

  10. Analysis of Waste Leak and Toxic Chemical Release Accidents from Waste Feed Delivery (WFD) Diluent System

    Energy Technology Data Exchange (ETDEWEB)

    WILLIAMS, J.C.

    2000-09-15

    Radiological and toxicological consequences are calculated for 4 postulated accidents involving the Waste Feed Delivery (WFD) diluent addition systems. Consequences for the onsite and offsite receptor are calculated. This analysis contains technical information used to determine the accident consequences for the River Protection Project (RPP) Final Safety Analysis Report (FSAR).

  11. Analysis of Waste Leak and Toxic Chemical Release Accidents from Waste Feed Delivery (WFD) Diluent System

    International Nuclear Information System (INIS)

    WILLIAMS, J.C.

    2000-01-01

    Radiological and toxicological consequences are calculated for 4 postulated accidents involving the Waste Feed Delivery (WFD) diluent addition systems. Consequences for the onsite and offsite receptor are calculated. This analysis contains technical information used to determine the accident consequences for the River Protection Project (RPP) Final Safety Analysis Report (FSAR)

  12. Leakage in the Juelich research reactor

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    On August 17, 1978, a leakage occurred in the DIDO research reactor. Early in the afternoon of this day, the valves of the coolant loop had been checked with the reactor shut off. When the mechanics wanted to oil a tight valve and drilled a hole in the valve cover for this, heavy water started to leak (leakage in the valve membrane). The mechanics left the shielded valve space at once; directly after having a shower, they underwent a radiation protection examination. It was found that none of the mechanics had been exposed to an excessive dose. When other mechanics in protective suits had closed the leak in the valve, a total of 150 liters had leaked into the sump pump at the valve entrance. They were pumped back into the cooling system. About 5 liters of water were evaporated and, via the stack, escaped into the environment. The activity released was about 40 curie; this is less than the permissible amount of 60 curie per week during normal operation. Neither the KFA personnel nor the inhabitants of Juelich and its surroundings were in danger at any moment. Calculations so far yield a maximum radiation exposure below 1mrem at the point of maximum exposure. The cooling circuit could be entered again only one day after the incident. The present shut-off phase of the reactor is not unduly prolonged by this accident. (orig./HP) [de

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  14. The causing model of accidents and preventing system of small mines

    Energy Technology Data Exchange (ETDEWEB)

    Cao, S.; Zhang, L.; Liu, Y.; Li, Y. [Chongqing University, Chongqing (China)

    2008-06-15

    From an analysis of data on fatal accidents in small coal mines in a southern region of China over a period of three years, the time and type of accidents was discussed by applying statistical methods. It is shown that accidents frequently occur at the end of spring and all through summer. Roof accidents and gas disasters constitute severe accidents and traffic accidents are also important. It was found that most accidents are caused by dangerous behaviour of personnel and the unsafe state of equipment combined with economic interest. The three-factor causing model (TFC model) was proposed. Unsafe behaviour is a direct cause influenced by staff and workers while the unsafe nature of equipment is an indirect cause of accidents influence by natural conditions and the level of technical equipment in the mines. A system of accident prevention in small coal collieries was established with the TFC model. In this, scientific management is an important factor. 13 refs., 4 figs., 1 tab.

  15. Factors correlated with traffic accidents as a basis for evaluating Advanced Driver Assistance Systems.

    Science.gov (United States)

    Staubach, Maria

    2009-09-01

    This study aims to identify factors which influence and cause errors in traffic accidents and to use these as a basis for information to guide the application and design of driver assistance systems. A total of 474 accidents were examined in depth for this study by means of a psychological survey, data from accident reports, and technical reconstruction information. An error analysis was subsequently carried out, taking into account the driver, environment, and vehicle sub-systems. Results showed that all accidents were influenced by errors as a consequence of distraction and reduced activity. For crossroad accidents, there were further errors resulting from sight obstruction, masked stimuli, focus errors, and law infringements. Lane departure crashes were additionally caused by errors as a result of masked stimuli, law infringements, expectation errors as well as objective and action slips, while same direction accidents occurred additionally because of focus errors, expectation errors, and objective and action slips. Most accidents were influenced by multiple factors. There is a safety potential for Advanced Driver Assistance Systems (ADAS), which support the driver in information assimilation and help to avoid distraction and reduced activity. The design of the ADAS is dependent on the specific influencing factors of the accident type.

  16. Analyses of systems availability and operator actions to support the development of severe accident procedures

    International Nuclear Information System (INIS)

    Lutz, R.J. Jr.; Scobel, J.H.

    1989-01-01

    This paper reports on traditional analyses of severe accidents, such as those presented in Probabilistic Risk Assessment (PRA) studies of nuclear power stations, that have generally been performed on the assumption that all means of cooling the reactor core are lost and that no operator actions to mitigate the consequences or progression of the severe accident are performed. The assumption to neglect the availability of safety systems and operator actions which do not prevent core melting can lead to erroneous conclusions regarding the plant severer accident profile. Recent work in severe accident management has identified the need to perform analyses which consider all systems availabilities and operator actions, irrespective of their contribution to the prevention of core melting. These new analyses indicate that the traditional analyses result in overfly pessimistic predictions of the time of core melting and the subsequent potential for recovery of core cooling prior to core melting. Additionally, since the traditional analyses do not model all of the operator actions which are prescribed, the impact of additional severe accident operator actions on the progression and consequences of the accident cannot be reliably identified. Further, the more detailed analysis can change the focus of the importance of various system to the prevention of core damage and the mitigation of severe accident consequences. Finally, the simplicity of the traditional analyses can have a considerable impact on severe accident decision making, particularly in the evaluation of alternate plant design features and the priorities for research studies

  17. Building of communication system for nuclear accident emergency disposal based on IP multimedia subsystem

    Science.gov (United States)

    Wang, Kang; Gao, Guiqing; Qin, Yuanli; He, Xiangyong

    2018-05-01

    The nuclear accident emergency disposal must be supported by an efficient, real-time modularization and standardization communication system. Based on the analysis of communication system for nuclear accident emergency disposal which included many functions such as the internal and external communication, multiply access supporting and command center. Some difficult problems of the communication system were discussed such as variety access device type, complex composition, high mobility, set up quickly, multiply business support, and so on. Taking full advantages of the IP Multimedia Subsystem (IMS), a nuclear accident emergency communication system was build based on the IMS. It was studied and implemented that some key unit and module functions of communication system were included the system framework implementation, satellite access, short-wave access, load/vehicle-mounted communication units. The application tests showed that the system could provide effective communication support for the nuclear accident emergency disposal, which was of great practical value.

  18. Privacy leakage in binary biometric systems : from gaussian to binary data

    NARCIS (Netherlands)

    Ignatenko, T.; Willems, F.M.J.; Campisi, P.

    2013-01-01

    In this chapter we investigate biometric key-binding systems for i.i.d. Gaussian biometric sources. In these systems two terminals observe two correlated biometric sequences. Moreover, a secret key, which is independent of the biometric sequences, is selected at the first terminal. The first

  19. Review of current status for designing severe accident management support system

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwang Sub

    2000-05-01

    The development of operator support system (OSS) is ongoing in many other countries due to the complexity both in design and in operation for nuclear power plant. The computerized operator support system includes monitoring of some critical parameters, early detection of plant transient, monitoring of component status, plant maintenance, and safety parameter display, and the operator support system for these areas are developed and are being used in some plants. Up to now, the most operator support system covers the normal operation, abnormal operation, and emergency operation. Recently, however, the operator support system for severe accident is to be developed in some countries. The study for the phenomena of severe accident is not performed sufficiently, but, based on the result up to now, the operator support system even for severe accident will be developed in this study. To do this, at first, the current status of the operator support system for normal/abnormal/emergency operation is reviewed, and the positive aspects and negative aspects of systems are analyzed by their characteristics. And also, the major items that should be considered in designing the severe accident operator support system are derived from the review. With the survey of domestic and foreign operator support systems, they are reviewed in terms of the safety parameter display system, decision-making support system, and procedure-tracking system. For the severe accident, the severe accident management guideline (SAMG) which is developed by Westinghouse is reviewed; the characteristics, structure, and logical flow of SAMG are studied. In addition, the critical parameters for severe accident, which are the basis for operators decision-making in severe accident management and are supplied to the operators and the technical support center, are reviewed, too.

  20. Review of current status for designing severe accident management support system

    International Nuclear Information System (INIS)

    Jeong, Kwang Sub

    2000-05-01

    The development of operator support system (OSS) is ongoing in many other countries due to the complexity both in design and in operation for nuclear power plant. The computerized operator support system includes monitoring of some critical parameters, early detection of plant transient, monitoring of component status, plant maintenance, and safety parameter display, and the operator support system for these areas are developed and are being used in some plants. Up to now, the most operator support system covers the normal operation, abnormal operation, and emergency operation. Recently, however, the operator support system for severe accident is to be developed in some countries. The study for the phenomena of severe accident is not performed sufficiently, but, based on the result up to now, the operator support system even for severe accident will be developed in this study. To do this, at first, the current status of the operator support system for normal/abnormal/emergency operation is reviewed, and the positive aspects and negative aspects of systems are analyzed by their characteristics. And also, the major items that should be considered in designing the severe accident operator support system are derived from the review. With the survey of domestic and foreign operator support systems, they are reviewed in terms of the safety parameter display system, decision-making support system, and procedure-tracking system. For the severe accident, the severe accident management guideline (SAMG) which is developed by Westinghouse is reviewed; the characteristics, structure, and logical flow of SAMG are studied. In addition, the critical parameters for severe accident, which are the basis for operators decision-making in severe accident management and are supplied to the operators and the technical support center, are reviewed, too

  1. [Hospital information system performance for road traffic accidents analysis in a hospital recruitment based area].

    Science.gov (United States)

    Jannot, A-S; Fauconnier, J

    2013-06-01

    Road traffic accidents in France are mainly analyzed through reports completed by the security forces (police and gendarmerie). But the hospital information systems can also identify road traffic accidents via specific documentary codes of the International Classification of Diseases (ICD-10). The aim of this study was therefore to determine whether hospital stays consecutive to road traffic accident were truly identified by these documentary codes in a facility that collects data routinely and to study the consistency of results from hospital information systems and from security forces during the 2002-2008 period. We retrieved all patients for whom a documentary code for road traffic accident was entered in 2002-2008. We manually checked the concordance of documentary code for road traffic accident and trauma origin in 350 patient files. The number of accidents in the Grenoble area was then inferred by combining with hospitalization regional data and compared to the number of persons injured by traffic accidents declared by the security force. These hospital information systems successfully report road traffic accidents with 96% sensitivity (95%CI: [92%, 100%]) and 97% specificity (95%CI: [95%, 99%]). The decrease in road traffic accidents observed was significantly less than that observed was significantly lower than that observed in the data from the security force (45% for security force data against 27% for hospital data). Overall, this study shows that hospital information systems are a powerful tool for studying road traffic accidents morbidity in hospital and are complementary to security force data. Copyright © 2013 Elsevier Masson SAS. All rights reserved.

  2. 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.

  3. 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

    International Nuclear Information System (INIS)

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

    2017-01-01

    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.

  4. Analysis of Fukushima unit 2 accident considering the operating conditions of RCIC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il, E-mail: sikim@kaeri.re.kr; Park, Jong Hwa; Ha, Kwang Soon; Cho, Song-Won; Song, JinHo

    2016-03-15

    Highlights: • Fukushima unit 2 accident was analyzed using MELCOR 1.8.6. • RCIC operating conditions were assumed and best case was selected. • Effect of RCIC operating condition on accident scenario was found. - Abstract: A severe accident in Fukushima occurred on March 11, 2011 and units 1, 2 and 3 were damaged severely. A tsunami following an earthquake made the supply of electricity power stop, and the safety systems, which use AC or DC power in plants could not operate properly. It is supposed that the degree of core degradation of unit 2 is less serious than in the other plants, and it was estimated that the operation of reactor core isolation cooling (RCIC) system at the initial stage of the accident minimized the core damage through decay heat removal. Although the operating conditions of the RCIC system are not known clearly, it can be important to analyze the accident scenario of unit 2. In this study, best case of the Fukushima unit 2 accident was presented considering the operating conditions of the RCIC system. The effects of operating condition on core degradation and fission product release rate to environment were also examined. In addition, importance of torus room flooding level in the accident analysis was discussed. MELCOR 1.8.6 was used in this research, and the geometries of plant and operating conditions of safety system were obtained from TEPCO through OECD/NEA BSAF Project.

  5. Accident Management System Based on Vehicular Network for an Intelligent Transportation System in Urban Environments

    Directory of Open Access Journals (Sweden)

    Yusor Rafid Bahar Al-Mayouf

    2018-01-01

    Full Text Available As cities across the world grow and the mobility of populations increases, there has also been a corresponding increase in the number of vehicles on roads. The result of this has been a proliferation of challenges for authorities with regard to road traffic management. A consequence of this has been congestion of traffic, more accidents, and pollution. Accidents are a still major cause of death, despite the development of sophisticated systems for traffic management and other technologies linked with vehicles. Hence, it is necessary that a common system for accident management is developed. For instance, traffic congestion in most urban areas can be alleviated by the real-time planning of routes. However, the designing of an efficient route planning algorithm to attain a globally optimal vehicle control is still a challenge that needs to be solved, especially when the unique preferences of drivers are considered. The aim of this paper is to establish an accident management system that makes use of vehicular ad hoc networks coupled with systems that employ cellular technology in public transport. This system ensures the possibility of real-time communication among vehicles, ambulances, hospitals, roadside units, and central servers. In addition, the accident management system is able to lessen the amount of time required to alert an ambulance that it is required at an accident scene by using a multihop optimal forwarding algorithm. Moreover, an optimal route planning algorithm (ORPA is proposed in this system to improve the aggregate spatial use of a road network, at the same time bringing down the travel cost of operating a vehicle. This can reduce the incidence of vehicles being stuck on congested roads. Simulations are performed to evaluate ORPA, and the results are compared with existing algorithms. The evaluation results provided evidence that ORPA outperformed others in terms of average ambulance speed and travelling time. Finally, our

  6. Development of high impedance measurement system for water leakage detection in implantable neuroprosthetic devices.

    Science.gov (United States)

    Yousif, Aziz; Kelly, Shawn K

    2016-08-01

    There has been a push for a greater number of channels in implantable neuroprosthetic devices; but, that number has largely been limited by current hermetic packaging technology. Microfabricated packaging is becoming reality, but a standard testing system is needed to prepare these devices for clinical trials. Impedance measurements of electrodes built into the packaging layers may give an early warning of device failure and predict device lifetime. Because the impedance magnitudes of such devices can be on the order of gigaohms, a versatile system was designed to accommodate ultra-high impedances and allow future integrated circuit implementation in current neural prosthetic technologies. Here we present the circuitry, control software, and preliminary testing results of our designed system.

  7. 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.

  8. Sisifo-gas a computerised system to support severe accident training and management

    International Nuclear Information System (INIS)

    Castro, A.; Buedo, J.L.; Borondo, L.; Lopez, N.

    2001-01-01

    Nuclear Power Plants (NPP) will have to be prepared to face the management of severe accidents, through the development of Severe Accident Guides and sophisticated systems of calculation, as a supporting to the decision-making. SISIFO-GAS is a flexible computerized tool, both for the supporting to accident management and for education and training in severe accident. It is an interactive system, a visual and an easily handle one, and needs no specific knowledge in MAAP code to make complicate simulations in conditions of severe accident. The system is configured and adjusted to work in a BWR/6 technology plant with Mark III Containment, as it is Cofrentes NPP. But it is easily portable to every other kind of reactor, having the level 2 PSA (probabilistic safety analysis) of the plant to be able to establish the categories of the source term and the most important sequences in the progression of the accident. The graphic interface allows following in a very intuitive and formative way the evolution and the most relevant events in the accident, in the both system's way of work, training and management. (authors)

  9. Development of Highly Survivable Power and Communication System for NPP Instruments under Severe Accident

    International Nuclear Information System (INIS)

    Yoo, Seung J.; Gu, Beom W.; Nguyen, Duy T.; Choi, Bo H.; Rim, Chun T.; Lee, So I.

    2014-01-01

    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

  10. The scenario-based system of workers training to prevent accidents during decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    Jeong, KwanSeong; Choi, ByungSeon; Moon, JeiKwon; Hyun, DongJun; Lee, JongHwan; Kim, IkJune; Kim, GeunHo; Seo, JaeSeok

    2014-01-01

    Highlights: • This paper is meant to develop the training system to prevent accidents during decommissioning of nuclear facilities. • Requirements of the system were suggested. • Data management modules of the system were designed. • The system was developed on virtual reality environment. - Abstract: This paper is meant to develop the training system to prevent accidents during decommissioning of nuclear facilities. Requirements of the system were suggested. Data management modules of the system were designed. The system was developed on virtual reality environment. The performance test of the system was proved to be appropriate to decommissioning of nuclear facilities

  11. A STUDY OF LEAKAGE OF TRACE METALS FROM CORROSION OF THE MUNICIPAL DRINKING WATER DISTRIBUTION SYSTEM

    Directory of Open Access Journals (Sweden)

    M.R SHA MANSOURI

    2003-09-01

    Full Text Available Introduction: A high portion of lead and copper concentration in municipal drinking water is related to the metallic structure of the distribution system and facets. The corrosive water in pipes and facets cause dissolution of the metals such as Pb, Cu, Cd, Zn, Fe and Mn into the water. Due to the lack of research work in this area, a study of the trace metals were performed in the drinking water distribution system in Zarin Shahr and Mobareke of Isfahan province. Methods: Based on the united states Environmental protection Agency (USEPA for the cities over than 50,000 population such as Zarin Shahr and Mobareke, 30 water samples from home facets with the minimum 6 hours retention time of water in pipes, were collected. Lead and cadmium concentration were determined using flameless Atomic Absorption. Cupper, Zinc, Iron and Manganese were determined using Atomic Absorption. Results: The average concentration of Pb, Cd, Zn, Fe and Mn in water distribution system fo Zarin Shahr were 5.7, 0.1, 80, 3042, 23065 and in Mobareke were 7.83, 0.8,210,3100, 253, 17µg respectively. The cocentration of Pb, Cd and Zn were zero at the beginning of the water samples from the municipal drinking water distribution system for both cities. Conclusion: The study showed that the corrosion by products (such as Pb, Cd and Zn was the results of dissolution of the galvanized pipes and brass facets. Lead concentration in over that 10 percent of the water samples in zarin shahr exceeded the drinking water standard level, which emphasize the evaluation and control of corrosion in drinking water distribution systems.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  13. Causative Chain Difference for each Type of Accidents in Japanese Maritime Traffic Systems (MTS

    Directory of Open Access Journals (Sweden)

    Wanginingastuti Mutmainnah

    2017-09-01

    Full Text Available Causative chain (CC is a failure chain that cause accident as an outcome product of the second step of MOP model, namely line relation analysis (LRA. This CC is a connection of several causative factors (CF, an outcome product of first step of MOP model, namely corner analysis (CA. MOP Model is an abbreviation from 4M Overturned Pyramid, created by authors by combining 2 accident analysis models. There are two steps in this model, namely CA and LRA. Utilizing this model can know what is CF that happen dominantly to the accidents and what is a danger CC that characterize accidents in a certain place and certain period. By knowing the characteristics, the preventive action can be decided to decrease the number of accident in the next period. The aim of this paper is providing the development of MOP Model that has been upgraded and understanding the characteristics of each type accident. The data that is analyzed in this paper is Japanese accidents from 2008 until 2013, which is available on Japan Transportation Safety Board (JTSB’s website. The analysis shows that every type of accidents has a unique characteristic, shown by their CFs and CCs. However, Man Factor is still playing role to the system dominantly.

  14. Gross Containment Leakage Monitoring System (GCLM) applied to accidental impairment of containment integrity determination

    International Nuclear Information System (INIS)

    Dinu, Camelia; Talpalariu, A.; Constantinescu, G.

    2007-01-01

    The Prioritization of Generic Safety Issues (NUREG-0933 of October 2006), section 1 task II.E.4 item II.E.4.3 recommends that a method of periodic or continuous testing has to be available, in order to detect unknown gross openings in the nuclear power plants containment structure. The Palisades incident and three other incidents are exemplified, when the reactor was operated for about 1.5 years, while the containment isolation valves in a purge system bypass line were unknowingly locked in the open position. It was estimated that the presence of a GCLM system could identify an unknown breach and reduce the expected unavailability of containment due to containment integrity breach events, to a 1.6x10 -3 /year demand. (authors)

  15. Passive depressurization accident management strategy for boiling water reactors

    International Nuclear Information System (INIS)

    Liu, Maolong; Erkan, Nejdet; Ishiwatari, Yuki; Okamoto, Koji

    2015-01-01

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident

  16. Passive depressurization accident management strategy for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong, E-mail: liuml@vis.t.u-tokyo.ac.jp [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Erkan, Nejdet [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Hitachi-GE Nuclear Energy, Ltd. (Japan); Okamoto, Koji [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan)

    2015-04-01

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident.

  17. The leakage problem in vacuum system. Realization of a mass spectrometer detecting leaks

    International Nuclear Information System (INIS)

    Geller, R.

    1954-11-01

    In the first part of this paper we consider the problem of leaks in vacuum systems, and their detection. We consider in particular the method of detection by means of a helium spectrometer. The second part deals with the experimental set p. The analyser and the ion source have been studied in great detail, and we have also discussed the technological and mechanical aspects of the apparatus and its performances. (author) [fr

  18. System 80+TM PRA insights on severe accident prevention and mitigation

    International Nuclear Information System (INIS)

    Finnicum, D.J.; Jacob, M.C.; Schneider, R.E.; Weston, R.A.

    2004-01-01

    The System 80 + design is ABB-CE's standardized evolutionary Advanced Light Water Reactor (ALWR) design. It incorporates design enhancements based on Probabilistic Risk Assessment (PRA) insights, guidance from the ALWR Utility Requirements Document (URD), and US NRC's Severe Accident Policy. Major severe accident prevention and mitigation design features of the System 80 + design are described. The results of the System 80 + PRA are presented and the insights gained from the PRA sensitivity analyses are discussed. ABB-CE considered defense-in-depth for accident prevention and mitigation early in the design process and used robust design features to ensure that the System 80 + design achieved a low core damage frequency, low containment conditional failure probability, and excellent deterministic containment performance under severe accident conditions and to ensure that the risk was properly allocated among design features and between prevention and mitigation. (author)

  19. The role of systems availability and operator actions in accident management

    International Nuclear Information System (INIS)

    Lutz, R.J. Jr.; Scobel, J.H.

    1988-01-01

    Traditional analyses of severe accidents, such as those presented in Probabilistic Risk Assessment (PRA) studies of nuclear power stations, have generally been performed on the assumption that all means of cooling the reactor core are lost and that no operator actions to mitigate the consequences or progression of the severe accident are performed. The assumption to neglect the availability of safety systems and operator actions which do not prevent core melting can lead to erroneous conclusions regarding the plant severe accident profile. Recent work in severe accident management has identified the need to perform analyses which consider all systems availabilities and operator actions, irrespective of their contribution to the prevention of core melting. These new analyses have far reaching conclusions. The analysis results indicate an unacceptably high degree of simplicity in the present severe accident analyses for Probabilistic Risk Assessment studies; the simplicity is in the assumption that systems availabilities and operator actions which do not impact core melt frequency can be neglected in the severe accident analyses. This results in overly pessimistic predictions of the time of core melting and the subsequent potential for recovery of core cooling prior to core melting. This simplicity can have a considerable impact on severe accident decision making, particularly in the evaluation of alternate plant design features and the priorities for research studies

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

    International Nuclear Information System (INIS)

    Chakrabarti, Santanu; Massand, O.P.

    1998-01-01

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

  1. Biogenic methane leakage on the Aquitaine Shelf: fluid system characterization from source to emission

    Science.gov (United States)

    Michel, Guillaume; Dupré, Stéphanie; Baltzer, Agnès; Imbert, Patrice; Ehrhold, Axel; Battani, Anne; Deville, Eric

    2017-04-01

    The recent discovery of biogenic methane emissions associated with methane-derived authigenic carbonate mounds along the Aquitaine Shelf edge offshore SW France (140 to 220 m water depth) questions about the initiation and temporal evolution of this fluid system (80 km N-S and 8 km E-W). Based on a multi-data study (including multibeam echosounder, subbottom profiler, single channel sparker seismic, 80 traces air gun seismic data and well cuttings and logs), different scenarii are proposed for the organic matter source levels and migration pathways of the methane. Several evidence of the presence of gas are observed on seismic data and interpreted to be linked to the biogenic system. Single channel sparker seismic lines exhibit an acoustic blanking (between 75-100 ms TWT below seafloor and the first multiple) below the present-day seepage area and westwards up to 8 km beyond the shelf-break. An air gun seismic line exhibits chaotic reflections along 8 km below the seepage area from the seabed down to 700 ms TWT below seafloor. Based on 1) the local geothermal gradient about 26 °C/km and 2) the window for microbial methanogenesis ranging from 4 to 56 °C, the estimation of the bottom limit for biogenic generation window is about 1.5 km below seafloor. Cuttings from 3 wells of the area within the methanogenesis window show average TOC (Total Organic Carbon) of 0.5 %; however, one well shows some coal levels with 30-35 % TOC in the Oligocene between 1490 and 1540 m below seafloor. Geochemical analysis on crushed cuttings evidenced heavy hydrocarbons up to mid-Paleogene, while shallower series did not evidence any. In the first scenario, we propose that methane is sourced from the Neogene prograding system. The 0.5% average TOC is sufficient to generate a large volume of methane over the thickness of this interval (up to 1 km at the shelf break area). In the second scenario, methane would be sourced from the Oligocene coals; however their spatial extension with regard

  2. Containment leakage rate testing requirements

    International Nuclear Information System (INIS)

    Arndt, E.G.

    1992-01-01

    This report presents the status of several documents under revision or development that provide requirements and guidance for testing nuclear power plant containment systems for leakage rates. These documents include the general revision to 10 CFR Part 50, Appendix J; the regulatory guide affiliated with the revision to Appendix J; the national standard that the regulatory guide endorses, ANSI/ANS-56.8, 'Containment System Leakage Rate Testing Requirements'; and the draft industry Licensing Topical Report, 'Standardized Program for Primary Containment Integrity Testing'. The actual or potential relationships between these documents are also explored

  3. NPP post-accident monitoring system based on unmanned aircraft vehicle:concept, design principles

    International Nuclear Information System (INIS)

    Sachenko, A.A.; Kochan, V.V.; Kharchenko, V.S.; Yanovskij, M.Eh.; Yastrebenetskij, M.A.; Fesenko, G.V.

    2016-01-01

    The paper presents a concept of designing the post-accident system for monitoring the equipment and territory of nuclear power plant after a severe accident based on unmanned aircraft vehicle (UAVs). Wired power and communications networks are found out as the most vulnerable ones during the accident monitoring, and informativity, reliability and veracity are recognized as system basic parameters. It is proposed to equip measurement and control modules with backup wireless communication channels and deploy the repeaters network based on UAVs to ensure the informativity. Modules possess the backup power battery, and repeaters appear in the appropriate places after the accident to provide the survivability. Moreover, an optimization of UAVs' location is proposed according to the minimum energy consumption criterion. To ensure the veracity, it is expected to design the noise-immune protocol for message exchange and archiving and self-diagnostics of all system components

  4. Analysis of helium purification system capability during water ingress accident in RDE

    Science.gov (United States)

    Sriyono; Kusmastuti, Rahayu; Bakhri, Syaiful; Sunaryo, Geni Rina

    2018-02-01

    The water ingress accident caused by steam generator tube rupture (SGTR) in RDE (Experimental Power Reactor) must be anticipated. During the accident, steam from secondary system diffused and mixed with helium gas in the primary coolant. To avoid graphite corrosion in the core, steam will be removed by Helium purification system (HPS). There are two trains in HPS, first train for normal operation and the second for the regeneration and accident. The second train is responsible to clean the coolant during accident condition. The second train is equipped with additional component, i.e. water cooler, post accident blower, and water separator to remove this mixture gas. During water ingress, the water release from rupture tube is mixed with helium gas. The water cooler acts as a steam condenser, where the steam will be separated by water separator from the helium gas. This paper analyses capability of HPS during water ingress accident. The goal of the research is to determine the time consumed by HPS to remove the total amount of water ingress. The method used is modelling and simulation of the HPS by using ChemCAD software. The BDBA and DBA scenarios will be simulated. In BDBA scenario, up to 110 kg of water is assumed to infiltrate to primary coolant while DBA is up to 35 kg. By using ChemCAD simulation, the second train will purify steam ingress maximum in 0.5 hours. The HPS of RDE has a capability to anticipate the water ingress accident.

  5. Guidelines for system modeling: pre-accident human errors, rev.0

    International Nuclear Information System (INIS)

    Kang, Dae Il; Jung, W. D.; Lee, Y. H.; Hwang, M. J.; Yang, J. E.

    2004-01-01

    The evaluation results of Human Reliability Analysis (HRA) of pre-accident human errors in the probabilistic safety assessment (PSA) for the Korea Standard Nuclear Power Plant (KSNP) using the ASME PRA standard show that more than 50% of 10 items to be improved are related to the identification and screening analysis for them. Thus, we developed a guideline for modeling pre-accident human errors for the system analyst to resolve some items to be improved for them. The developed guideline consists of modeling criteria for the pre-accident human errors (identification, qualitative screening, and common restoration errors) and detailed guidelines for pre-accident human errors relating to testing, maintenance, and calibration works of nuclear power plants (NPPs). The system analyst use the developed guideline and he or she applies it to the system which he or she takes care of. The HRA analyst review the application results of the system analyst. We applied the developed guideline to the auxiliary feed water system of the KSNP to show the usefulness of it. The application results of the developed guideline show that more than 50% of the items to be improved for pre-accident human errors of auxiliary feed water system are resolved. The guideline for modeling pre-accident human errors developed in this study can be used for other NPPs as well as the KSNP. It is expected that both use of the detailed procedure, to be developed in the future, for the quantification of pre-accident human errors and the guideline developed in this study will greatly enhance the PSA quality in the HRA of pre-accident human errors

  6. Guidelines for system modeling: pre-accident human errors, rev.0

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dae Il; Jung, W. D.; Lee, Y. H.; Hwang, M. J.; Yang, J. E

    2004-01-01

    The evaluation results of Human Reliability Analysis (HRA) of pre-accident human errors in the probabilistic safety assessment (PSA) for the Korea Standard Nuclear Power Plant (KSNP) using the ASME PRA standard show that more than 50% of 10 items to be improved are related to the identification and screening analysis for them. Thus, we developed a guideline for modeling pre-accident human errors for the system analyst to resolve some items to be improved for them. The developed guideline consists of modeling criteria for the pre-accident human errors (identification, qualitative screening, and common restoration errors) and detailed guidelines for pre-accident human errors relating to testing, maintenance, and calibration works of nuclear power plants (NPPs). The system analyst use the developed guideline and he or she applies it to the system which he or she takes care of. The HRA analyst review the application results of the system analyst. We applied the developed guideline to the auxiliary feed water system of the KSNP to show the usefulness of it. The application results of the developed guideline show that more than 50% of the items to be improved for pre-accident human errors of auxiliary feed water system are resolved. The guideline for modeling pre-accident human errors developed in this study can be used for other NPPs as well as the KSNP. It is expected that both use of the detailed procedure, to be developed in the future, for the quantification of pre-accident human errors and the guideline developed in this study will greatly enhance the PSA quality in the HRA of pre-accident human errors.

  7. Natural hazard impacts on transport systems: analyzing the data base of transport accidents in Russia

    Science.gov (United States)

    Petrova, Elena

    2015-04-01

    We consider a transport accident as any accident that occurs during transportation of people and goods. It comprises of accidents involving air, road, rail, water, and pipeline transport. With over 1.2 million people killed each year, road accidents are one of the world's leading causes of death; another 20-50 million people are injured each year on the world's roads while walking, cycling, or driving. Transport accidents of other types including air, rail, and water transport accidents are not as numerous as road crashes, but the relative risk of each accident is much higher because of the higher number of people killed and injured per accident. Pipeline ruptures cause large damages to the environment. That is why safety and security are of primary concern for any transport system. The transport system of the Russian Federation (RF) is one of the most extensive in the world. It includes 1,283,000 km of public roads, more than 600,000 km of airlines, more than 200,000 km of gas, oil, and product pipelines, 115,000 km of inland waterways, and 87,000 km of railways. The transport system, especially the transport infrastructure of the country is exposed to impacts of various natural hazards and weather extremes such as heavy rains, snowfalls, snowdrifts, floods, earthquakes, volcanic eruptions, landslides, snow avalanches, debris flows, rock falls, fog or icing roads, and other natural factors that additionally trigger many accidents. In June 2014, the Ministry of Transport of the RF has compiled a new version of the Transport Strategy of the RF up to 2030. Among of the key pillars of the Strategy are to increase the safety of the transport system and to reduce negative environmental impacts. Using the data base of technological accidents that was created by the author, the study investigates temporal variations and regional differences of the transport accidents' risk within the Russian federal regions and a contribution of natural factors to occurrences of different

  8. Evaluation of severe accident safety system value based on averting financial risks

    International Nuclear Information System (INIS)

    Hatch, S.W.; Benjamin, A.S.; Bennett, P.R.

    1983-01-01

    The Severe Accident Risk Reduction Program is being performed to benchmark the risks from nuclear power plants and to assess the benefits and impacts of a set of severe accident safety features. This paper describes the program in general and presents some preliminary results. These results include estimates of the financial risks associated with the operation of six reference plants and the value of severe accident prevention and mitigation safety systems in averting these risks. The results represent initial calculations and will be iterated before being used to support NRC decisions

  9. K Basins floor sludge retrieval system knockout pot basket fuel burn accident

    International Nuclear Information System (INIS)

    HUNT, J.W.

    1998-01-01

    The K Basins Sludge Retrieval System Preliminary Hazard Analysis Report (HNF-2676) identified and categorized a series of potential accidents associated with K Basins Sludge Retrieval System design and operation. The fuel burn accident was of concern with respect to the potential release of contamination resulting from a runaway chemical reaction of the uranium fuel in a knockout pot basket suspended in the air. The unmitigated radiological dose to an offsite receptor from this fuel burn accident is calculated to be much less than the offsite risk evaluation guidelines for anticipated events. However, because of potential radiation exposure to the facility worker, this accident is precluded with a safety significant lifting device that will prevent the monorail hoist from lifting the knockout pot basket out of the K Basin water pool

  10. Study on the code system for the off-site consequences assessment of severe nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

  11. Study on the code system for the off-site consequences assessment of severe nuclear accident

    International Nuclear Information System (INIS)

    Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk

    2016-01-01

    The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents

  12. Advanced Detection Technology of Trace-level Borate for SG Leakage Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seban; Kang, Dukwon; Kim, Seungil; Kim, Hyunki; Heo, Jun; Sung, Jinhyun [Radiation Eng. Center, Shihung (Korea, Republic of); Lee, Dongbum [Academic Support Dept., Seoul (Korea, Republic of)

    2013-05-15

    Many studies have been reported for monitoring technology of steam generator, however, all of these methods have their own limitations. The leakage monitoring technology of steam generator of PWR has also got a limit due to the adoption of specific radionuclides (N-16, Ar-41, H-3, Xe, etc.) generated by nuclear fission, which are available only when reactor output is 20% or more. Most of domestic NPPs apply the N-16 technique for monitoring tube leakage but it has some problem that it is difficult to calculate the leakage rate because neutron flux are not completely formed during low power operation. For example, tube leakage of steam generator occurred in the Uljin nuclear power plant in 2002 during coast down operation for periodic plant maintenance. This plant could not prevent a rupture accident in advance because N-16 method is not possible the leak monitoring less than 20% reactor power. The development of excellent alternative monitoring technology that can monitor the real-time leakage is required under a variety of operating conditions like start-up and abnormal conditions of NPPs. This study was performed to lay a foundation in monitoring the leakage of steam generator coping with the lower output and low power operational condition using trace level of boron which is non-radioactive nuclide to inject control neutron injection. In this study, non-radioactive nuclide boron ion, which existed in the secondary system water, as leakage monitoring indicator was investigated for the separation of complex cation and anion phase. Borate was detected by using borate concentrator column coupled with the ion-exclusion column analytical column, revealing the problem of overlapped peak between fluoride and boron ions. Meanwhile, ion-exchange column could confirm the possibility as a leakage monitoring indicator of steam generator, despite the peak of glycolic acid salts was slightly overlapped. It will be needed for further research regarding the selectivity of the

  13. Advanced Detection Technology of Trace-level Borate for SG Leakage Monitoring

    International Nuclear Information System (INIS)

    Lee, Seban; Kang, Dukwon; Kim, Seungil; Kim, Hyunki; Heo, Jun; Sung, Jinhyun; Lee, Dongbum

    2013-01-01

    Many studies have been reported for monitoring technology of steam generator, however, all of these methods have their own limitations. The leakage monitoring technology of steam generator of PWR has also got a limit due to the adoption of specific radionuclides (N-16, Ar-41, H-3, Xe, etc.) generated by nuclear fission, which are available only when reactor output is 20% or more. Most of domestic NPPs apply the N-16 technique for monitoring tube leakage but it has some problem that it is difficult to calculate the leakage rate because neutron flux are not completely formed during low power operation. For example, tube leakage of steam generator occurred in the Uljin nuclear power plant in 2002 during coast down operation for periodic plant maintenance. This plant could not prevent a rupture accident in advance because N-16 method is not possible the leak monitoring less than 20% reactor power. The development of excellent alternative monitoring technology that can monitor the real-time leakage is required under a variety of operating conditions like start-up and abnormal conditions of NPPs. This study was performed to lay a foundation in monitoring the leakage of steam generator coping with the lower output and low power operational condition using trace level of boron which is non-radioactive nuclide to inject control neutron injection. In this study, non-radioactive nuclide boron ion, which existed in the secondary system water, as leakage monitoring indicator was investigated for the separation of complex cation and anion phase. Borate was detected by using borate concentrator column coupled with the ion-exclusion column analytical column, revealing the problem of overlapped peak between fluoride and boron ions. Meanwhile, ion-exchange column could confirm the possibility as a leakage monitoring indicator of steam generator, despite the peak of glycolic acid salts was slightly overlapped. It will be needed for further research regarding the selectivity of the

  14. Application of GIS in prediction and assessment system of off-site accident consequence for NPP

    International Nuclear Information System (INIS)

    Wang Xingyu; Shi Zhongqi

    2002-01-01

    The assessment and prediction software system of off-site accident consequence for Guangdong Nuclear Power Plant (GNARD2.0) is a GIS-based software system. The spatial analysis of radioactive materials and doses with geographic information is available in this system. The structure and functions of the GNARD system and the method of applying ArcView GIS are presented

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

    International Nuclear Information System (INIS)

    Gieseke, J.A.

    1985-01-01

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

  16. Classification Of Road Accidents From The Perspective Of Vehicle Safety Systems

    Directory of Open Access Journals (Sweden)

    Jirovský Václav

    2015-11-01

    Full Text Available Modern road accident investigation and database structures are focused on accident analysis and classification from the point of view of the accident itself. The presented article offers a new approach, which will describe the accident from the point of view of integrated safety vehicle systems. Seven main categories have been defined to specify the level of importance of automated system intervention. One of the proposed categories is a new approach to defining the collision probability of an ego-vehicle with another object. This approach focuses on determining a 2-D reaction space, which describes all possible positions of the vehicle or other moving object in the specified amount of time in the future. This is to be used for defining the probability of the vehicles interacting - when the intersection of two reaction spaces exists, an action has to be taken on the side of ego-vehicle. The currently used 1-D quantity of TTC (time-to-collision can be superseded by the new reaction space variable. Such new quantity, whose basic idea is described in the article, enables the option of counting not only with necessary braking time, but mitigation by changing direction is then easily feasible. Finally, transparent classification measures of a probable accident are proposed. Their application is highly effective not only during basic accident comparison, but also for an on-board safety system.

  17. In Vitro Evaluation of Leakage at Implant-Abutment Connection of Three Implant Systems Having the Same Prosthetic Interface Using Rhodamine B

    Directory of Open Access Journals (Sweden)

    Antoine Berberi

    2014-01-01

    Full Text Available Objectives. Hollow space between implant and abutment may act as reservoir for commensal and/or pathogenic bacteria representing a potential source of tissue inflammation. Microbial colonization of the interfacial gap may ultimately lead to infection and bone resorption. Using Rhodamine B, a sensitive fluorescent tracer dye, we aim in this study to investigate leakage at implant-abutment connection of three implant systems having the same prosthetic interface. Materials and Methods. Twenty-one implants (seven Astra Tech, seven Euroteknika, and seven Dentium with the same prosthetic interface were connected to their original abutments, according to the manufacturers’ recommendation. After determination of the inner volume of each implant systems, the kinetic quantification of leakage was evaluated for each group using Rhodamine B (10−2 M. For each group, spectrophotometric analysis was performed to detect leakage with a fluorescence spectrophotometer at 1 h (T0 and 48 h (T1 of incubation time at room temperature. Results. Astra Tech had the highest inner volume (6.8 μL, compared to Dentium (4 μL and Euroteknika (2.9 μL. At T0 and T1, respectively, the leakage volume and percentage of each system were as follows: Astra Tech 0.043 μL or 1.48% (SD 0.0022, 0.08 μL or 5.56% (SD 0.0074, Euroteknika 0.09 μL or 6.93% (SD 0.0913, 0.21 μL or 20.55% (SD 0.0035, and Dentium 0.07 μL or 4.6% (SD 0.0029, 0.12 μL or 10.47% (SD 0.0072. Conclusion. The tested internal conical implant-abutment connections appear to be unable to prevent leakage. In average, Astra Tech implants showed the highest inner volume and the least leakage.

  18. System Design Strategies of Post-Accident Monitoring System for a PGSFR in Korea

    International Nuclear Information System (INIS)

    Jang, Gwi-sook; Jeong, Kwang-il; Keum, Jong-yong; Seong, Seung-hwan

    2013-06-01

    Monitoring systems of a PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) in Korea provide alarms, integrity information in the reactor building, sodium-water reaction information in the steam generator, fuel failure information, and supporting information for maintenance and inspection. In particular, a Post-Accident Monitoring System (PAMS) provides primary information for operators to assess the plant conditions and perform their role in bringing the plant to a safe condition during an accident. Some PAM variables can be allocated as more two types. It is important for system designers to confirm the suitability of the selection of PAM variables. In addition, the PAMS is a position 4 display against common cause failures of safety I and C systems. The position 4 display should be independent and diverse from the safety I and C systems. The diversity of safety I and C equipment has led to an increase in the design and verification and validation cost. Thus, this paper proposes the system design strategies on the PAMS design problems of the PGSFR in KOREA. The results will be input into a conceptual system design for the PAMS of the PGSFR in KOREA. (authors)

  19. Water level measurement system in reactor pressure vessel of BWR and hydrogen concentration monitoring system for severe accident

    International Nuclear Information System (INIS)

    Kuroda, Hidehiko; Okazaki, Koki; Shiraishi, Fujio; Kenjyo, Hiroaki; Isoda, Koichiro

    2013-01-01

    TEPCO's Fukushima Daiichi Nuclear Power Station Accident caused severe accident to lose functions of many instrumentation systems. As a result, many important plant parameters couldn't be monitored. In order to monitor plant parameters in the case of severe accident, new instrumentation systems available in the severe conditions are being developed. Water level in reactor pressure vessel and hydrogen concentration in primary containment vessel are one of the most important parameters. Performance test results about water level measurement sensor and hydrogen sensor in severe environmental conditions are described. (author)

  20. Status and functioning of the European Commission's major accident reporting system

    International Nuclear Information System (INIS)

    Kirchsteiger, C.

    1999-01-01

    This paper describes the background, functioning and status of the European Commission's Major Accident Reporting System (MARS), dedicated to collect, in a consistent way, data on major industrial accidents involving dangerous substances from the Member States of the European Union, to analyse and statistically process them, and to create subsets of all non-confidential accidents data and analysis results for export to all Member States. This modern information exchange and analysis tool is made up of two connected parts: one for each local unit (i.e., for the Competent Authority of each EU Member State), and one central part for the European Commission. The local, as well as the central parts of this information network, can serve both as data logging systems and, on different levels of complexity, as data analysis tools. The central database allows complex cluster and pattern analysis, identifying and analysing the succession of the disruptive factors leading to an accident. On this basis, 'lessons learned' can be formulated for the industry for the purposes of further accident prevention. Further, results from analysing data of major industrial accidents reported to MARS are presented. It can be shown that some of the main assumptions in the new 'Seveso II Directive' can directly be validated from MARS data. (Copyright (c) 1999 Elsevier Science B.V., Amsterdam. All rights reserved.)

  1. Land-use Leakage

    Energy Technology Data Exchange (ETDEWEB)

    Calvin, Katherine V.; Edmonds, James A.; Clarke, Leon E.; Bond-Lamberty, Benjamin; Kim, Son H.; Wise, Marshall A.; Thomson, Allison M.; Kyle, G. Page

    2009-12-01

    Leakage occurs whenever actions to mitigate greenhouse gas emissions in one part of the world unleash countervailing forces elsewhere in the world so that reductions in global emissions are less than emissions mitigation in the mitigating region. While many researchers have examined the concept of industrial leakage, land-use policies can also result in leakage. We show that land-use leakage is potentially as large as or larger than industrial leakage. We identify two potential land-use leakage drivers, land-use policies and bioenergy. We distinguish between these two pathways and run numerical experiments for each. We also show that the land-use policy environment exerts a powerful influence on leakage and that under some policy designs leakage can be negative. International “offsets” are a potential mechanism to communicate emissions mitigation beyond the borders of emissions mitigating regions, but in a stabilization regime designed to limit radiative forcing to 3.7 2/m2, this also implies greater emissions mitigation commitments on the part of mitigating regions.

  2. Data leakage quantification

    NARCIS (Netherlands)

    Vavilis, S.; Petkovic, M.; Zannone, N.; Atluri, V.; Pernul, G.

    2014-01-01

    The detection and handling of data leakages is becoming a critical issue for organizations. To this end, data leakage solutions are usually employed by organizations to monitor network traffic and the use of portable storage devices. These solutions often produce a large number of alerts, whose

  3. Modeling valve leakage

    International Nuclear Information System (INIS)

    Bell, S.R.; Rohrscheib, R.

    1994-01-01

    The American Society of Mechanical Engineers (ASME) Code requires individual valve leakage testing for Category A valves. Although the U.S. Nuclear Regulatory Commission (USNRC) has recognized that it is more appropriate to test containment isolation valves in groups, as allowed by 10 CFR 50, Appendix J, a utility seeking relief from these Code requirements must provide technical justification for the relief and establish a conservative alternate acceptance criteria. In order to provide technical justification for group testing of containment isolation valves, Illinois Power developed a calculation (model) for determining the size of a leakage pathway in a valve disc or seat for a given leakage rate. The model was verified experimentally by machining leakage pathways of known size and then measuring the leakage and comparing this value to the calculated value. For the range of values typical of leakage rate testing, the correlation between the experimental values and calculated values was quote good. Based upon these results, Illinois Power established a conservative acceptance criteria for all valves in the inservice testing (IST) program and was granted relief by the USNRC from the individual leakage testing requirements of the ASME Code. This paper presents the results of Illinois Power's work in the area of valve leakage rate testing

  4. Solved and unsolved problems in boiler systems. Learning from accidents

    International Nuclear Information System (INIS)

    Ozawa, Mamoru

    2000-01-01

    This paper begins with a brief review on the similarity law of conventional fossil-fuel-fired boilers. The concept is based on the fact that the heat release due to combustion in the furnace is restricted by the furnace volume but the heat absorption is restricted by the heat transfer surface area. This means that a small-capacity boiler has relatively high specific furnace heat release rate, about 10 MW/m 3 , and on the contrary a large-capacity boiler has lower value. The surface-heat-flux limit is mainly dominated by the CHF inside the water-wall tubes of the boiler furnace, about 350 kW/m 2 . This heat-flux limit is almost the same order independently on the capacity of boilers. For the safety of water-walls, it is essential to retain suitable water circulation, i.e. circulation ratio and velocity of water. This principle is a common knowledge of boiler designer, but actual situation is not the case. Newly designed boilers often suffer from similar accidents, especially burnout due to circulation problems. This paper demonstrates recent accidents encountered in practical boilers, and raises problems of rather classical but important two-phase flow and heat transfer. (author)

  5. Risk Analysis on Leakage Failure of Natural Gas Pipelines by Fuzzy Bayesian Network with a Bow-Tie Model

    Directory of Open Access Journals (Sweden)

    Xian Shan

    2017-01-01

    Full Text Available Pipeline is the major mode of natural gas transportation. Leakage of natural gas pipelines may cause explosions and fires, resulting in casualties, environmental damage, and material loss. Efficient risk analysis is of great significance for preventing and mitigating such potential accidents. The objective of this study is to present a practical risk assessment method based on Bow-tie model and Bayesian network for risk analysis of natural gas pipeline leakage. Firstly, identify the potential risk factors and consequences of the failure. Then construct the Bow-tie model, use the quantitative analysis of Bayesian network to find the weak links in the system, and make a prediction of the control measures to reduce the rate of the accident. In order to deal with the uncertainty existing in the determination of the probability of basic events, fuzzy logic method is used. Results of a case study show that the most likely causes of natural gas pipeline leakage occurrence are parties ignore signage, implicit signage, overload, and design defect of auxiliaries. Once the leakage occurs, it is most likely to result in fire and explosion. Corresponding measures taken on time will reduce the disaster degree of accidents to the least extent.

  6. 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.

  7. Model for Electromagnetic Information Leakage

    OpenAIRE

    Mao Jian; Li Yongmei; Zhang Jiemin; Liu Jinming

    2013-01-01

    Electromagnetic leakage will happen in working information equipments; it could lead to information leakage. In order to discover the nature of information in electromagnetic leakage, this paper combined electromagnetic theory with information theory as an innovative research method. It outlines a systematic model of electromagnetic information leakage, which theoretically describes the process of information leakage, intercept and reproduction based on electromagnetic radiation, and ana...

  8. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Jin; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures.

  9. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    International Nuclear Information System (INIS)

    Lee, Young Jin; Chung, Bub Dong

    2004-01-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures

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

    International Nuclear Information System (INIS)

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

    1987-04-01

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

  11. REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1982-01-01

    REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

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

    International Nuclear Information System (INIS)

    Ireland, J.R.

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    Putman, V.L.

    1995-09-01

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

  14. [Analysis of accidents for magnetically induced displacement of the large ferromagnetic material in magnetic resonance systems].

    Science.gov (United States)

    Yamatani, Yuya; Doi, Tsukasa; Ueyama, Tsuyoshi; Nishiki, Shigeo; Ogura, Akio; Kawamitsu, Hideaki; Tsuchihashi, Toshio; Okuaki, Tomoyuki; Matsuda, Tsuyoshi

    2013-01-01

    To improve magnetic resonance (MR) safety, we surveyed the accidents caused by large ferromagnetic materials brought into MR systems accidentally. We sent a questionnaire to 700 Japanese medical institutions and received 405 valid responses (58%). A total of 97 accidents in 77 institutions were observed and we analyzed them regarding incidental rate, the detail situation and environmental factors. The mean accident rate of each institute was 0.7/100,000 examinations, which was widely distributed (0-25.6/100,000) depending on the institute. In this survey, relatively small institutes with less than 500 beds tend to have these accidents more frequently (paccidents than those with less than 10 daily examinations. The institutes with 6-10 MR examinations daily have significantly more accidents than that with more than 10 daily MR examinations (paccidents were considered to be "prejudice" and "carelessness" but some advocate "ignorance." Though we could not find significant reduction in the institutes that have lectures and training for MR safety, we should continue lectures and training for MR safety to reduce accidents due to "ignorance."

  15. Acoustic control of sodium leakage in valve gates of NPP

    International Nuclear Information System (INIS)

    Trykov, E.L.; Kovtun, S.N.; Anan'ev, A.A.; Yugov, S.I.

    2014-01-01

    Short description of sodium bench and acoustic investigation results on leakage monitoring of valves DN10 and DN40 are given. It is shown that acoustic method can be used successfully to control the leakages of sodium valves. Leakages on both type of valves increase the acoustic signal dispersion by 2-3 orders. For each type of valve acoustic system of leakage determination allows to conduct the preliminary graduation of signal dispersion on the sodium discharge rate. It make possible not only to record the leakage presence but also to determine the sodium discharge rate through the valve during the leakage [ru

  16. Innovative leakage monitoring for local and district heating systems. Development of a new device; Innovative Leckageueberwachung fuer Nah- und Fernwaermesysteme. Entwicklung eines neuen Geraets

    Energy Technology Data Exchange (ETDEWEB)

    Neugebauer, Horst [Stadtwerke Rosenheim (Germany). Abt. Fernwaermenetz; Schober, Wilhelm [Elnic, Rosenheim (Germany)

    2013-07-15

    District heating plants and local heating systems are one of the most important future technologies for the transition from fossil fuels to renewable energies. In order to hedge the cost and to extend the life of these systems, it is even more important in the future to pay more attention to a preventive monitoring of leakages of pipeline routes. Together with the development specialists Elnic (Rosenheim, Federal Republic of Germany), the Stadtwerke Rosenheim (Federal Republic of Germany) have brought a new leak detection system on the market.

  17. Technique of research of severe accidents and substantiation of safety of nuclear systems

    International Nuclear Information System (INIS)

    Ivanov, E.A.; Tchenov, S.V.

    2001-01-01

    Work is devoted to development of possible ways of solution of the problems of nuclear safety substantiation. We believe that safety in severe accidents is one of significant factors, which restrict value of nuclear industry in future power production. In connection with it we can conclude followed items: -) Substantiation of safety in severe accidents in nuclear system should be built on a deterministic way of guaranteed exception of heavy consequences; -) It is easy that this aim can be achieved by modeling in functions of common type; -) Main purpose of this work is to show that it is possible to estimate physical allowed state of system in emergency and find of trajectory of heaviest scenarios by optimization procedure; and -) In this work we have developed new method and computer code purposed for study of accident conditions of water cooled un-managed nuclear systems such as cooling ponds of spent fuel, experimental facilities etc. (authors)

  18. Effect of crack size on gas leakage characteristics in a confined space

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Kun Hyuk; Ryou, Hong Sun; Yoon, Kee Bong; Lee, Hy Uk; Bang, Joo Won [Chung-Ang University, Seoul (Korea, Republic of); Li, Longnan; Choi, Jin Wook; Kim, Dae Joong [Sogang University, Seoul (Korea, Republic of)

    2016-07-15

    We numerically investigated the influence of crack size on gas leakage characteristics in a confined space. The real scale model of underground Combined cycle power plant (CCPP) was taken for simulating gas leakage characteristics for different crack sizes such as 10 mm, 15 mm and 20 mm. The commercial code of Fluent (v.16.1) was used for three-dimensional simulation. In particular, a risk region showing such a probability of ignition was newly suggested with the concept of Lower flammable limit (LFL) of methane gas used in the present study to characterize the gas propagation and the damage area in space. From the results, the longitudinal and transverse leakage distances were estimated and analyzed for quantitative evaluation of risk area. The crack size was found to have a great impact on the longitudinal leakage distance, showing an increasing tendency with the crack size. In case of a crack size of 20 mm, the longitudinal leakage distance suddenly increased after 180 s, whereas it remained constant after 2 s in the other cases. This is because a confinement effect, which is caused by circulation flows in the whole space, increased the gas concentration near the gas flow released from the crack. The confinement effect is thus closely associated with the released mass flow rate changing with the crack size. This result would be useful in designing the gas detector system for preventing accidents in the confined space as like CCPP.

  19. The system of emergency cards for primary actions in accident at radioactive material transport in Russia

    International Nuclear Information System (INIS)

    Ananiev, V.V.; Ermakov, S.V.; Ershov, V.N.; Stovbur, V.I.; Shvedov, M.O.

    2004-01-01

    In the paper are reviewed the current and new designed system of the emergency cards for consignments of radioactive materials in Russian Federation, within the framework of a uniform state system of warning and liquidation of consequences of extraordinary situations and functional subsystem of warning and liquidation of accident situations of Federal Agency for Atomic Energy

  20. Current statistical tools, systems and bodies concerned with safety and accident statistics.

    NARCIS (Netherlands)

    Koornstra, M.J.

    1996-01-01

    There are a wide range of differences in the methods used nationally to classify and record road accidents. The current use of road safety information systems and the few systems available for international use are discussed. Recommendations are made for a more efficient, less costly, and improved

  1. Investigation into the causes of accidents on scraper systems in the gold and platinum mining sectors

    CSIR Research Space (South Africa)

    Moseme, R

    2003-11-01

    Full Text Available and cleaning operations of the scraper winch systems that require identification. This research report identifies the risk and hazards associated with scraper winch systems that may lead to potential accidents in the gold and platinum sector. The research also...

  2. The system of emergency cards for primary actions in accident at radioactive material transport in Russia

    Energy Technology Data Exchange (ETDEWEB)

    Ananiev, V.V. [Div. of the Decommission of Nuclear and Radiation-Hazardous Object of the Federal Agency for Atomic Energy, Moscow (Russian Federation); Ermakov, S.V.; Ershov, V.N.; Stovbur, V.I. [FGUP ' ' Emergency Response Centre of Minatom of Russia' ' , St-Petersburg (Russian Federation); Shvedov, M.O. [Div. of Nuclear and Radiation Safety of the Federal Agency for Atomic Energy, Moscow (Russian Federation)

    2004-07-01

    In the paper are reviewed the current and new designed system of the emergency cards for consignments of radioactive materials in Russian Federation, within the framework of a uniform state system of warning and liquidation of consequences of extraordinary situations and functional subsystem of warning and liquidation of accident situations of Federal Agency for Atomic Energy.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  4. Possibility of the development of a Serbian protection system against chemical accidents

    Directory of Open Access Journals (Sweden)

    Dejan R. Inđić

    2012-10-01

    Full Text Available The paper presents a draft of a system model for responding in case of chemical accidents in accordance with the current legislation regarding the environment protection, the structure and elements of the existing response system in case of chemical accidents, other works dealing with the issue as well as the prospects planned by those responsible for the environmental protection. The paper discuss the possibilities of different institutions and agencies of the Republic of Serbia to engage in specialized methods of cooperation and protection against chemical hazards in accordance with Article X of the Convention on the Prohibition of Chemical Weapons.

  5. Radiofrequency radiation leakage from microwave ovens

    International Nuclear Information System (INIS)

    Lahham, A.; Sharabati, A.

    2013-01-01

    This work presents data on the amount of radiation leakage from 117 microwave ovens in domestic and restaurant use in the West Bank, Palestine. The study of leakage is based on the measurements of radiation emissions from the oven in real-life conditions by using a frequency selective field strength measuring system. The power density from individual ovens was measured at a distance of 1 m and at the height of centre of door screen. The tested ovens were of different types, models with operating powers between 1000 and 1600 W and ages ranging from 1 month to >20 y, including 16 ovens with unknown ages. The amount of radiation leakage at a distance of 1 m was found to vary from 0.43 to 16.4 μW cm -1 with an average value equalling 3.64 μW cm -2 . Leakages from all tested microwave ovens except for seven ovens (∼6 % of the total) were below 10 μW cm -2 . The highest radiation leakage from any tested oven was ∼16.4 μW cm -2 , and found in two cases only. In no case did the leakage exceed the limit of 1 μWcm -1 recommended by the ICNIRP for 2.45-GHz radiofrequency. This study confirms a linear correlation between the amount of leakage and both oven age and operating power, with a stronger dependence of leakage on age. (authors)

  6. Multi-phase model development to assess RCIC system capabilities under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kirkland, Karen Vierow [Texas A & M Univ., College Station, TX (United States); Ross, Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Beeny, Bradley [Texas A & M Univ., College Station, TX (United States); Luthman, Nicholas [Texas A& M Engineering Experiment Station, College Station, TX (United States); Strater, Zachary [Texas A & M Univ., College Station, TX (United States)

    2017-12-23

    The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that the system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.

  7. Study on integrated approach of Nuclear Accident Hazard Predicting, Warning, and Optimized Controlling System based on GIS

    International Nuclear Information System (INIS)

    Tang Lijuan; Huang Shunxiang; Wang Xinming

    2012-01-01

    The issue of nuclear safety becomes the attention focus of international society after the nuclear accident happened in Fukushima. Aiming at the requirements of the prevention and controlling of Nuclear Accident establishment of Nuclear Accident Hazard Predicting, Warning and optimized Controlling System (NAPWS) is a imperative project that our country and army are desiderating, which includes multiple fields of subject as nuclear physics, atmospheric science, security science, computer science and geographical information technology, etc. Multiplatform, multi-system and multi-mode are integrated effectively based on GIS, accordingly the Predicting, Warning, and Optimized Controlling technology System of Nuclear Accident Hazard is established. (authors)

  8. Development of reactor accident diagnostic system DISKET using knowledge engineering technique

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Kohsaka, Atsuo; Yamamoto, Minoru.

    1986-01-01

    An accident diagnostic system DISKET has been developed to identify the cause and the type of an abnormal transient of a nuclear power plant. The system is based on the knowledge engineering (KE) and consists of an inference engine IERIAS and a knowledge base. The main features of DISKET are the following : (1) Time-varying characteristics of transients can be treated. (2) Knowledge base can be divided into several knowledge units to handle a lot of rules effectively. (3) Programming language UTILISP, which is a dialect of LISP, is used to manipulate symbolic data effectively. For the verification of DISKET, performance tests have been conducted for several types of accidents. The knowledge base used in the tests was generated from the data of various types of transients produced by a PWR plant simulator. The results of verification studies showed a good applicability of DISKET to reactor accident diagnosis. (author)

  9. Applying of Reliability Techniques and Expert Systems in Management of Radioactive Accidents

    International Nuclear Information System (INIS)

    Aldaihan, S.; Alhbaib, A.; Alrushudi, S.; Karazaitri, C.

    1998-01-01

    Accidents including radioactive exposure have variety of nature and size. This makes such accidents complex situations to be handled by radiation protection agencies or any responsible authority. The situations becomes worse with introducing advanced technology with high complexity that provide operator huge information about system working on. This paper discusses the application of reliability techniques in radioactive risk management. Event tree technique from nuclear field is described as well as two other techniques from nonnuclear fields, Hazard and Operability and Quality Function Deployment. The objective is to show the importance and the applicability of these techniques in radiation risk management. Finally, Expert Systems in the field of accidents management are explored and classified upon their applications

  10. Qualitative analysis of the man-organization system in accident conditions for nuclear installations

    International Nuclear Information System (INIS)

    Farcasiu, Mita; Prisecaru, Ilie

    2010-01-01

    In this paper a model of the human performance investigation of accident conditions in the operation of the nuclear installation is developed. A framework for analyses of the human action in the man-organization system context is achieved. The goal of this model is to identify the possible roots causing human errors which could occur during the evolution of the accident by the qualitative analysis of the interfaces in man-organization system. These interfaces represent the main elements which characterize the implication of the organization in human performance. The results of this paper are the interfaces of the man-organization and their circumstances in which human performance could fail. Also, another result is a pre-designed framework which could help in the investigation of an accident. (authors)

  11. Accidents involving Brazilian indigenous treated at urgent and emergency services of the Unified Health System.

    Science.gov (United States)

    Souza, Edinilsa Ramos de; Njaine, Kathie; Mascarenhas, Márcio Dênis Medeiros; Oliveira, Maria Conceição de

    2016-12-01

    Abstract We analyzed the accidents with Brazilian indigenous treated at urgent and emergency services of the Unified Health System (SUS). Data were obtained from the 2014 Viva Survey, which included 86 services from 24 capitals and the Federal District. The demographic profile of the indigenous, the event and the attendance were characterized. Most of the attended people were male in the 20-39 years age group. Falls and traffic accidents were the main reasons for attendance. Alcohol use was informed by 5.6% of the attended people, a figure that increases to 19.1% in traffic accidents, 26.1% among drivers and 22.8% among motorcyclists. There was a statistical difference between genders in relation to age, disability, place of occurrence of the event, work-related event and victim's condition in the traffic accident. We emphasize the importance of providing visibility to accidents with indigenous and engage them in the prevention of such events. Data reliability depends on the adequate completion in indigenous health information systems.

  12. Universal leakage elimination

    International Nuclear Information System (INIS)

    Byrd, Mark S.; Lidar, Daniel A.; Wu, L.-A.; Zanardi, Paolo

    2005-01-01

    'Leakage' errors are particularly serious errors which couple states within a code subspace to states outside of that subspace, thus destroying the error protection benefit afforded by an encoded state. We generalize an earlier method for producing leakage elimination decoupling operations and examine the effects of the leakage eliminating operations on decoherence-free or noiseless subsystems which encode one logical, or protected qubit into three or four qubits. We find that by eliminating a large class of leakage errors, under some circumstances, we can create the conditions for a decoherence-free evolution. In other cases we identify a combined decoherence-free and quantum error correcting code which could eliminate errors in solid-state qubits with anisotropic exchange interaction Hamiltonians and enable universal quantum computing with only these interactions

  13. Developing a Minimum Data Set for an Information Management System to Study Traffic Accidents in Iran.

    Science.gov (United States)

    Mohammadi, Ali; Ahmadi, Maryam; Gharagozlu, Alireza

    2016-03-01

    Each year, around 1.2 million people die in the road traffic incidents. Reducing traffic accidents requires an exact understanding of the risk factors associated with traffic patterns and behaviors. Properly analyzing these factors calls for a comprehensive system for collecting and processing accident data. The aim of this study was to develop a minimum data set (MDS) for an information management system to study traffic accidents in Iran. This descriptive, cross-sectional study was performed in 2014. Data were collected from the traffic police, trauma centers, medical emergency centers, and via the internet. The investigated resources for this study were forms, databases, and documents retrieved from the internet. Forms and databases were identical, and one sample of each was evaluated. The related internet-sourced data were evaluated in their entirety. Data were collected using three checklists. In order to arrive at a consensus about the data elements, the decision Delphi technique was applied using questionnaires. The content validity and reliability of the questionnaires were assessed by experts' opinions and the test-retest method, respectively. An (MDS) of a traffic accident information management system was assigned to three sections: a minimum data set for traffic police with six classes, including 118 data elements; a trauma center with five data classes, including 57 data elements; and a medical emergency center, with 11 classes, including 64 data elements. Planning for the prevention of traffic accidents requires standardized data. As the foundation for crash prevention efforts, existing standard data infrastructures present policymakers and government officials with a great opportunity to strengthen and integrate existing accident information systems to better track road traffic injuries and fatalities.

  14. Highly Reliable Power and Communication System for Essential Instruments under a Severe Accident of NPPs

    International Nuclear Information System (INIS)

    Yoo, S. J.; Choi, B. H.; Jung, S. Y.; Rim, Chun T.

    2013-01-01

    In this paper, three survivable strategies to overcome the problems listed above are proposed for the essential instruments under the severe accident of NPPs. First, wire/wireless multi power systems are adopted to the essential instruments for continuous power supply. Second, wire/wireless communication systems are proposed for reliable transmission of measuring information among instruments and operators. Third, a physical protection system such as a harness and a heat isolation box is introduced to ensure operable conditions for the proposed systems. In this paper, a highly reliable strategy, which consists of wire/wireless multi power and communication systems and physical protection system is proposed to ensure the survival of the essential instruments under harsh external conditions. The wire/wireless multi power and communication systems are designed to transfer power and data in spite of the failure of conventional wired systems. The physical protection system provides operable environments to the instruments. Therefore, the proposed system can be considered as a candidate of practical and urgent remedy for NPPs under the severe accident. After the Fukushima nuclear accident, survivability of essential instruments has been emphasized for immediate and accurate response. The essential instruments can measure environment conditions such as temperature, pressure, radioactivity and corium behavior inside nuclear power plants (NPPs) under a severe accident. Access to the inside of NPPs is restricted to human beings because of hazardous environment such as high radioactivity, high temperature and high pressure. Thus, monitoring the inside of NPPs is necessary for avoiding damage from the severe accident. Even though there were a number of instruments in Fukushima Daiichi NPP, they failed to obtain exact monitoring information. According to the details of the Fukushima nuclear accident, following problems can be counted as strong candidates of this instruments

  15. Highly Reliable Power and Communication System for Essential Instruments under a Severe Accident of NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, S. J.; Choi, B. H.; Jung, S. Y.; Rim, Chun T. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    In this paper, three survivable strategies to overcome the problems listed above are proposed for the essential instruments under the severe accident of NPPs. First, wire/wireless multi power systems are adopted to the essential instruments for continuous power supply. Second, wire/wireless communication systems are proposed for reliable transmission of measuring information among instruments and operators. Third, a physical protection system such as a harness and a heat isolation box is introduced to ensure operable conditions for the proposed systems. In this paper, a highly reliable strategy, which consists of wire/wireless multi power and communication systems and physical protection system is proposed to ensure the survival of the essential instruments under harsh external conditions. The wire/wireless multi power and communication systems are designed to transfer power and data in spite of the failure of conventional wired systems. The physical protection system provides operable environments to the instruments. Therefore, the proposed system can be considered as a candidate of practical and urgent remedy for NPPs under the severe accident. After the Fukushima nuclear accident, survivability of essential instruments has been emphasized for immediate and accurate response. The essential instruments can measure environment conditions such as temperature, pressure, radioactivity and corium behavior inside nuclear power plants (NPPs) under a severe accident. Access to the inside of NPPs is restricted to human beings because of hazardous environment such as high radioactivity, high temperature and high pressure. Thus, monitoring the inside of NPPs is necessary for avoiding damage from the severe accident. Even though there were a number of instruments in Fukushima Daiichi NPP, they failed to obtain exact monitoring information. According to the details of the Fukushima nuclear accident, following problems can be counted as strong candidates of this instruments

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  17. Coolant leakage detecting device

    International Nuclear Information System (INIS)

    Yamauchi, Kiyoshi; Kawai, Katsunori; Ishihara, Yoshinao.

    1995-01-01

    The device of the present invention judges an amount of leakage of primary coolants of a PWR power plant at high speed. Namely, a mass of coolants contained in a pressurizer, a volume controlling tank and loop regions is obtained based on a preset relational formula and signals of each of process amount, summed up to determine the total mass of coolants for every period of time. The amount of leakage for every period of time is calculated by a formula of Karman's filter based on the total mass of the primary coolants for every predetermined period of time, and displays it on CRT. The Karman's filter is formed on every formula for several kinds of states formed based on the preset amount of the leakage, to calculate forecasting values for every mass of coolants. An adaptable probability for every preset leakage amount is determined based on the difference between the forecast value and the observed value and the scattering thereof. The adaptable probability is compared with a predetermined threshold value, which is displayed on the CRT. This device enables earlier detection of leakage and identification of minute leakage amount as compared with the prior device. (I.S.)

  18. 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.)

  19. Normal accidents

    International Nuclear Information System (INIS)

    Perrow, C.

    1989-01-01

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

  20. A systemic analysis of South Korea Sewol ferry accident - Striking a balance between learning and accountability.

    Science.gov (United States)

    Kee, Dohyung; Jun, Gyuchan Thomas; Waterson, Patrick; Haslam, Roger

    2017-03-01

    The South Korea Sewol ferry accident in April 2014 claimed the lives of over 300 passengers and led to criminal charges of 399 personnel concerned including imprisonment of 154 of them as of Oct 2014. Blame and punishment culture can be prevalent in a more hierarchical society like South Korea as shown in the aftermath of this disaster. This study aims to analyse the South Korea ferry accident using Rasmussen's risk management framework and the associated AcciMap technique and to propose recommendations drawn from an AcciMap-based focus group with systems safety experts. The data for the accident analysis were collected mainly from an interim investigation report by the Board of Audit and Inspection of Korea and major South Korean and foreign newspapers. The analysis showed that the accident was attributed to many contributing factors arising from front-line operators, management, regulators and government. It also showed how the multiple factors including economic, social and political pressures and individual workload contributed to the accident and how they affected each other. This AcciMap was presented to 27 safety researchers and experts at 'the legacy of Jens Rasmussen' symposium adjunct to ODAM2014. Their recommendations were captured through a focus group. The four main recommendations include forgive (no blame and punishment on individuals), analyse (socio-technical system-based), learn (from why things do not go wrong) and change (bottom-up safety culture and safety system management). The findings offer important insights into how this type of accident should be understood, analysed and the subsequent response. Copyright © 2016 Elsevier Ltd. All rights reserved.

  1. Analysis of accidental loss of pool coolant due to leakage in a PWR SFP

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-01-01

    Highlights: • Accidental loss of pool coolant due to leakage in a PWR SFP was studied using MAAP5. • The effect of emergency ventilation on the accident progression was investigated. • The effect of emergency injection on the accident progression was discussed. - Abstract: A large loss of pool coolant/water accident may be caused by extreme accidents such as the pool wall or bottom floor punctures due to a large aircraft strike. The safety of SFP under this circumstance is very important. Large amounts of radioactive materials would be easily released into the environment if a severe accident happened in the SFP, because the spent fuel pool (SFP) in a PWR nuclear power station (NPS) is often located in the fuel handing building outside the reactor containment. To gain insight into the loss of pool coolant accident progression for a pressurized water reactor (PWR) SFP, a computational model was established by using the Modular Accident Analysis Program (MAAP5). Important factors such as Zr oxidation by air, air natural circulation and thermal radiation were considered for partial and complete drainage accidents without mitigation measures. The calculation indicated that even if the residual water level was in the active fuel region, there was a chance to effectively remove the decay heat through axial heat conduction (if the pool cooling system failed) or steam cooling (if the pool cooling system was working). For sensitivity study, the effects of emergency ventilation and water injection on the accident progression were analyzed. The analysis showed that for the current configuration of high-density storage racks, it was difficult to cool the spent fuels by air natural circulation. Enlarging the space between the adjacent assemblies was a way of increasing air natural circulation flow rate and maintaining the coolability of SFP. Water injection to the bottom of the SFP helped to recover water inventory, quenching the high temperature assemblies to prevent

  2. Reduction of PWR containment pressure after hypothetical accidents by water-cooling of the outer containment surface - annular space spray system

    International Nuclear Information System (INIS)

    Cremer, J.; Dietrich, D.P.; Roedder, P.

    1980-12-01

    The consequences of a core melt-out accident in the vicinity of a nuclear power station are determined by the integrity of the safety containment. This can be adversely affected by different events during the course of the core melt-out accident. The most important phenomenon is the contact between the melt and sump water. Due to the evaporation of the sump water, there is a continuous rise in pressure of the safety containment, which finally leads to failure due to excess pressure. In order to reduce the fission product release due to the resulting leakage, one must try to reduce the pressure as quickly as possible. As heat cannot be removed from the steel containment to the environment because of the thick concrete containment, it is best to bypass the insulating effect of the concrete by cooling the steel containment from outside. The aim of this investigation is therefore to work out a technically relatively simple system, which offers the possibility of backfitting, setting to work and repair. Such a system is an annular space spray system, by which the annular space between the concrete and steel containment has water pumped to the level of the dome and evenly sprayed over the top hemisphere. Mobile pumps on fire engines belonging to the fire brigade are sufficient to supply the cooling water and these will be available some hours after the accident occurs. The used spray water without any radioactive components is collected outside the reactor building and/or drained off. (orig./GL) [de

  3. analysis of reactivity accidents in MTR for various protection system parameters and core condition

    International Nuclear Information System (INIS)

    Mohamed, F.M.

    2011-01-01

    Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99 Mo production. The old core comprising 29 fuel elements and one Co Irradiation Device (CID) and the new core comprising 27 fuel elements, CID, and two 99 Mo production boxes. The in core irradiation has the advantage of no special cooling or irradiation loop is required. The purpose of the present work is the analysis of reactivity accidents (RIA) for ETRR-2 cores. The analysis was done to evaluate the accidents from different point of view:1- Analysis of the new core for various Reactor Protection System (RPS) parameters 2- Comparison between the two cores. 3- Analysis of the 99 Mo production boxes.PARET computer code was employed to compute various parameters. Initiating events in RIA involve various modes of reactivity insertion, namely, prompt critical condition (p=1$), accidental ejection of partial and complete CID uncontrolled withdrawal of a control rod accident, and sudden cooling of the reactor core. The time histories of reactor power, energy released, and the maximum fuel, clad and coolant temperatures of fuel elements and LEU plates were calculated for each of these accidents. The results show that the maximum clad temperatures remain well below the clad melting of both fuel and uranium plates during these accidents. It is concluded that for the new core, the RIA with scram will not result in fuel or uranium plate failure.

  4. 1983 international intercomparison of nuclear accident dosimetry systems at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Swaja, R.E.; Greene, R.T.; Sims, C.S.

    1985-04-01

    An international intercomparison of nuclear accident dosimetry systems was conducted during September 12-16, 1983, at Oak Ridge National Laboratory (ORNL) using the Health Physics Research Reactor operated in the pulse mode to simulate criticality accidents. This study marked the twentieth in a series of annual accident dosimetry intercomparisons conducted at ORNL. Participants from ten organizations attended this intercomparison and measured neutron and gamma doses at area monitoring stations and on phantoms for three different shield conditions. Results of this study indicate that foil activation techniques are the most popular and accurate method of determining accident-level neutron doses at area monitoring stations. For personnel monitoring, foil activation, blood sodium activation, and thermoluminescent (TL) methods are all capable of providing accurate dose estimates in a variety of radiation fields. All participants in this study used TLD's to determine gamma doses with very good results on the average. Chemical dosemeters were also shown to be capable of yielding accurate estimates of total neutron plus gamma doses in a variety of radiation fields. While 83% of all neutron measurements satisfied regulatory standards relative to reference values, only 39% of all gamma results satisfied corresponding guidelines for gamma measurements. These results indicate that continued improvement in accident dosimetry evaluation and measurement techniques is needed

  5. Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.

  6. Hydrogen-control systems for severe LWR accident conditions - a state-of-technology report

    International Nuclear Information System (INIS)

    Hilliard, R.K.; Postma, A.K.; Jeppson, D.W.

    1983-03-01

    This report reviews the current state of technology regarding hydrogen safety issues in light water reactor plants. Topics considered in this report relate to control systems and include combustion prevention, controlled combustion, minimization of combustion effects, combination of control concepts, and post-accident disposal. A companion report addresses hydrogen generation, distribution, and combustion. The objectives of the study were to identify the key safety issues related to hydrogen produced under severe accident conditions, to describe the state of technology for each issue, and to point out ongoing programs aimed at resolving the open issues

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

    International Nuclear Information System (INIS)

    Seo, K. S.; Lee, J. C.; Bang, K. S.; Choi, W. S.; Lee, S. H.; Seo, J. S.; Kim, K. Y.; Jeon, J. E.

    2011-06-01

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

  8. Fuel relocation modeling in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Tentner, A.M.; Miles, K.J.

    1985-01-01

    SAS4A is a new code system which has been designed for analyzing the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modeling the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel motion experiment analyses are also presented

  9. Some aspects of thyroid system status in persons exposed to the Chernobyl accident

    International Nuclear Information System (INIS)

    Cheban, A.K.; Afanasyev, D.E.; Boyarskaya, O.Y.

    1997-01-01

    The thyroid system status estimation held in post-accidental period dynamics among 7868 children evacuated from the 30-km Chernobyl zone and resident now in Slavutich city (Cs-137 contaminated area), among contaminated regions permanent residents, among native kievites and evacuated from 30-km zone. The thyroid pathology incidence dependence on residence place during Chernobyl Accident and after that was revealed. The immune-inflammatory thyroid disorders are characteristic for 30-km zone migrants, goitre different forms - for the radionuclides contaminated territories residents. No thyroid function abnormalities frequency confidential increase was registered during the research activities run. The total serum cholesterol level application unavailability is revealed in Chernobyl accident survivors thyroid hormones metabolic effects estimation. Data concerning Chernobyl accident consequences cleaning up participants (CACCP) presented additionally. (author)

  10. Fuel relocation modeling in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Tentner, A.M.; Miles, K.J.; Kalimullah; Hill, D.J.

    1986-01-01

    The SAS4A code system has been designed for the analysis of the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modelong the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel relocation are emphasized. 12 refs

  11. Modernization of the accident localisation system and relevant dose exposure on unit four of KNPP

    International Nuclear Information System (INIS)

    Valtchev, G.; Neshkova, M.; Nikilov, A.

    2005-01-01

    In 2001 a modernization of the accident localisation system (ALS) on Unit 4 was accomplished. The outage duration was longer then usually and special dose budget was elaborated. All ALS work was performed by external organisation. An ALARA implementation was recognised priority. The really accumulated collective doses were analysed and conclusions drawn. A short film on CD was prepared. (authors)

  12. [Injuries caused by traffic accidents: passive safety and restraint systems in automobiles].

    Science.gov (United States)

    Zuppichini, F; Orlandi, E; Genna, M; Rodella, L; Ricci, G; Arienzo, A; Dorrucci, V; Inaspettato, G

    1986-10-01

    In this article are considered the multiple instruments today employed in cars, in order to prevent or ameliorate the lesions caused to the occupants in case of road accident. The acquisitions in the differentiated structure of the car, in the windshield, in the components of the passenger cell are described, and the peculiar importance of the restraint systems is evidenced.

  13. Pathology of the reproductive system and thyroid of women liquidators of Chernobyl accident

    International Nuclear Information System (INIS)

    Babkin, A.A.; Merkulova, I.P.

    2014-01-01

    Data of the annual health follow up of the 100 women-liquidators of the Chernobyl accident performed by Republic centre of medical rehabilitation and balneotherapy have been analyzed. The high frequency of thyroid disease as well as the reproductive system pathology revealed: they were detected in 96% and 87% patients correspondingly. Oncological diseases were detected in 25% of studied cohort. (authors)

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

    International Nuclear Information System (INIS)

    Kuznecov, M.V.

    1992-01-01

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

  15. Full scale simulations of accidents on spent-nuclear-fuel shipping systems

    International Nuclear Information System (INIS)

    Yoshimura, H.R.

    1978-01-01

    In 1977 and 1978, five first-of-a-kind full scale tests of spent-nuclear-fuel shipping systems were conducted at Sandia Laboratories. The objectives of this broad test program were (1) to assess and demonstrate the validity of current analytical and scale modeling techniques for predicting damage in accident conditions by comparing predicted results with actual test results, and (2) to gain quantitative knowledge of extreme accident environments by assessing the response of full scale hardware under actual test conditions. The tests were not intended to validate the present regulatory standards. The spent fuel cask tests fell into the following configurations: crashes of a truck-transport system into a massive concrete barrier (100 and 130 km/h); a grade crossing impact test (130 km/h) involving a locomotive and a stalled tractor-trailer; and a railcar shipping system impact into a massive concrete barrier (130 km/h) followed by fire. In addition to collecting much data on the response of cask transport systems, the program has demonstrated thus far that current analytical and scale modeling techniques are valid approaches for predicting vehicular and cask damage in accident environments. The tests have also shown that the spent casks tested are extremely rugged devices capable of retaining their radioactive contents in very severe accidents

  16. Application research of cloud computing in emergency system platform of nuclear accidents

    International Nuclear Information System (INIS)

    Zhang Yan; Yue Huiguo; Lin Quanyi; Yue Feng

    2013-01-01

    This paper described the key technology of the concept of cloud computing, service type and implementation methods. Combined with the upgrade demand of nuclear accident emergency system platform, the paper also proposed the application design of private cloud computing platform, analyzed safety of cloud platform and the characteristics of cloud disaster recovery. (authors)

  17. Phenomena occurring in the reactor coolant system during severe core damage accidents

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1989-01-01

    The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. In the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. In contrast to normal operating conditions, severe core damage accidents are characterized by significant temporal and spatial variations in heat and mass fluxes, and by eventual geometrical changes within the RCS. Furthermore, the difficulties in describing the system in the severe accident mode are compounded by the occurrence of chemical reactions. These reactions can influence both the thermal and the mass transport behavior of the system. In addition, behavior of the reactor vessel internals and of materials released from the core region (especially the radioactive fission products) in the course of the accident likewise become of concern to the analyst. This report addresses these concerns. 9 refs., 1 tab

  18. Environmental decision support system on base of geoinformational technologies for the analysis of nuclear accident consequences

    International Nuclear Information System (INIS)

    Haas, T.C.; Maigan, M.; Arutyunyan, R.V.; Bolshov, L.A.; Demianov, V.V.

    1996-01-01

    The report deals with description of the concept and prototype of environmental decision support system (EDSS) for the analysis of late off-site consequences of severe nuclear accidents and analysis, processing and presentation of spatially distributed radioecological data. General description of the available software, use of modem achievements of geostatistics and stochastic simulations for the analysis of spatial data are presented and discussed

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  20. Comparison of accident risks in different energy systems: Comments from Russian specialists

    International Nuclear Information System (INIS)

    2000-01-01

    Many articles on accident risk analysis of different energy systems in comparison with nuclear power share certain stereotypical features. For example: When assessing the risks associated with the operation of such facilities, they ignore the effects of the upgrading of RBMK reactors which was carried out after the Chernobyl accident. In their integrated assessment of the radiological consequences of the Chernobyl accident they use numerous studies which frequently contain unreliable source data and unfounded predictions, and they ignore many socio-political factors which considerably increased the damage caused by the accident. Unfortunately, the study in question, despite its topicality and originality of approach, is also not without such shortcomings. After the Chernobyl accident, reconstruction and safety enhancement measures were implemented at nuclear power plants with RBMK reactors which were without precedent in world practice and have continued to this day. According to probabilistic safety assessments (PSA) carried out with the assistance of international experts, the probability of serious accidents at RBMKs has decreased by a factor of two or more thanks to the above mentioned measures. The mean weighted safety index for all operational RBMK reactors is 10 -4 l/year and is decreasing thanks to the ongoing and planned reconstruction of all units. All operational nuclear power plants with RBMK reactors are thus on a par with the successfully operating Soviet WWERs and western boiling water reactors (BWRs) and pressurized water reactors (PWRs), and satisfy the IAEA recommendations regarding the risk level of older generation nuclear power plants. The authors of the IAEA Bulletin article give estimates of the remote radiological consequences of the Chernobyl accident which range from an estimated 10,000 to 30,000 fatal cases of radiation-induced cancer, and the literature on the subject contains even more extreme estimates. However, our 14 years

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

    Science.gov (United States)

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

    2018-02-01

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

  2. Design and implementation of an identification system in construction site safety for proactive accident prevention.

    Science.gov (United States)

    Yang, Huanjia; Chew, David A S; Wu, Weiwei; Zhou, Zhipeng; Li, Qiming

    2012-09-01

    Identifying accident precursors using real-time identity information has great potential to improve safety performance in construction industry, which is still suffering from day to day records of accident fatality and injury. Based on the requirements analysis for identifying precursor and the discussion of enabling technology solutions for acquiring and sharing real-time automatic identification information on construction site, this paper proposes an identification system design for proactive accident prevention to improve construction site safety. Firstly, a case study is conducted to analyze the automatic identification requirements for identifying accident precursors in construction site. Results show that it mainly consists of three aspects, namely access control, training and inspection information and operation authority. The system is then designed to fulfill these requirements based on ZigBee enabled wireless sensor network (WSN), radio frequency identification (RFID) technology and an integrated ZigBee RFID sensor network structure. At the same time, an information database is also designed and implemented, which includes 15 tables, 54 queries and several reports and forms. In the end, a demonstration system based on the proposed system design is developed as a proof of concept prototype. The contributions of this study include the requirement analysis and technical design of a real-time identity information tracking solution for proactive accident prevention on construction sites. The technical solution proposed in this paper has a significant importance in improving safety performance on construction sites. Moreover, this study can serve as a reference design for future system integrations where more functions, such as environment monitoring and location tracking, can be added. Copyright © 2011 Elsevier Ltd. All rights reserved.

  3. An expert system for the quantification of fault rates in construction fall accidents.

    Science.gov (United States)

    Talat Birgonul, M; Dikmen, Irem; Budayan, Cenk; Demirel, Tuncay

    2016-01-01

    Expert witness reports, prepared with the aim of quantifying fault rates among parties, play an important role in a court's final decision. However, conflicting fault rates assigned by different expert witness boards lead to iterative objections raised by the related parties. This unfavorable situation mainly originates due to the subjectivity of expert judgments and unavailability of objective information about the causes of accidents. As a solution to this shortcoming, an expert system based on a rule-based system was developed for the quantification of fault rates in construction fall accidents. The aim of developing DsSafe is decreasing the subjectivity inherent in expert witness reports. Eighty-four inspection reports prepared by the official and authorized inspectors were examined and root causes of construction fall accidents in Turkey were identified. Using this information, an evaluation form was designed and submitted to the experts. Experts were asked to evaluate the importance level of the factors that govern fall accidents and determine the fault rates under different scenarios. Based on expert judgments, a rule-based expert system was developed. The accuracy and reliability of DsSafe were tested with real data as obtained from finalized court cases. DsSafe gives satisfactory results.

  4. Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

    International Nuclear Information System (INIS)

    Su Xia

    2013-01-01

    The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

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

    International Nuclear Information System (INIS)

    Sdouz, Gert

    2006-01-01

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

  6. The intelligent system for accident identification in nuclear power plant

    International Nuclear Information System (INIS)

    Hernandez, Jorge Luis.

    1998-01-01

    Accidental situations in NPP are of greet concern for operators, the facility, regulatory bodies and the environment. This work proposes a design of intelligent system aimed to assist the operator in the process of decision making when initiator events with higher relative contribution to the reactor core damage occur. The intelligent System uses the results of the pre operational Probabilistic Safety Assessment and the Thermal hydraulic Safety Analyses of the NPP Juragua as source for building its knowledge base. The nucleus of the system is presented as a design of an intelligent hybrid system from the combination of the artificial intelligence techniques: fussy logic and artificial neural networks. The system works with variables from the process of the firsts circuit, second circuit and the containment and it is presented as a model for the integration of safety analyses in the process of decision making by the operator when tackling with accidental situations

  7. The intelligent system for accident identification in NPP

    International Nuclear Information System (INIS)

    Hernandez, Jorge Luis.

    1998-01-01

    Accidental situations in NPP are of greet concern for operators, the facility, regulatory bodies and the environment. This work proposes a design of intelligent system aimed to assist the operator in the process of decision making when initiator events with higher relative contribution to the reactor core damage occur. The intelligent System uses the results of the pre operational Probabilistic Safety Assessment and the Thermal hydraulic Safety Analyses of the NPP Juragua as source for building its knowledge base. The nucleus of the system is presented as a design of an intelligent hybrid system from the combination of the artificial intelligence techniques: fussy logic and artificial neural networks. The system works with variables from the process of the firsts circuit, second circuit and the containment and it is presented as a model for the integration of safety analyses in the process of decision making by the operator when tackling with accidental situations

  8. Coolant leakage detection device

    International Nuclear Information System (INIS)

    Ito, Takao.

    1983-01-01

    Purpose: To surely detect the coolant leakage at a time when the leakage amount is still low in the intra-reactor inlet pipeway of FBR type reactor. Constitution: Outside of the intra-reactor inlet piping for introducing coolants at low temperature into a reactor core, an outer closure pipe is furnished. The upper end of the outer closure pipe opens above the liquid level of the coolants in the reactor, and a thermocouple is inserted to the opening of the upper end. In such a structure, if the coolants in the in-reactor piping should leak to the outer closure pipe, coolants over-flows from the opening thereof, at which the thermocouple detects the temperature of the coolants at a low temperature, thereby enabling to detect the leakage of the coolants at a time when it is still low. (Kamimura, M.)

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

    International Nuclear Information System (INIS)

    Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi

    1997-01-01

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

  10. Accident diagnosis system based on real-time decision tree expert system

    Science.gov (United States)

    Nicolau, Andressa dos S.; Augusto, João P. da S. C.; Schirru, Roberto

    2017-06-01

    Safety is one of the most studied topics when referring to power stations. For that reason, sensors and alarms develop an important role in environmental and human protection. When abnormal event happens, it triggers a chain of alarms that must be, somehow, checked by the control room operators. In this case, diagnosis support system can help operators to accurately identify the possible root-cause of the problem in short time. In this article, we present a computational model of a generic diagnose support system based on artificial intelligence, that was applied on the dataset of two real power stations: Angra1 Nuclear Power Plant and Santo Antônio Hydroelectric Plant. The proposed system processes all the information logged in the sequence of events before a shutdown signal using the expert's knowledge inputted into an expert system indicating the chain of events, from the shutdown signal to its root-cause. The results of both applications showed that the support system is a potential tool to help the control room operators identify abnormal events, as accidents and consequently increase the safety.

  11. Monitoring of oil leakage from a ship propulsion system using IR camera and wavelet analysis for prevention of health and ecology risks and engine faults

    Energy Technology Data Exchange (ETDEWEB)

    Soda, J.; Beros, S. [University of Split (Croatia). Faculty of Electrical Engineering, Mechanical Engineering and Naval Architecture; Antonic, R.; Vujovic, I. [University of Split (Croatia) Maritime Faculty; Kuzmanic, I.

    2009-03-15

    It is a well known fact that oil leakage from ship diesel engines is harmful both for the environment and the ship engine and therefore has to be observed and alarmed. The present paper proposes a system for overcoming described problems by installing a computer vision system. The used algorithm of pattern recognition system is based on the use of wavelet structures. Additionally, one of the problems in the system is the compensation of camera movements due to engine vibration. The compensation part of the computer vision solution is used to improve position determination. The position determination is improved more that 300 % when using farras wavelets. (Abstract Copyright [2009], Wiley Periodicals, Inc.) [German] (Abstract Copyright [2009], Wiley Periodicals, Inc.)

  12. ABOUT THE SPECIAL INVESTIGATIONS OF THE PROTECTION OF THE TECHNICAL SECURITY SYSTEMS AGAINST INFORMATION LEAKAGE DUE TO THE ACOUSTO-ELECTRICAL TRANSFORMATIONS

    Directory of Open Access Journals (Sweden)

    A. P. Durakovskiy

    2016-12-01

    Full Text Available None of the critically important facilities can operate without the engineered safety system. Functionally varied security networks or a fire alarm system can refer to this system as well as safety and reliability which are provided by secured energy, water and heating supply. In the process of attestation according to the requirements of information security of information objects with such technical means, it is necessary to conduct special investigations of protection against leakage of acoustic speech information through the channels of the acousto-electrical transformations (AET. There are major aspects in the data leak via AET, which currently include the following: lack of and /or obtaining legal and safety norms to regulate specified parameters; lack of the automated hardware and software system for some AET variations to carry out measurements; lack of specified safety equipment for some AET variations; lack of shelter security units; high costs of AET measurement and control units; and low measurement repeatability.

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

    International Nuclear Information System (INIS)

    Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

    2016-01-01

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

  14. Experience with neutron flux monitoring systems qualified for post-accident monitoring

    International Nuclear Information System (INIS)

    Shugars, H.G.; Miller, J.F.

    1995-01-01

    In this paper we discuss the environmental requirements for excore neutron flux monitors that are qualified for use during and after postulated accidents in Pressurized Water Reactors (PWRs). We emphasize PWRs designed in the United States, which are similar to those used also in parts of Western Europe and Eastern Asia. We then discuss design features of the flux monitoring systems necessary to address the environmental, functional, and regulatory requirements, and the experience with these systems. (author). 9 refs, 2 figs

  15. Channel follower leakage restrictor

    International Nuclear Information System (INIS)

    Williamson, H.E.; Smith, B.A.

    1977-01-01

    An improved means is provided to control coolant leakage between the flow channel and the lower tie plate of a nuclear fuel assembly. The means includes an opening in the lower tie plate and a movable element adjacent thereto. The coolant pressure within the tie plate biases the movable means toward the inner surface of the surrounding flow channel to compensate for any movement of the flow channel away from the lower tie plate to thereby control the leakage of coolant flow from the fuel assemblies to the spaces among the fuel assemblies of the core. 9 figures

  16. Technology evaluation for space station atmospheric leakage

    Energy Technology Data Exchange (ETDEWEB)

    Lemon, D.K.; Friesel, M.A.; Griffin, J.W.; Skorpik, J.R.; Shepard, C.L.; Antoniak, Z.I.; Kurtz, R.J.

    1990-02-01

    A concern in operation of a space station is leakage of atmosphere through seal points and through the walls as a result of damage from particle (space debris and micrometeoroid) impacts. This report describes a concept for a monitoring system to detect atmosphere leakage and locate the leak point. The concept is based on analysis and testing of two basic methods selected from an initial technology survey of potential approaches. 18 refs., 58 figs., 5 tabs.

  17. Postulated accident conditions for air cleaning systems and radiological dose assessments for containment options

    International Nuclear Information System (INIS)

    Hilliard, R.K.; Postma, A.K.

    1975-01-01

    Ambient conditions and performance requirements for emergency air cleaning systems applicable to commercial LMFBR plants were studied. The focus of this study centered on aerosol removal under hypothetical core disruptive accident conditions. Effort completed includes a review of air cleaning systems related to LMFBR plants, selection of three reference containment system designs, postulation of the EACS design basis accident (EACS-DBA), analysis of thermal conditions resulting from the DBA, analysis of aerosol transport behavior following the DBA, and an estimate of bone dose at the site boundary for each of the reference plant designs. Reference plant concepts were a single containment system (e.g., FFTF), a double containment system (e.g., CRBRP with closed head compartment), and a containment-confinement design in which an inerted, sealed primary volume was located within a ventilated building whose exhaust was filtered. The reference design basis accident selected here involved release to the inner containment system of 1 percent of non-volatile solids and plutonium, 25 percent of core halogens, 25 percent of core volatile solids, 100 percent of core noble gases, 68 lbs of sodium vapor and 5000 lbs of liquid sodium. 13 references. (U.S.)

  18. System for step-wise accident protection of nuclear reactors

    International Nuclear Information System (INIS)

    Rubek, J.; Kuklik, B.; Bednarik, K.

    1991-01-01

    A system comprising electric switching circuits is proposed for the control of a WWER type reactor shutdown in case of turbine failure or another abnormal situation. The fastest reactor shutdown mode is only resorted to if the pressures in the primary and secondary circuits would otherwise increase above tolerable limits and safety valves would open. The temperature and pressure stress of the nuclear power plant components and fuel is reduced. In this manner, the losses emerging during turbine failures due to false alarms are minimized. The contacts of the system switch if the turbines are relieved to the power of the unit home consumption, if the first or second turbine fails by closing the quick-acting valves, if a signal for blocking the by-pass stations of the operated turbines appears, or if the electric supply of the control system and of the turbo-set protection fails. (M.D.). 1 fig

  19. Development of accident event trees and evaluation of safety system failure modes for the nuclear ultra large crude carrier

    International Nuclear Information System (INIS)

    Lewe, C.K.; Coffey, R.S.; Goodwin, E.F.; Maltese, J.G.; Pyatt, D.W.

    1978-01-01

    A method of applying the probabilistic accident event tree methodology to safety assessments of a nuclear powered Ultra Large Crude Carrier is presented. Also presented are the procedures by which an external accident initiating event, such as a ship collision, may be correlated with the probabilities of damage to the ship's safety systems and to their ultimate availabilities to perform required safety functions

  20. Pre-test analysis of ATLAS SBO with RCP seal leakage scenario using MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Quang Huy; Lee, Sang Young; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    This study presents a pre-test calculation for the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) SBO experiment with RCP seal leakage scenario. Initially, turbine-driven auxfeed water pumps are used. Then, outside cooling water injection method is used for long term cooling. The analysis results would be useful for conducting the experiment to verify the APR 1400 extended SBO optimum mitigation strategy using outside cooling water injection in future. The pre-test calculation for ATLAS extended SBO with RCP seal leakage and outside cooling water injection scenario is performed. After Fukushima nuclear accident, the capability of coping with the extended station blackout (SBO) becomes important. Many NPPs are applying FLEX approach as main coping strategies for extended SBO scenarios. In FLEX strategies, outside cooling water injection to reactor cooling system (RCS) and steam generators (SGs) is considered as an effective method to remove residual heat and maintain the inventory of the systems during the accident. It is worthwhile to examine the soundness of outside cooling water injection method for extended SBO mitigation by both calculation and experimental demonstration. From the calculation results, outside cooling water injection into RCS and SGs is verified as an effective method during extended SBO when RCS and SGs depressurization is sufficiently performed.

  1. Reactivity Accidents in CAREM-25 Core with and Without Safety Systems Actuation

    International Nuclear Information System (INIS)

    Gimenez, Marcelo; Vertullo, Alicia; Schlamp, Miguel

    2000-01-01

    A reactivity accident in CAREM core can be provoked by different initiating events, a cold water injection in pressure vessel, a secondary side steam line breakage and a failure in the absorbing rods drive system.The present work analyses inadverted control rod withdraws transients.Maximum worth control rod (2.5 $) at normal velocity (1 cm/s) is adopted for the simulations (Reactivity ramp of 0.018 $/s).Different scenarios considering actuation of first shutdown system (FSS), second shutdown system (SSS) and selflimiting conditions were modeled.Results of the accident with actuation of FSS show that safety margins are well above critical values (DNBR and CPR).In the cases with failure of the FSS and success of SSS or selflimited, safety margins are below critical values, however, the SSS provides a reduction of elapsed time under advised margins

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  3. Geographic Information System (GIS) capabilities in traffic accident information management: a qualitative approach

    Science.gov (United States)

    Ahmadi, Maryam; Valinejadi, Ali; Goodarzi, Afshin; Safari, Ameneh; Hemmat, Morteza; Majdabadi, Hesamedin Askari; Mohammadi, Ali

    2017-01-01

    Background Traffic accidents are one of the more important national and international issues, and their consequences are important for the political, economical, and social level in a country. Management of traffic accident information requires information systems with analytical and accessibility capabilities to spatial and descriptive data. Objective The aim of this study was to determine the capabilities of a Geographic Information System (GIS) in management of traffic accident information. Methods This qualitative cross-sectional study was performed in 2016. In the first step, GIS capabilities were identified via literature retrieved from the Internet and based on the included criteria. Review of the literature was performed until data saturation was reached; a form was used to extract the capabilities. In the second step, study population were hospital managers, police, emergency, statisticians, and IT experts in trauma, emergency and police centers. Sampling was purposive. Data was collected using a questionnaire based on the first step data; validity and reliability were determined by content validity and Cronbach’s alpha of 75%. Data was analyzed using the decision Delphi technique. Results GIS capabilities were identified in ten categories and 64 sub-categories. Import and process of spatial and descriptive data and so, analysis of this data were the most important capabilities of GIS in traffic accident information management. Conclusion Storing and retrieving of descriptive and spatial data, providing statistical analysis in table, chart and zoning format, management of bad structure issues, determining the cost effectiveness of the decisions and prioritizing their implementation were the most important capabilities of GIS which can be efficient in the management of traffic accident information. PMID:28848627

  4. Geographic Information System (GIS) capabilities in traffic accident information management: a qualitative approach.

    Science.gov (United States)

    Ahmadi, Maryam; Valinejadi, Ali; Goodarzi, Afshin; Safari, Ameneh; Hemmat, Morteza; Majdabadi, Hesamedin Askari; Mohammadi, Ali

    2017-06-01

    Traffic accidents are one of the more important national and international issues, and their consequences are important for the political, economical, and social level in a country. Management of traffic accident information requires information systems with analytical and accessibility capabilities to spatial and descriptive data. The aim of this study was to determine the capabilities of a Geographic Information System (GIS) in management of traffic accident information. This qualitative cross-sectional study was performed in 2016. In the first step, GIS capabilities were identified via literature retrieved from the Internet and based on the included criteria. Review of the literature was performed until data saturation was reached; a form was used to extract the capabilities. In the second step, study population were hospital managers, police, emergency, statisticians, and IT experts in trauma, emergency and police centers. Sampling was purposive. Data was collected using a questionnaire based on the first step data; validity and reliability were determined by content validity and Cronbach's alpha of 75%. Data was analyzed using the decision Delphi technique. GIS capabilities were identified in ten categories and 64 sub-categories. Import and process of spatial and descriptive data and so, analysis of this data were the most important capabilities of GIS in traffic accident information management. Storing and retrieving of descriptive and spatial data, providing statistical analysis in table, chart and zoning format, management of bad structure issues, determining the cost effectiveness of the decisions and prioritizing their implementation were the most important capabilities of GIS which can be efficient in the management of traffic accident information.

  5. ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    WILLIAMS, J.C.

    2003-11-15

    This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.

  6. Methane leakage during the evolution of petroleum systems in the Western Canada Sedimentary Basin and the Central Graben area of the North Sea

    Science.gov (United States)

    Berbesi, L. A.; di Primio, R.; Anka, Z.; Horsfield, B.

    2012-04-01

    Around 500 to 600 Tg (1 Tg = 1012 g) of methane enter the atmosphere every year, mainly as product of microbial processes and combustion of fossil fuels and burning biomass. The importance of another source, the geologic emissions of methane, is up to now only loosely constrained. In this study, we addressed the potential methane emissions during the geological evolution of the Western Canada sedimentary basin (WCSB), which holds the largest oil sand accumulations in the world, and the Central Graben area of the North Sea. In the case of the WCSB, thermogenic gas generation and leakage at the sediment surface were addressed through 3D petroleum systems modeling. In this basin, the accumulated oil experienced intense biodegradation that resulted in large masses of biogenic methane. We quantified this latter mass though a two-step mass balance approach. Firstly, we estimated the rate of petroleum degradation and the magnitude of petroleum loss. After this, we calculated the mass of biogenic methane generated using a model that assumes hexadecane (C16H34) as representative of the saturated compounds (Zengler et al., 1999). Our 3D model suggests that 90000-150000 Tg of dry gas (mostly methane) could have leaked during the interval from 80 Ma to 60 Ma. Therefore, uniform leakage rates would have been in the order of 10-3-10-2 Tg yr-1. Biogenic methane generation could have taken place at rates of 10-4 to 10-2 Tg yr-1. However, the effective mass of thermogenic and biogenic methane reaching the atmosphere might have been up to 90% lower than calculated here due to methanotrophic consumption in soils (Etiope and Klusman, 2002). We addressed the thermogenic gas generation and leakage in the Central Graben through two different methods. The first is based on a previous 3D petroleum system modeling of the region (Neumann, 2006). The second consisted of calculating the mass of generated petroleum based on source rock extension and properties (Schmoker, 1994), and then

  7. Performance-based containment leakage testing

    International Nuclear Information System (INIS)

    Cybulskis, P.

    1995-01-01

    The U.S. Nuclear Regulatory Commission (NRC) is reviewing regulatory requirements in an effort to revise those that are marginal to safety but impose significant burdens on licensees. Identification of requirements marginal to safety and development and evaluation of alternatives utilize the NRC safety goals and insights from probabilistic risk assessments (PRAs). Since earlier studies found design-basis containment leakage to be a minor contributor to reactor accident risk, containment leakage testing has been selected as a candidate for change in regulations. This paper summarizes the technical analyses supporting the NRC proposal to amend Appendix J of 10 CFR Part 50 as its first effort to decrease unnecessary regulatory burdens on licensees

  8. Highlights from the literature on accident causation and system safety: Review of major ideas, recent contributions, and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Saleh, J.H., E-mail: jsaleh@gatech.ed [School of Aerospace Engineering, Georgia Institute of Technology (United States); Marais, K.B. [School of Aeronautics and Astronautics, Purdue University (United States); Bakolas, E.; Cowlagi, R.V. [School of Aerospace Engineering, Georgia Institute of Technology (United States)

    2010-11-15

    This work constitutes a short guide to the extensive but fragmented literature on accident causation and system safety. After briefly motivating the interest in accident causation and discussing the notion of a safety value chain, we delve into our multi-disciplinary review with discussions of Man Made Disasters, Normal Accident, and the High Reliability Organizations (HRO) paradigm. The HRO literature intersects an extensive literature on safety culture, a subject we then briefly touch upon. Following this discussion, we note that while these social and organizational contributions have significantly enriched our understanding of accident causation and system safety, they have important deficiencies and are lacking in their understanding of technical and design drivers of system safety and accident causation. These missing ingredients, we argue, were provided in part by the development of Probabilistic Risk Assessment (PRA). The idea of anticipating possible accident scenarios, based on the system design and configuration, as well as its technical and operational characteristics, constitutes an important contribution of PRA, which builds on and extends earlier contributions made by the development of Fault Tree and Event Tree Analysis. We follow the discussion of PRA with an exposition of the concept of safety barriers and the principle of defense-in-depth, both of which emphasize the functions and 'safety elements [that should be] deliberately inserted' along potential accident trajectories to prevent, contain, or mitigate accidents. Finally, we discuss two ideas that are emerging as foundational in the literature on system safety and accident causation, namely that system safety is a 'control problem', and that it requires a 'system theoretic' approach to be dealt with. We clarify these characterizations and indicate research opportunities to be pursued along these directions. We conclude this work with two general recommendations

  9. Highlights from the literature on accident causation and system safety: Review of major ideas, recent contributions, and challenges

    International Nuclear Information System (INIS)

    Saleh, J.H.; Marais, K.B.; Bakolas, E.; Cowlagi, R.V.

    2010-01-01

    This work constitutes a short guide to the extensive but fragmented literature on accident causation and system safety. After briefly motivating the interest in accident causation and discussing the notion of a safety value chain, we delve into our multi-disciplinary review with discussions of Man Made Disasters, Normal Accident, and the High Reliability Organizations (HRO) paradigm. The HRO literature intersects an extensive literature on safety culture, a subject we then briefly touch upon. Following this discussion, we note that while these social and organizational contributions have significantly enriched our understanding of accident causation and system safety, they have important deficiencies and are lacking in their understanding of technical and design drivers of system safety and accident causation. These missing ingredients, we argue, were provided in part by the development of Probabilistic Risk Assessment (PRA). The idea of anticipating possible accident scenarios, based on the system design and configuration, as well as its technical and operational characteristics, constitutes an important contribution of PRA, which builds on and extends earlier contributions made by the development of Fault Tree and Event Tree Analysis. We follow the discussion of PRA with an exposition of the concept of safety barriers and the principle of defense-in-depth, both of which emphasize the functions and 'safety elements [that should be] deliberately inserted' along potential accident trajectories to prevent, contain, or mitigate accidents. Finally, we discuss two ideas that are emerging as foundational in the literature on system safety and accident causation, namely that system safety is a 'control problem', and that it requires a 'system theoretic' approach to be dealt with. We clarify these characterizations and indicate research opportunities to be pursued along these directions. We conclude this work with two general recommendations: (1) that more fundamental

  10. Assessing Agulhas leakage

    NARCIS (Netherlands)

    van Sebille, E.

    2009-01-01

    Agulhas leakage, the water that flows from the Indian Ocean to the Atlantic Ocean, plays an important role in the circulation of the Atlantic Ocean. The magnitude of this flux of warm and saline Indian Ocean water into the much colder and fresher Atlantic Ocean can be related to the strength of the

  11. Integrated Three-Voltage-Booster DC-DC Converter to Achieve High Voltage Gain with Leakage-Energy Recycling for PV or Fuel-Cell Power Systems

    Directory of Open Access Journals (Sweden)

    Chih-Lung Shen

    2015-09-01

    Full Text Available In this paper, an integrated three-voltage-booster DC-DC (direct current to direct current converter is proposed to achieve high voltage gain for renewable-energy generation systems. The proposed converter integrates three voltage-boosters into one power stage, which is composed of an active switch, a coupled-inductor, five diodes, and five capacitors. As compared with conventional high step-up converters, it has a lower component count. In addition, the features of leakage-energy recycling and switching loss reduction can be accomplished for conversion efficiency improvement. While the active switch is turned off, the converter can inherently clamp the voltage across power switch and suppress voltage spikes. Moreover, the reverse-recovery currents of all diodes can be alleviated by leakage inductance. A 200 W prototype operating at 100 kHz switching frequency with 36 V input and 400 V output is implemented to verify the theoretical analysis and to demonstrate the feasibility of the proposed high step-up DC-DC converter.

  12. Signal attenuation due to cavity leakage

    International Nuclear Information System (INIS)

    Sherman, M.H.; Modera, M.P.

    1988-01-01

    The propagation of sound waves in fluids requires information about three properties of the system: capacitance (compressibility), resistance (friction), and inductance (inertia). Acoustical design techniques to date have tended to ignore the frictional effects associated with airflow across the envelope of the acoustic cavity (e.g., resistive vents). Since such leakage through the cavity envelope is best expressed with a power law dependence on the pressure, standard Fourier techniques that rely on linearity cannot be used. In this article, the theory relevant to nonlinear leakage is developed and equations presented. Potential applications of the theory to techniques for quantifying the leakage of buildings are presented. Experimental results from pressure decays in a full-scale test structure are presented and the leakage so measured is compared with independent measurements to demonstrate the technique

  13. CO2 leakage from carbon dioxide capture and storage (CCS) systems affects organic matter cycling in surface marine sediments.

    Science.gov (United States)

    Rastelli, Eugenio; Corinaldesi, Cinzia; Dell'Anno, Antonio; Amaro, Teresa; Greco, Silvestro; Lo Martire, Marco; Carugati, Laura; Queirós, Ana M; Widdicombe, Stephen; Danovaro, Roberto

    2016-12-01

    Carbon dioxide capture and storage (CCS), involving the injection of CO 2 into the sub-seabed, is being promoted worldwide as a feasible option for reducing the anthropogenic CO 2 emissions into the atmosphere. However, the effects on the marine ecosystems of potential CO 2 leakages originating from these storage sites have only recently received scientific attention, and little information is available on the possible impacts of the resulting CO 2 -enriched seawater plumes on the surrounding benthic ecosystem. In the present study, we conducted a 20-weeks mesocosm experiment exposing coastal sediments to CO 2 -enriched seawater (at 5000 or 20,000 ppm), to test the effects on the microbial enzymatic activities responsible for the decomposition and turnover of the sedimentary organic matter in surface sediments down to 15 cm depth. Our results indicate that the exposure to high-CO 2 concentrations reduced significantly the enzymatic activities in the top 5 cm of sediments, but had no effects on subsurface sediment horizons (from 5 to 15 cm depth). In the surface sediments, both 5000 and 20,000 ppm CO 2 treatments determined a progressive decrease over time in the protein degradation (up to 80%). Conversely, the degradation rates of carbohydrates and organic phosphorous remained unaltered in the first 2 weeks, but decreased significantly (up to 50%) in the longer term when exposed at 20,000 ppm of CO 2 . Such effects were associated with a significant change in the composition of the biopolymeric carbon (due to the accumulation of proteins over time in sediments exposed to high-pCO 2 treatments), and a significant decrease (∼20-50% at 5000 and 20,000 ppm respectively) in nitrogen regeneration. We conclude that in areas immediately surrounding an active and long-lasting leak of CO 2 from CCS reservoirs, organic matter cycling would be significantly impacted in the surface sediment layers. The evidence of negligible impacts on the deeper sediments should be

  14. Preliminary analysis of accident in SST-1 current feeder system

    International Nuclear Information System (INIS)

    Roy, Swati; Kanabar, Deven; Garg, Atul; Singh, Amit; Tanna, Vipul; Prasad, Upendra; Srinivasan, R.

    2017-01-01

    Steady-state Tokamak-1 (SST-1) has 16 superconducting Toroidal field (TF) and 9 superconducting poloidal field (PF) coils rated for 10kA DC. All the TF are connected in series and are operated in DC condition whereas PF coils are individually operated in pulse mode during SST-1 campaigns. SST-1 current feeder system (CFS) houses 9 pairs of PF current leads and 1 pair of TF current leads. During past SST-1 campaign, there were arcing incidents within SST-1 CFS chamber which caused significant damage to PF superconducting current leads as well as its Helium cooling lines of the current leads. This paper brings out the preliminary analysis of the mentioned arcing incident, possible reasons and its investigation thereby laying out the sequence of events. From this analysis and observations, various measures to avoid such arcing incidents have also been proposed. (author)

  15. Integral isolation valve systems for loss of coolant accident protection

    Science.gov (United States)

    Kanuch, David J.; DiFilipo, Paul P.

    2018-03-20

    A nuclear reactor includes a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel having vessel penetrations that exclusively carry flow into the nuclear reactor and at least one vessel penetration that carries flow out of the nuclear reactor. An integral isolation valve (IIV) system includes passive IIVs each comprising a check valve built into a forged flange and not including an actuator, and one or more active IIVs each comprising an active valve built into a forged flange and including an actuator. Each vessel penetration exclusively carrying flow into the nuclear reactor is protected by a passive IIV whose forged flange is directly connected to the vessel penetration. Each vessel penetration carrying flow out of the nuclear reactor is protected by an active IIV whose forged flange is directly connected to the vessel penetration. Each active valve may be a normally closed valve.

  16. [Nordic accident classification system used in the Danish National Hospital Registration System to register causes of severe traumatic brain injury].

    Science.gov (United States)

    Engberg, Aase Worsaa; Penninga, Elisabeth Irene; Teasdale, Thomas William

    2007-11-05

    The purpose was to illustrate the use of the accident classification system worked out by the Nordic Medico-Statistical Committee (NOMESCO). In particular, registration of causes of severe traumatic brain injury according to the system as part of the Danish National Hospital Registration System was studied. The study comprised 117 patients with very severe traumatic brain injury (TBI) admitted to the Brain Injury Unit of the University Hospital in Hvidovre, Copenhagen, from 1 October 2000 to 30 September 2002. Prospective NOMESCO coding at discharge was compared to independent retrospective coding based on hospital records, and to coding from other wards in the Danish National Hospital Registration System. Furthermore, sets of codes in the Danish National Hospital Registration System for consecutive admissions after a particular accident were compared. Identical results of prospective and independent retrospective coding were found for 65% of 588 single codes, and complete sets of codes for the same accident were identical only in 28% of cases. Sets of codes for the first admission in a hospital course corresponded to retrospective coding at the end of the course in only 17% of cases. Accident code sets from different wards, based on the same injury, were identical in only 7% of cases. Prospective coding by the NOMESCO accident classification system proved problematic, both with regard to correctness and completeness. The system--although logical--seems too complicated compared to the resources invested in the coding. The results of this investigation stress the need for better management and for better instruction to those who carry out the registration.

  17. University student's education after Fukushima nuclear leakage crisis

    International Nuclear Information System (INIS)

    Dou Daying; Gu Jianzhong; Zheng Jianying

    2012-01-01

    Fukushima nuclear leakage crisis after 3.11 earthquake and tsunami, the horrible INES-7 accident warns the colleagues all over the world. Own much to the experts' reports on public media, INES classification, basic knowledge of nuclear reactor, nuclear safety and protection had been discussed and brain-stormed in detail. (authors)

  18. Remotely Piloted Aircraft Systems and a Wireless Sensors Network for Radiological Accidents

    Directory of Open Access Journals (Sweden)

    A. Reyes-Muñoz

    2016-01-01

    Full Text Available In critical radiological situations, the real time information that we could get from the disaster area becomes of great importance. However, communication systems could be affected after a radiological accident. The proposed network in this research consists of distributed sensors in charge of collecting radiological data and ground vehicles that are sent to the nuclear plant at the moment of the accident to sense environmental and radiological information. Afterwards, data would be analyzed in the control center. Collected data by sensors and ground vehicles would be delivered to a control center using Remotely Piloted Aircraft Systems (RPAS as a message carrier. We analyze the pairwise contacts, as well as visiting times, data collection, capacity of the links, size of the transmission window of the sensors, and so forth. All this calculus was made analytically and compared via network simulations.

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

    International Nuclear Information System (INIS)

    Heams, T.J.; Williams, D.A.; Johns, N.A.; Mason, A.; Bixler, N.E.; Grimley, A.J.; Wheatley, C.J.; Dickson, L.W.; Osborn-Lee, I.; Domagala, P.; Zawadzki, S.; Rest, J.; Alexander, C.A.; Lee, R.Y.

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  1. Research on alarm triggered fault-diagnosis expert system for U-shaped tube breaking accident of steam generators

    International Nuclear Information System (INIS)

    Qian Hong; Luo Jianbo; Jin Weixiao; Zhou Jinming; Wang Du

    2015-01-01

    According to the U-shaped tube breaking accident of steam generator (SGTR), this paper designs a fault-diagnosis expert system based on the alarm triggering. By analyzing the fault mechanism of SGTR accidents, the fault symptom is obtained. The parameters of the belief rule are set up based on the simulation experiment. The information fusion is conducted on the fault-diagnosis results from multiple expert systems to obtain the final diagnose result. The test result shows that the expert system can diagnose the SGTR accident accurately and rapidly, and provide with the operation guidance. (authors)

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

    International Nuclear Information System (INIS)

    Nomoto, Tsutomu

    2015-01-01

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

  3. Environmental impact of accidents involving radioactive material shipping systems

    International Nuclear Information System (INIS)

    Yoshimura, H.R.; Pope, R.B.; Huerta, M.; Nilson, R.H.

    1978-01-01

    Four full-scale spent fuel cask crash tests have been performed, including two head-on truck-barrier impacts (100 and 135 km/h), one railcar-barrier impact (130 km/h), and one locomotive grade crossing impact (130 km/h). Releases to the environment were limited to seepage of about 100 cc of cavity liquid from the cask head in the 135 km/h truck impact test and a slight head seal air leak in the 130 km/h locomotive grade crossing test. These releases were well within the limits specified by the NRC regulations, would have been easily cleaned up, and would have caused little effect on the environment and virtually no risk to the public. To further evaluate cask capability, the crashed spent-fuel rail cask system was fire tested. The cask withstood 90 minutes of a fully engulfing hydrocarbon pool fire while maintaining its structural integrity. At approximately 100 minutes into the fire test, the outer shell of the cask cracked resulting in the partial loss of lead radiation shielding. The failure of the shell was attributed to poor quality control during the original fabrication of the cask in the early 1960's. Present regulatory standards would prevent such occurrences in casks built and licensed today. In addition, the test was much more severe than the qualification criteria specified by present licensing requirements. 4 tables, 13 figures

  4. Portable Filtered Air Suction System for Released Radioactive Gases Prevention under a Severe Accident of NPPs

    International Nuclear Information System (INIS)

    Gu, Beom W.; Choi, Su Y.; Rim, Chun T.

    2013-01-01

    In this paper, the portable filtered air suction system (PoFASS) for released radioactive gases prevention under a severe accident of NPP is proposed. This technology can prevent the release of the radioactive gases to the atmosphere and it can be more economical than FVCS because PoFASS can cover many NPPs with its high mobility. The conceptual design of PoFASS, which has the highest cost effectiveness and robustness to the environment condition such as wind velocity and precipitation, is suggested and the related previous research is introduced in this paper. The portable filtered air suction system (PoFASS) for released radioactive gases prevention can play a key role to mitigate the severe accident of NPP with its high cost effectiveness and robustness to the environment conditions. As further works, the detail design of PoFASS to fabricate a prototype for a demonstration will be proceeded. When released radioactive gases from the broken containment building in the severe accident of nuclear power plants (NPPs) such as the Chernobyl and Fukushima accidents occur, there are no ways to prevent the released radioactive gases spreading in the air. In order to solve this problem, several European NPPs have adopted the filtered vented containment system (FVCS), which can avoid the containment failure through a pressure relief capability to protect the containment building against overpressure. However, the installation cost of FVCS for a NPP is more than $10 million and this system has not been widely welcomed by NPP operating companies due to its high cost

  5. Can cultural differences lead to accidents? Team cultural differences and sociotechnical system operations.

    Science.gov (United States)

    Strauch, Barry

    2010-04-01

    I discuss cultural factors and how they may influence sociotechnical system operations. Investigations of several major transportation accidents suggest that cultural factors may have played a role in the causes of the accidents. However, research has not fully addressed how cultural factors can influence sociotechnical systems. I review literature on cultural differences in general and cultural factors in sociotechnical systems and discuss how these differences can affect team performance in sociotechnical systems. Cultural differences have been observed in social and interpersonal dimensions and in cognitive and perceptual styles; these differences can affect multioperator team performance. Cultural factors may account for team errors in sociotechnical systems, most likely during high-workload, high-stress operational phases. However, much of the research on cultural factors has methodological and interpretive shortcomings that limit their applicability to sociotechnical systems. Although some research has been conducted on the role of cultural differences on team performance in sociotechnical system operations, considerable work remains to be done before the effects of these differences can be fully understood. I propose a model that illustrates how culture can interact with sociotechnical system operations and suggest avenues of future research. Given methodological challenges in measuring cultural differences and team performance in sociotechnical system operations, research in these systems should use a variety of methodologies to better understand how culture can affect multioperator team performance in these systems.

  6. Analysis of the Processes in Spent Fuel Pools in Case of Loss of Heat Removal due to Water Leakage

    Directory of Open Access Journals (Sweden)

    Algirdas Kaliatka

    2013-01-01

    Full Text Available The safe storage of spent fuel assemblies in the spent fuel pools is very important. These facilities are not covered by leaktight containment; thus, the consequences of overheating and melting of fuel in the spent fuel pools can be very severe. On the other hand, due to low decay heat of fuel assemblies, the processes in pools are slow in comparison with processes in reactor core during LOCA accident. Thus, the accident management measures play a very important role in case of some accidents in spent fuel pools. This paper presents the analysis of possible consequences of fuel overheating due to leakage of water from spent fuel pool. Also, the accident mitigation measure, the late injection of water was evaluated. The analysis was performed for the Ignalina NPP Unit 2 spent fuel pool, using system thermal hydraulic code for severe accident analysis ATHLET-CD. The phenomena, taking place during such accident, are discussed. Also, benchmarking of results of the same accident calculation using ASTEC and RELAP/SCDAPSIM codes is presented here.

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

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1991-01-01

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

  8. Accident identification system with automatic detection of abnormal condition using quantum computation

    International Nuclear Information System (INIS)

    Nicolau, Andressa dos Santos; Schirru, Roberto; Lima, Alan Miranda Monteiro de

    2011-01-01

    Transient identification systems have been proposed in order to maintain the plant operating in safe conditions and help operators in make decisions in emergency short time interval with maximum certainty associated. This article presents a system, time independent and without the use of an event that can be used as a starting point for t = 0 (reactor scram, for instance), for transient/accident identification of a pressurized water nuclear reactor (PWR). The model was developed in order to be able to recognize the normal condition and three accidents of the design basis list of the Nuclear Power Plant Angra 2, postulated in the Final Safety Analysis Report (FSAR). Were used several sets of process variables in order to establish a minimum set of variables considered necessary and sufficient. The optimization step of the identification algorithm is based upon the paradigm of Quantum Computing. In this case, the optimization metaheuristic Quantum Inspired Evolutionary Algorithm (QEA) was implemented and works as a data mining tool. The results obtained with the QEA without the time variable are compatible to the techniques in the reference literature, for the transient identification problem, with less computational effort (number of evaluations). This system allows a solution that approximates the ideal solution, the Voronoi Vectors with only one partition for the classes of accidents with robustness. (author)

  9. Colorectal Anastomotic Leakage: New perspectives

    NARCIS (Netherlands)

    F. Daams (Freek)

    2014-01-01

    markdownabstract__Abstract__ This thesis provides new perspectives on colorectal anastomotic leakages. In both experimental and clinical studies, aspects of prevention, early identification, treatment and consequences of anastomotic leakage are discussed.

  10. Validation and application of the system code ATHLET-CD for BWR severe accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

    2016-10-15

    Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

  11. Fusion radioactivity confinement and application to postulated ITER accidents

    International Nuclear Information System (INIS)

    Piet, S.J.; Brereton, S.J.

    1991-01-01

    An assessment of the ITER radioactivity confinement shows reduction of potential accidental releases to the environment by two orders of magnitude. Important credits are the 1% volume/day confinement leakage rate, radioactivity decay for short-lived isotopes, resumption of detritiation/negative pressure within seven days of the accident, and wind meander during the slow confinement leakage. Achieving this two order of magnitude credit in practice requires appropriate design details, especially the leakage rate and detritiation/negative pressure equipment, and research to validate some key assumptions. The confinement maximizes dependence on passive safety features, thereby working toward using fusion's potential safety advantages. The confinement includes several confinement zones with varying human access allowances. Some confinement areas are normally isolated from the environment, the closed ventilation zone. Some areas have an inert cover gas to inhibit combustion. If future assessments of accidental overpressure show the need, we propose a filter/vent system. This report documents our work for the ITER Conceptual Design Activity (CDA). The report is consistent with the final CDA design reports and descriptions, except that our analysis includes a filter/vent. For gaseous or vapor tritium and for most activated aerosols, the reference release fraction is about 2%. For short-lived tungsten-rhenium aerosols, the reference release fraction is somewhat lower, as low as 0.5% for some accident scenarios. Even without resumption of detritiation/decontamination or negative pressure within seven days of the accident, the release fraction for stays below 4%

  12. Quantifying information leakage of randomized protocols

    DEFF Research Database (Denmark)

    Biondi, Fabrizio; Legay, Axel; Malacaria, Pasquale

    2015-01-01

    The quantification of information leakage provides a quantitative evaluation of the security of a system. We propose the usage of Markovian processes to model deterministic and probabilistic systems. By using a methodology generalizing the lattice of information approach we model refined attackers...... capable to observe the internal behavior of the system, and quantify the information leakage of such systems. We also use our method to obtain an algorithm for the computation of channel capacity from our Markovian models. Finally, we show how to use the method to analyze timed and non-timed attacks...

  13. Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code

    International Nuclear Information System (INIS)

    Meng Lin; Rui Hu; Yun Su; Ronghua Zhang; Yanhua Yang

    2005-01-01

    Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent

  14. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 Degree-Sign C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated

  15. Modeling of Spray System Operation under Hydrogen and Steam Emissions in NPP Containment during Severe Accident

    Directory of Open Access Journals (Sweden)

    Vadim E. Seleznev

    2011-01-01

    Full Text Available The paper describes one of the variants of mathematical models of a fluid dynamics process inside the containment, which occurs in the conditions of operation of spray systems in severe accidents at nuclear power plant. The source of emergency emissions in this case is the leak of the coolant or rupture at full cross-section of the main circulating pipeline in a reactor building. Leak or rupture characteristics define the localization and the temporal law of functioning of a source of emergency emission (or accrued operating of warmed up hydrogen and steam in the containment. Operation of this source at the course of analyzed accident models should be described by the assignment of the relevant Dirichlet boundary conditions. Functioning of the passive autocatalytic recombiners of hydrogen is described in the form of the complex Newton boundary conditions.

  16. Design considerations for post accident monitoring system of a research reactor

    International Nuclear Information System (INIS)

    Jang, Gwi Sook; Park, Je Yun; Kim, Young Ki

    2012-01-01

    The Post Accident Monitoring System (PAMS) provides primary information for operators to assess the plant conditions and perform their role in bringing the plant to a safe condition during an accident. The PAMS of NPP (Nuclear Power Plant) in KOREA provides the continuous display of the PAM category 1 parameters specified in R.G 1.97, Rev. 03. Recently the PAMS of NPP has been designed according to R.G 1.97, Rev. 04. There is no PAMS at the HANARO in KOREA, but recently RRs (Research Reactors) around the world are going to have PAMS for various multi purposes. We should determine the design considerations for PAMS in a Korean RR based on the design state analysis. Thus, this paper proposes strategies on the design considerations for the PAMS of a Korean RR

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  18. Study on the establishment of retrospective dosimetry system for nuclear radiation accident(II)

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Jae Shik; Chai, Ha Seok; Lee, Jong Ok [Chungnam National Univ., Taejon (Korea, Republic of)

    1999-03-15

    This study was driven forward centering around physical techniques in retrospective dosimetry system for encountering nuclear radiation accident. The results obtained through this study are summarized as follow : the minimal facilities based on physical techniques should be assured at KINS for appropriate operation and establishment of retrospective accident dosimetry system, the necessary apparatus and man power for retrospective dose assessment by physical techniques might be operated flexibly, however, CL and TL/OSL readers should be equipped with the highest priority, a series of comparative examination of several physical techniques for retrospective dose assessment revealed that most of the irradiated materials around accident sites are usable for the dose assessment, if a priori study on the dosimetrical characteristics of those materials is preceded in accordance with the species of the collectable samples, the results of the study on the CL-dose response and radiation energy dependence of sugar and sorbitol, showed the nonlinearity in CL-dose relationship at the range of low dose(less than 5 Gy), and it led us to perform a study on the correction of the nonlinearity, and in the later study, CL output showed heavy dependence on radiation energy in the energy below around 100 keV and accordingly, a study on the correction for the energy dependence was also carried out, ve were able to obtain good results as a first attempt to carry out such corrections.

  19. Development of filter module for passive filtration and accident gas release confinement system for NPP

    International Nuclear Information System (INIS)

    Yelizarov, P.G.; Efanov, A.D.; Martynov, P.N.; Masalov, D.P.; Osipov, V.P.; Yagodkin, I.V.

    2005-01-01

    Full text of publication follows: One of the urgent problems of the safe NPP operation is air cleaning from radioactive aerosols and volatile iodine compounds under the accident operation conditions of NPP. A principally new passive accident gas release confinement system is used as the basis of the designs of new generation reactor power blocks under the-beyond-design-basis accident conditions with total loss of current. The basic structural component of the passive filtration system (PFS) is the filter-sorber being heated up to 300 deg. C. The filter-sorber represents a design consisting of 150 connected in parallel two-step filtering modules. The first step is intended to clean air from radioactive aerosols, the second one - to clean air from radioactive iodine and its volatile compounds. The filter-sorber is located in the upper point of the exterior protection shell. Due to natural convection, it provides confinement of r/a impurities and controlled steam-gas release from the inter-shell space into atmosphere. The basic specific design feature is the two-section design of the PFS filter module consisting of a coarse-cleaning section and a fine-cleaning section. A combination of layer-by-layer put filtering materials on the basis of glass fiber and metal fiber. The pilot PFS filter module specimen tests run in conditions modeling accident situation indicated that at a filtration rate of 0,3 cm/s the aerodynamic resistance of the module does not exceed 12 Pa, the filtration effectiveness equals 99,99 % in terms of aerosol, no less than 99,9% in terms of radioactive 131 I and no less than 99,0% in terms of organic compounds of iodine (CH 3 131 I); the dust capacity amounts to a value above 50 g/m 2 . The obtained results of tests comply with the design requirements imposed on the PFS filter-sorber module. (authors)

  20. Development of training system to prevent accidents during decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    Jeong, Kwanseong; Moon, Jeikwon; Choi, Byungseon; Hyun, Dongjun; Lee, Jonghwan; Kim, Ikjune; Kim, Geunho; Seo, Jaeseok

    2014-01-01

    Decommissioning workers need familiarization with working environments because working environment is under high radioactivity and work difficulty during decommissioning of nuclear facilities. On-the-job training of decommissioning works could effectively train decommissioning workers but this training approach could consume much costs and poor modifications of scenarios. The efficiency of virtual training system could be much better than that of physical training system. This paper was intended to develop the training system to prevent accidents for decommissioning of nuclear facilities. The requirements for the training system were drawn. The data management modules for the training system were designed. The training system of decommissioning workers was developed on the basis of virtual reality which is flexibly modified. The visualization and measurement in the training system were real-time done according as changes of the decommissioning scenario. It can be concluded that this training system enables the subject to improve his familiarization about working environments and to prevent accidents during decommissioning of nuclear facilities

  1. Development of training system to prevent accidents during decommissioning of nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwanseong; Moon, Jeikwon; Choi, Byungseon; Hyun, Dongjun; Lee, Jonghwan; Kim, Ikjune; Kim, Geunho; Seo, Jaeseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Decommissioning workers need familiarization with working environments because working environment is under high radioactivity and work difficulty during decommissioning of nuclear facilities. On-the-job training of decommissioning works could effectively train decommissioning workers but this training approach could consume much costs and poor modifications of scenarios. The efficiency of virtual training system could be much better than that of physical training system. This paper was intended to develop the training system to prevent accidents for decommissioning of nuclear facilities. The requirements for the training system were drawn. The data management modules for the training system were designed. The training system of decommissioning workers was developed on the basis of virtual reality which is flexibly modified. The visualization and measurement in the training system were real-time done according as changes of the decommissioning scenario. It can be concluded that this training system enables the subject to improve his familiarization about working environments and to prevent accidents during decommissioning of nuclear facilities.

  2. Water Leakage and Nitrate Leaching Characteristics in the Winter Wheat–Summer Maize Rotation System in the North China Plain under Different Irrigation and Fertilization Management Practices

    Directory of Open Access Journals (Sweden)

    Shufeng Chen

    2017-02-01

    Full Text Available Field experiments were carried out in Huantai County from 2006 to 2008 to evaluate the effects of different nitrogen (N fertilization and irrigation management practices on water leakage and nitrate leaching in the dominant wheat–maize rotation system in the North China Plain (NCP. Two N fertilization (NF1, the traditional one; NF2, fertilization based on soil testing and two irrigation (IR1, the traditional one; IR2, irrigation based on real-time soil water content monitoring management practices were designed in the experiments. Water and nitrate amounts leaving the soil layer at a depth of 2.0 m below the soil surface were calculated and compared. Results showed that the IR2 effectively reduced water leakage and nitrate leaching amounts in the two-year period, especially in the winter wheat season. Less than 10 percent irrigation water could be saved in a dry winter wheat season, but about 60 percent could be saved in a wet winter wheat season. Besides, 58.8 percent nitrate under single NF2IR1 and 85.2 percent under NF2IR2 could be prevented from leaching. The IR2 should be considered as the best management practice to save groundwater resources and prevent nitrate from leaching. The amounts of N input play a great role in affecting nitrate concentrations in the soil solutions in the winter wheat–summer maize rotation system. The NF2 significantly reduced N inputs and should be encouraged in ordinary agricultural production. Thus, nitrate leaching and groundwater contamination could be alleviated, but timely N supplement might be needed under high precipitation condition.

  3. Sustainable management of leakage from wastewater pipelines.

    Science.gov (United States)

    DeSilva, D; Burn, S; Tjandraatmadja, G; Moglia, M; Davis, P; Wolf, L; Held, I; Vollertsen, J; Williams, W; Hafskjold, L

    2005-01-01

    Wastewater pipeline leakage is an emerging concern in Europe, especially with regards to the potential effect of leaking effluent on groundwater contamination and the effects infiltration has on the management of sewer reticulation systems. This paper describes efforts by Australia, in association with several European partners, towards the development of decision support tools to prioritize proactive rehabilitation of wastewater pipe networks to account for leakage. In the fundamental models for the decision support system, leakage is viewed as a function of pipeline system deterioration. The models rely on soil type identification across the service area to determine the aggressiveness of the pipe environment and for division of the area into zones based on pipe properties and operational conditions. By understanding the interaction between pipe materials, operating conditions, and the pipe environment in the mechanisms leading to pipe deterioration, the models allow the prediction of leakage rates in different zones across a network. The decision support system utilizes these models to predict the condition of pipes in individual zones, and to optimize the utilization of rehabilitation resources by targeting the areas with the highest leakage rates.

  4. Radiological aspects of nuclear accident scenarios. Volume 2 the Rade-Aid system post-Chernobyl action

    International Nuclear Information System (INIS)

    Sinnaeve, J.

    1991-01-01

    In the event of a nuclear accident, there is a need for a rapid assessment of the resulting levels of environmental contamination in order to facilitate decisions on possible countermeasures. Volume 2 describes the RADE-AID project to develop a computer system which can be used to support the formulation of decisions on countermeasures following an accidental release of radionuclides. The system is intended as an aid following an actual accident and a tool for assistance in planning and training

  5. A system of safety management practices and worker engagement for reducing and preventing accidents: an empirical and theoretical investigation.

    Science.gov (United States)

    Wachter, Jan K; Yorio, Patrick L

    2014-07-01

    The overall research objective was to theoretically and empirically develop the ideas around a system of safety management practices (ten practices were elaborated), to test their relationship with objective safety statistics (such as accident rates), and to explore how these practices work to achieve positive safety results (accident prevention) through worker engagement. Data were collected using safety manager, supervisor and employee surveys designed to assess and link safety management system practices, employee perceptions resulting from existing practices, and safety performance outcomes. Results indicate the following: there is a significant negative relationship between the presence of ten individual safety management practices, as well as the composite of these practices, with accident rates; there is a significant negative relationship between the level of safety-focused worker emotional and cognitive engagement with accident rates; safety management systems and worker engagement levels can be used individually to predict accident rates; safety management systems can be used to predict worker engagement levels; and worker engagement levels act as mediators between the safety management system and safety performance outcomes (such as accident rates). Even though the presence of safety management system practices is linked with incident reduction and may represent a necessary first-step in accident prevention, safety performance may also depend on mediation by safety-focused cognitive and emotional engagement by workers. Thus, when organizations invest in a safety management system approach to reducing/preventing accidents and improving safety performance, they should also be concerned about winning over the minds and hearts of their workers through human performance-based safety management systems designed to promote and enhance worker engagement. Copyright © 2013 The Authors. Published by Elsevier Ltd.. All rights reserved.

  6. Quantifying Information Leakage of Randomized Protocols

    DEFF Research Database (Denmark)

    Biondi, Fabrizio; Wasowski, Andrzej; Legay, Axel

    2013-01-01

    The quantification of information leakage provides a quantitative evaluation of the security of a system. We propose the usage of Markovian processes to model and analyze the information leakage of deterministic and probabilistic systems. We show that this method generalizes the lattice...... of information approach and is a natural framework for modeling refined attackers capable to observe the internal behavior of the system. We also use our method to obtain an algorithm for the computation of channel capacity from our Markovian models. Finally, we show how to use the method to analyze timed...

  7. Accidents - Chernobyl accident

    International Nuclear Information System (INIS)

    2004-01-01

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

  8. Severe accident testing of a personnel airlock

    International Nuclear Information System (INIS)

    Clauss, D.B.; Parks, M.B.; Julien, J.T.; Peters, S.W.

    1988-01-01

    Sandia National Laboratories (Sandia) is investigating the leakage potential of mechanical penetrations as part of a research program on containment integrity under severe accident loads for the U.S. Nuclear Regulatory Commission (NRC). Barnes et al. (1984) and Shackelford et al. (1985) identified leakage from personnel airlocks as an important failure mode of containments subject to severe accident loads. However, these studies were based on relatively simple analysis methods. The complex structural interaction between the door, gasket, and bulkhead in personnel airlocks makes analytical evaluation of leakage difficult. In order to provide data to validate methods for evaluating the leakage potential, a full-size personnel airlock was subject to simulated severe accident loads consisting of pressure and temperature up to 300 psig and 800 degrees F. The test was conducted at Chicago Bridge and Iron under contract to Sandia. The authors provide a detailed report on the test program

  9. Analysis and model testing of a Super Tiger Type B waste transport system in accident environments

    International Nuclear Information System (INIS)

    May, R.A.; Yoshimura, H.R.; Romesberg, L.E.; Joseph, B.J.

    1980-01-01

    Sandia National Laboratories is investigating the response of a Type B packaging containing drums of contact-handled transuranic waste (CH-TRU) as a part of a program to evaluate the adequacy of experimental and analytical methods for assessing the safety of waste transport systems in accident environments. A US NRC certified Type B package known as the Super Tiger was selected for the study. This overpack consists of inner and outer steel shells separated by rigid polyurethane foam and can be used for either highway or rail transportation. Tests using scale models of the vehicular system are being conducted in conjunction with computer analyses

  10. Use of probabilistic safety assessment in structuring conceptual design of accident mitigation systems

    Energy Technology Data Exchange (ETDEWEB)

    Nishiura, Hiroshi; Urata, Shigeru; Tsujikura, Yonezo [Kansai Electric Power Co., Inc., Osaka (Japan); Kuroiwa, Katsuya; Fujimoto, Haruo

    2000-07-01

    When there is an opportunity to develop a new safety design, it should be a rational design that serves its intended purpose while giving due consideration to factors such as reliability, economic efficiency, and others. Therefore, we have aimed to establish a methodical conceptual design process for accident mitigation systems as part of the core cooling system. In this consideration, we have proposed a process made up of 4 steps and have confirmed that the PSA method can be used as a tool in this process. (author)

  11. Use of probabilistic safety assessment in structuring conceptual design of accident mitigation systems

    International Nuclear Information System (INIS)

    Nishiura, Hiroshi; Urata, Shigeru; Tsujikura, Yonezo; Kuroiwa, Katsuya; Fujimoto, Haruo

    2000-01-01

    When there is an opportunity to develop a new safety design, it should be a rational design that serves its intended purpose while giving due consideration to factors such as reliability, economic efficiency, and others. Therefore, we have aimed to establish a methodical conceptual design process for accident mitigation systems as part of the core cooling system. In this consideration, we have proposed a process made up of 4 steps and have confirmed that the PSA method can be used as a tool in this process. (author)

  12. A decision support system for emergency response to major nuclear accidents

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  13. The United States Department of Energy (DOE) Computerized Accident/Incident Reporting System (CAIRS)

    International Nuclear Information System (INIS)

    Briscoe, G.J.

    1993-01-01

    The Department of Energy's (DOE) Computerized Accident/Incident Reporting System (CAIRS) is a comprehensive data base containing more than 50,000 investigation reports of injury/illness, property damage and vehicle accident cases representing safety data from 1975 to the present for more than 150 DOE contractor organizations. A special feature is that the text of each accident report is translated using a controlled dictionary and rigid sentence structure called Factor Relationship and Sequence of Events (FRASE) that enhances the ability to retrieve specific types of information and to perform detailed analyses. DOE summary and individual contractor reports are prepared quarterly and annually. In addition, ''Safety Performance Profile'' reports for individual organizations are prepared to provide advance information to appraisal teams, and special topical reports are prepared for areas of concern such as an increase in the number of security injuries or environmental releases. The data base is open to all DOE and Contractor registered users with no access restrictions other than that required by the Privacy Act

  14. Assisting emergency operating procedures execution with AMAS, an Accident Management Advisor System

    International Nuclear Information System (INIS)

    Guarro, S.; Milici, T.; Wu, J.S.; Apostolakis, G.

    2004-01-01

    In an accident situation, because any decisions that the operators make will depend on how instrumentation readings are ultimately interpreted, the issue of instrument uncertainty is of paramount importance. This uncertainty exists because instrument readings may not be available in the desired form - i.e., only indirect readings for a parameter of interest may exist, with uncertainty on which physical models may be used to deduce its value from these indirect indications -, or because readings may be coming from instruments whose accuracy and reliability in the face of the severe conditions produced by the accident are far from what may be expected under normal operating conditions. In following the EOPs, the operators must rely on instrumentation whose readings may not reflect the real situation. The Accident Management Advisor System (AMAS) is a decision aid intended to supplement plant Emergency Operating Procedures (EOPs) by accounting for instrumentation uncertainty, and by alerting the operators if they are on the wrong procedures, or otherwise performing an action that is not optimal in terms of preventing core damage. In AMAS, the availability and reliability of certain important instrument readings is treated in probabilistic, rather than deterministic terms. This issue is discussed in greater detail later in the paper, since it relates to one of the key characteristics of the AMAS decision aid. (author)

  15. Quantitative Safety Impact of Severe Accident Management Systems for EU-APR during Low Power Shutdown Operation

    International Nuclear Information System (INIS)

    Lee, Keunsung; Hwang, Do Hun; Chang, Hyun-bin

    2016-01-01

    In order to enlarge and to diversify the export market of APR1400, the EU-APR design was developed based on the APR1400 design to comply with the latest version of the European Utility Requirements (EUR) revision D. The EU-APR design has the distinguished and advanced severe accident management systems taken from the APR1400 to obtain a containment integrity for the beyond design basis accident, such as the Passive Ex-vessel retaining and Cooling System (PECS), the Severe Accident Containment Spray System (SACSS) and the Containment Filtered Vent System (CFVS). The risk associated with the nuclear power plant can be identified through the Probabilistic Safety Assessment (PSA). In the EUR chapter 1 and 17 of volume 2, the Criteria for Limited Impact (CLI) should be applied to the Level 2 PSA as a risk metrics. The fraction of exceeding CLI for the EU-APR during LPSD operation was calculated as 4.52% of the CDF under the condition that all severe accident management systems are credited. The PECS, SACSS and CFVS are considered as the severe accident management system which is EU-APR dedicated system. The exemption of each system leads to increase the fraction of exceeding CLI to 54.18%, 89.74% and 21.32% respectively. In case if all these systems are unavailable, the fraction of exceeding CLI is increased to 100%. The most effective system is the SACSS that the system reduces containment pressure and temperature

  16. Quantitative Safety Impact of Severe Accident Management Systems for EU-APR during Low Power Shutdown Operation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Keunsung; Hwang, Do Hun [KHNP CRI, Daejeon (Korea, Republic of); Chang, Hyun-bin [Future and Challenge Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    In order to enlarge and to diversify the export market of APR1400, the EU-APR design was developed based on the APR1400 design to comply with the latest version of the European Utility Requirements (EUR) revision D. The EU-APR design has the distinguished and advanced severe accident management systems taken from the APR1400 to obtain a containment integrity for the beyond design basis accident, such as the Passive Ex-vessel retaining and Cooling System (PECS), the Severe Accident Containment Spray System (SACSS) and the Containment Filtered Vent System (CFVS). The risk associated with the nuclear power plant can be identified through the Probabilistic Safety Assessment (PSA). In the EUR chapter 1 and 17 of volume 2, the Criteria for Limited Impact (CLI) should be applied to the Level 2 PSA as a risk metrics. The fraction of exceeding CLI for the EU-APR during LPSD operation was calculated as 4.52% of the CDF under the condition that all severe accident management systems are credited. The PECS, SACSS and CFVS are considered as the severe accident management system which is EU-APR dedicated system. The exemption of each system leads to increase the fraction of exceeding CLI to 54.18%, 89.74% and 21.32% respectively. In case if all these systems are unavailable, the fraction of exceeding CLI is increased to 100%. The most effective system is the SACSS that the system reduces containment pressure and temperature.

  17. Assessment of chemical processes for the post-accident decontamination of reactor-coolant systems. Final report

    International Nuclear Information System (INIS)

    Munson, L.F.; Card, C.J.; Divine, J.R.

    1983-02-01

    Previously used chemical decontamination processes and potentially useful new decontamination processes were examined for the usefulness following a reactor accident. Both generic fuel damage accidents and the accident at TMI-2 were considered. A total of fourteen processes were evaluated. Process evaluation included data in the following categories: technical description of the process, recorded past usage, effectiveness, process limitation, safety consideration, and waste management. These data were evaluated, and cost considerations were presented along with a description of the applicability of the process to TMI-2 and development and demonstration needs. Specific recommendations regarding a primary-system decontamination development program to support TMI-2 recovery were also presented

  18. An assessment of Class-9 (core-melt) accidents for PWR dry-containment systems

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Saito, M.

    1981-01-01

    The phenomenology of core-melt accidents in dry containments was examined for the purpose of identifying the margins of safety in such Class-9 situations. The scale (geometry) effects appear to crucially limit the extent (severity) of steam explosions. This together with the established reduced explosivity of the corium-A/water system, and the inherently high capability of dry containments (redinforced concrete, and shields in some cases, seismic design etc.) lead to the conclusion that failure due to steam explosions may be considered essentially incredible. These premixture scaling considerations also impact ultimate debris disposition and coolability and need additional development. A water-flooded reactor cavity would have beneficial effects in limiting (but not necessarily eliminating) melt-concrete interactions. Independently of the initial degree of quenching and/or scale of fragmentation, mechanisms exist that drive the system towards ultimate stability (coolability). Additional studies, with intermediate-scale prototypic materials are recommended to better explore these mechanisms. Containment heat removal systems must provide the crucial capability of mitigating such accidents. Passive systems should be explored and assessed against currently available and/or improved active systems taking into account the rather loose time constraints required for activation. It appears that containment margins for accommodating the hydrogen problem are limited. This problem appears to stand out not only in terms of potential consequences but also in terms of lack of any readily available and clear cut solutions at this time. (orig.)

  19. The Role of Trust and Interaction in Global Positioning System Related Accidents

    Science.gov (United States)

    Johnson, Chris W.; Shea, Christine; Holloway, C. Michael

    2008-01-01

    The Global Positioning System (GPS) uses a network of satellites to calculate the position of a receiver over time. This technology has revolutionized a wide range of safety-critical industries and leisure applications. These systems provide diverse benefits; supplementing the users existing navigation skills and reducing the uncertainty that often characterizes many route planning tasks. GPS applications can also help to reduce workload by automating tasks that would otherwise require finite cognitive and perceptual resources. However, the operation of these systems has been identified as a contributory factor in a range of recent accidents. Users often come to rely on GPS applications and, therefore, fail to notice when they develop faults or when errors occur in the other systems that use the data from these systems. Further accidents can stem from the over confidence that arises when users assume automated warnings will be issued when they stray from an intended route. Unless greater attention is paid to the role of trust and interaction in GPS applications then there is a danger that we will see an increasing number of these failures as positioning technologies become integral in the functioning of increasing numbers of applications.

  20. An Ontology-Underpinned Emergency Response System for Water Pollution Accidents

    Directory of Open Access Journals (Sweden)

    Xiaoliang Meng

    2018-02-01

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

  1. Problem of corium melt coolability in passive protection systems against severe accidents in the containment

    Directory of Open Access Journals (Sweden)

    Ali Kalvand

    2018-05-01

    Full Text Available Paper is devoted to the development of the mathematical model and analysis of the problem of corium melt interaction with low-temperature melting blocks in the passive protection systems against severe accidents at the NPP, which is of high importance for substantiation of the nuclear power safety, for building and successful op-erating of passive protection systems. In the third-generation reactors passive protection systems against severe accidents at the NPP are mandatory, therefore this paper is of importance for the nuclear power safety. A few configurations for the cooling blocks’ distribution have been considered and an analysis of the blocks’ melting and corium’s cooling in the pool under reactor vessel have been done, which can serve more effective for further improvement of the safety current systems and for the development of new ones. The ways for solution of the problems and the methods for their successful elaboration were discussed. The developed mathematical models and the analysis performed in the paper might be helpful for the design of passive protection systems of the cori-um melt retention inside the containment after corium melt eruption from the broken reactor vessel.

  2. Postulated accidents

    International Nuclear Information System (INIS)

    Ullrich, W.

    1980-01-01

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

  3. Incident warning systems : accident review. DRIVE II Project V2002 Horizontal Project for the Evaluation of Safety HOPES, Deliverable 17, Workpackage 31, Activity 31.2.

    NARCIS (Netherlands)

    Oppe, S. Lindeijer, J.E. & Barjonet, P.

    1995-01-01

    The objective of this accident review is to check what proportion of accidents recorded in the past could in principle have been prevented by using an incident warning system (IWS). The accident review was carried out for all three IWS test sites that are part of the HOPES evaluation study. These

  4. Determination of leakage areas in nuclear piping

    International Nuclear Information System (INIS)

    Keim, E.

    1997-01-01

    For the design and operation of nuclear power plants the Leak-Before-Break (LBB) behavior of a piping component has to be shown. This means that the length of a crack resulting in a leak is smaller than the critical crack length and that the leak is safely detectable by a suitable monitoring system. The LBB-concept of Siemens/KWU is based on computer codes for the evaluation of critical crack lengths, crack openings, leakage areas and leakage rates, developed by Siemens/KWU. In the experience with the leak rate program is described while this paper deals with the computation of crack openings and leakage areas of longitudinal and circumferential cracks by means of fracture mechanics. The leakage areas are determined by the integration of the crack openings along the crack front, considering plasticity and geometrical effects. They are evaluated with respect to minimum values for the design of leak detection systems, and maximum values for controlling jet and reaction forces. By means of fracture mechanics LBB for subcritical cracks has to be shown and the calculation of leakage areas is the basis for quantitatively determining the discharge rate of leaking subcritical through-wall cracks. The analytical approach and its validation will be presented for two examples of complex structures. The first one is a pipe branch containing a circumferential crack and the second one is a pipe bend with a longitudinal crack

  5. Determination of leakage areas in nuclear piping

    Energy Technology Data Exchange (ETDEWEB)

    Keim, E. [Siemens/KWU, Erlangen (Germany)

    1997-04-01

    For the design and operation of nuclear power plants the Leak-Before-Break (LBB) behavior of a piping component has to be shown. This means that the length of a crack resulting in a leak is smaller than the critical crack length and that the leak is safely detectable by a suitable monitoring system. The LBB-concept of Siemens/KWU is based on computer codes for the evaluation of critical crack lengths, crack openings, leakage areas and leakage rates, developed by Siemens/KWU. In the experience with the leak rate program is described while this paper deals with the computation of crack openings and leakage areas of longitudinal and circumferential cracks by means of fracture mechanics. The leakage areas are determined by the integration of the crack openings along the crack front, considering plasticity and geometrical effects. They are evaluated with respect to minimum values for the design of leak detection systems, and maximum values for controlling jet and reaction forces. By means of fracture mechanics LBB for subcritical cracks has to be shown and the calculation of leakage areas is the basis for quantitatively determining the discharge rate of leaking subcritical through-wall cracks. The analytical approach and its validation will be presented for two examples of complex structures. The first one is a pipe branch containing a circumferential crack and the second one is a pipe bend with a longitudinal crack.

  6. Leakage radiation interference microscopy.

    Science.gov (United States)

    Descrovi, Emiliano; Barakat, Elsie; Angelini, Angelo; Munzert, Peter; De Leo, Natascia; Boarino, Luca; Giorgis, Fabrizio; Herzig, Hans Peter

    2013-09-01

    We present a proof of principle for a new imaging technique combining leakage radiation microscopy with high-resolution interference microscopy. By using oil immersion optics it is demonstrated that amplitude and phase can be retrieved from optical fields, which are evanescent in air. This technique is illustratively applied for mapping a surface mode propagating onto a planar dielectric multilayer on a thin glass substrate. The surface mode propagation constant estimated after Fourier transformation of the measured complex field is well matched with an independent measurement based on back focal plane imaging.

  7. Pickering unit 1 containment leakage characterization

    International Nuclear Information System (INIS)

    Zakaib, G.D.

    1994-01-01

    Results of the design pressure test carried out on Pickering Reactor Building number 1 during late 1992 showed that the leakage rate of the building was close to the safety analysis value of 2.7% contained mass per hour at the design pressure of 41.4 kPa(g) and was significantly higher than that reported after the previous test conducted in the spring of 1987. This unexpected finding initiated the longest and the most comprehensive containment leakage investigation ever undertaken by Ontario Hydro. A thorough investigation of leakage behaviour by repeated testing, inspections, leak search and analysis was launched. The extensive leak search effort included items such as: leak source detection by soap solution application, use of ultrasonic detectors, fogging and tracer gas techniques, systematic systems isolation, thermal imaging of the exterior, and quantification of leak sites by flowmeter and bagging. Using a specially designed volumetric technique, the root cause of the problem was finally confirmed as being due to 'pressure dependent laminar leakage' through the hairline cracks in the dome concrete. Structural analysis indicated that the thermal gradients and pressure loading combined to cause the cracking early in the structure's operating history and that overall structural integrity has not been compromised. Leakage rate analysis using a new fluid mechanics model augmented by the effect of thermal strains indicated that the leakage could be significantly less under certain transient temperature gradient conditions. Several options for repairing the dome were considered by a multidisciplinary team and it was finally decided to apply a specially engineered multilayer elastomeric coating to the exterior concrete surface. When the unit was re-tested in October 1993, a dramatic ten-fold improvement in leakage rate (down to 0.25%/h at design pressure) was observed. This is lower than even the commissioning results and comparable to the performance of newer units

  8. Markovian Processes for Quantitative Information Leakage

    DEFF Research Database (Denmark)

    Biondi, Fabrizio

    Quantification of information leakage is a successful approach for evaluating the security of a system. It models the system to be analyzed as a channel with the secret as the input and an output as observable by the attacker as the output, and applies information theory to quantify the amount...... and randomized processes with Markovian models and to compute their information leakage for a very general model of attacker. We present the QUAIL tool that automates such analysis and is able to compute the information leakage of an imperative WHILE language. Finally, we show how to use QUAIL to analyze some...... of information transmitted through such channel, thus effectively quantifying how many bits of the secret can be inferred by the attacker by analyzing the system’s output. Channels are usually encoded as matrices of conditional probabilities, known as channel matrices. Such matrices grow exponentially...

  9. Leakage evaluation in the PCV (Primary Containment Vessel) using chemical and radiochemical data

    International Nuclear Information System (INIS)

    Maeda, Katsuji; Nagasawa, Katsumi

    1998-01-01

    Keeping the reliability of nuclear power plant operation, the primary coolant leakage in the PCV is strictly restricted by the Technical Specifications. It is very important to detect an indication of leakage and estimate the source of leakage to provide countermeasures. Usually the indication of leakage will be detected by increase of drain flow in the PCV sump. There are some possibilities of leakage sources in the PCV, such as reactor water, main steam, condensate, feedwater and closed cooling water. The leakage source contain different chemical and radiochemical species. This means that the leakage source can be presumed and detected by using chemical information from the PCV atmosphere and sump water. To detect the leakage indication and the source quickly and exactly, the PCV Leakage Detection Expert System has been developed. This paper describes how to evaluate the leakage indication and source in the PCV by using chemical and radiochemical data. (author)

  10. Accident localization system with jet condensers for VVER 440-V 230 NPP at Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Murani, J.

    1995-01-01

    The operational safety of the V1 nuclear power plant (NPP) is unsatisfactory and does not correspond to present requirements as to nuclear safety. Further NPP operation after 1995 is conditional on nuclear safety enhancement to a level comparable with that in West European countries. This aim should be achieved by a principal reconstruction involving in addition to others also backfitting the V1 NPP with technical facilities aimed at coping with a design basis accident (DBA).To cope with such an accident the Power Equipment Research Institute (VUEZ) designed an accident localization system with jet condensers. This system consists of (a) an air trap (one for each unit, mutually interconnected) with an expansion bell enclosed within, placed on a plate with 200 pipes of jet condensers passing through, and (b) a connecting duct between the hermetic zone and the air trap. The vertical jet condenser is an essential element of the system designed for steam condensation. Apart from condensation it serves as a water seal separating units 1 and 2.Demonstration tests of the jet condenser (model 1:1) condensing function were carried out at the testing unit of the All-Union Research Institute for NPP Operation (VNIIAES), Moscow in Kashir, 11-22 September 1992. These experiments proved the jet condenser ability to ensure complete condensation of the steam produced. Experimental verification of the sealing function (model 1:1) was carried out at the testing unit of the VUEZ Tlmace. These experiments concerning the dynamics and overpressure in the free space above the pool were close to the conditions in the air trap during DBA. The jet condenser height was proved to be sufficient to ensure the sealing function. Design and experimental work has been implemented in close cooperation with Russian experts Mr. V.N. Bulynin from the VNIIAES, Moscow, and Mr. M.V. Kuznecov from the Scientific and Engineering Center for Nuclear and Radiological Safety, Moscow. (orig.)

  11. Accident analysis of heat pipe cooled and AMTEC conversion space reactor system

    International Nuclear Information System (INIS)

    Yuan, Yuan; Shan, Jianqiang; Zhang, Bin; Gou, Junli; Bo, Zhang; Lu, Tianyu; Ge, Li; Yang, Zijiang

    2016-01-01

    Highlights: • A transient analysis code TAPIRS for HPS has been developed. • Three typical accidents are analyzed using TAPIRS. • The reactor system has the self-stabilization ability under accident conditions. - Abstract: A space power with high power density, light weight, low cost and high reliability is of crucial importance to future exploration of deep space. Space reactor is an excellent candidate because of its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe (HP) as core cooling component, is considered as one of the most promising choices and is widely studied all over the world. This paper develops a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) based on point reactor kinetics model, lumped parameter core heat transfer model, combined HP model (self-diffusion model, flat-front startup model and network model), energy conversion model of Alkali Metal Thermal-to-Electric Conversion units (AMTEC), and HP radiator model. Three typical accidents, i.e., control drum failure, AMTEC failure and partial loss of the heat transfer area of radiator are then analyzed using TAPIRS. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. The results show the following. (1) After the failure of one set of control drums, the reactor power finally reaches a stable value after two local peaks under the temperature feedback. The fuel temperature rises rapidly, however it is still under safe limit. (2) The fuel temperature is below a safe limit under the AMTEC failure and partial loss of the heat transfer area of radiator. This demonstrates the rationality of the system design and the potential applicability of the TAPIRS code for the future engineering application of

  12. The development of a nuclear accident risk information system(NARIS)

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Jung, Won Dea

    2001-03-01

    The computerized system, NARIS(Nuclear Accident Risk Information System) was developed in order to support the estimation of health effects and the establishment the effective risk reduction strategies. Using the system, we can analyze the distribution of health effects easily by displaying the results on the digital map of the site. Also, the thematic mapping allows the diverse analysis of the distribution of the health effects.The NARIS can be used in the emergency operation facilities in order to analyze the distribution of the health effects resulting from the severe accidents of a nuclear power plant. Also, the rapid analysis of the health effect is possible by storing the health effect results in the form of a database. Therefore, the staffs of the emergency operation facilities can establish the rapid and effective emergency response strategies. The module for the optimization of the costs and benefits and the decision making support will be added. The technical support for the establishment of the optimum and effective emergency response strategies will be possible using this system.

  13. Development of Highly Reliable Power and Communication System for Essential Instruments Under Severe Accidents in NPP

    Directory of Open Access Journals (Sweden)

    Bo Hwan Choi

    2016-10-01

    Full Text Available This article proposes a highly reliable power and communication system that guarantees the protection of essential instruments in a nuclear power plant under a severe accident. Both power and communication lines are established with not only conventional wired channels, but also the proposed wireless channels for emergency reserve. An inductive power transfer system is selected due to its robust power transfer characteristics under high temperature, high pressure, and highly humid environments with a large amount of scattered debris after a severe accident. A thermal insulation box and a glass-fiber reinforced plastic box are proposed to protect the essential instruments, including vulnerable electronic circuits, from extremely high temperatures of up to 627°C and pressure of up to 5 bar. The proposed wireless power and communication system is experimentally verified by an inductive power transfer system prototype having a dipole coil structure and prototype Zigbee modules over a 7-m distance, where both the thermal insulation box and the glass-fiber reinforced plastic box are fabricated and tested using a high-temperature chamber. Moreover, an experiment on the effects of a high radiation environment on various electronic devices is conducted based on the radiation test having a maximum accumulated dose of 27 Mrad.

  14. Development of highly reliable power and communication system for essential instruments under severe accidents in NPP

    International Nuclear Information System (INIS)

    Choi, Bo Hwan; Jang, Gi Chan; Shin, Sung Min; Kang, Hyun Gook; Rim, Chun Taek; Lee, Soo Ill

    2016-01-01

    This article proposes a highly reliable power and communication system that guarantees the protection of essential instruments in a nuclear power plant under a severe accident. Both power and communication lines are established with not only conventional wired channels, but also the proposed wireless channels for emergency reserve. An inductive power transfer system is selected due to its robust power transfer characteristics under high temperature, high pressure, and highly humid environments with a large amount of scattered debris after a severe accident. A thermal insulation box and a glass-fiber reinforced plastic box are proposed to protect the essential instruments, including vulnerable electronic circuits, from extremely high temperatures of up to 627 .deg. C and pressure of up to 5 bar. The proposed wireless power and communication system is experimentally verified by an inductive power transfer system prototype having a dipole coil structure and prototype Zigbee modules over a 7-m distance, where both the thermal insulation box and the glass-fiber reinforced plastic box are fabricated and tested using a high-temperature chamber. Moreover, an experiment on the effects of a high radiation environment on various electronic devices is conducted based on the radiation test having a maximum accumulated dose of 27 Mrad

  15. A route-specific system for risk assessment of radioactive materials transportation accidents

    International Nuclear Information System (INIS)

    Moore, J.E.; Sandquist, G.M.; Slaughter, D.M.

    1995-01-01

    A low-cost, powerful geographic information system (GIS) that operates on a personal computer was integrated into a software system to provide route specific assessment of the risks associated with the atmospheric release of radioactive and hazardous materials in transportation accidents. The highway transportation risk assessment (HITRA) software system described here combines a commercially available GIS (TransCAD) with appropriate models and data files for route- and accident-specific factors, such as meteorology, dispersion, demography, and health effects to permit detailed analysis of transportation risk assessment. The HITRA system allows a user to interactively select a highway or railroad route from a GIS database of major US transportation routes. A route-specific risk assessment is then performed to estimate downwind release concentrations and the resulting potential health effects imposed on the exposed population under local environmental and temporal conditions. The integration of GIS technology with current risk assessment methodology permits detailed analysis coupled with enhanced user interaction. Furthermore, HITRA provides flexibility and documentation for route planning, updating and improving the databases required for evaluating specific transportation routes, changing meteorological and environmental conditions, and local demographics

  16. Development of high-performance monitoring system under severe accident condition

    International Nuclear Information System (INIS)

    Takeuchi, Tomoaki; Tsuchiya, Kunihiro; Ishihara, Masahiro; Komanome, H.; Miura, K.

    2017-01-01

    A research and development of a monitoring system for NPPs situations even during severe accidents have been performed. The R and D consists of the three objectives. The major findings are briefly summarized in the followings: 1) Radiation-resistant monitoring camera. The image sensor with the photogate and three transistors was found to be advantageous in terms of dark current and sensitivity. In addition, radiation-resistant optical parts and signal circuits were successfully fabricated. The results suggested that the monitoring camera system with 10 6 Gy in radiation resistance was possible. 2) Radiation-resistant in-water wireless transmission system. A two-dimensional LED matrix with 10 6 Gy in radiation resistance and a camera were used as the transmission devices. The results of the in-water transmission tests suggested that stable wireless transmission between 5 m distance was possible even with bubble, turbidity, or obstacles. 3) Heat-resistant signal cable. In order to develop a cable that can transmit the data inside reactor pressure vessels, heat-proof tests were performed for candidate metallic sheath materials of mineral insulation (MI) cables. The results indicated MI cables which can be used at 1000degC in air were possible. These results indicate the feasibility of the monitoring system even during severe accidents. (author)

  17. Development of highly reliable power and communication system for essential instruments under severe accidents in NPP

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Bo Hwan; Jang, Gi Chan; Shin, Sung Min; Kang, Hyun Gook; Rim, Chun Taek [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Lee, Soo Ill [I and C Group, Korea Hydro and Nuclear Power Co., Ltd, Central Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This article proposes a highly reliable power and communication system that guarantees the protection of essential instruments in a nuclear power plant under a severe accident. Both power and communication lines are established with not only conventional wired channels, but also the proposed wireless channels for emergency reserve. An inductive power transfer system is selected due to its robust power transfer characteristics under high temperature, high pressure, and highly humid environments with a large amount of scattered debris after a severe accident. A thermal insulation box and a glass-fiber reinforced plastic box are proposed to protect the essential instruments, including vulnerable electronic circuits, from extremely high temperatures of up to 627 .deg. C and pressure of up to 5 bar. The proposed wireless power and communication system is experimentally verified by an inductive power transfer system prototype having a dipole coil structure and prototype Zigbee modules over a 7-m distance, where both the thermal insulation box and the glass-fiber reinforced plastic box are fabricated and tested using a high-temperature chamber. Moreover, an experiment on the effects of a high radiation environment on various electronic devices is conducted based on the radiation test having a maximum accumulated dose of 27 Mrad.

  18. The development of a nuclear accident risk information system(NARIS)

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Jung, Won Dea [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    The computerized system, NARIS(Nuclear Accident Risk Information System) was developed in order to support the estimation of health effects and the establishment the effective risk reduction strategies. Using the system, we can analyze the distribution of health effects easily by displaying the results on the digital map of the site. Also, the thematic mapping allows the diverse analysis of the distribution of the health effects.The NARIS can be used in the emergency operation facilities in order to analyze the distribution of the health effects resulting from the severe accidents of a nuclear power plant. Also, the rapid analysis of the health effect is possible by storing the health effect results in the form of a database. Therefore, the staffs of the emergency operation facilities can establish the rapid and effective emergency response strategies. The module for the optimization of the costs and benefits and the decision making support will be added. The technical support for the establishment of the optimum and effective emergency response strategies will be possible using this system. 23 figs., 1 tab. (Author)

  19. Issues and challenges for pedestrian active safety systems based on real world accidents.

    Science.gov (United States)

    Hamdane, Hédi; Serre, Thierry; Masson, Catherine; Anderson, Robert

    2015-09-01

    The purpose of this study was to analyze real crashes involving pedestrians in order to evaluate the potential effectiveness of autonomous emergency braking systems (AEB) in pedestrian protection. A sample of 100 real accident cases were reconstructed providing a comprehensive set of data describing the interaction between the vehicle, the environment and the pedestrian all along the scenario of the accident. A generic AEB system based on a camera sensor for pedestrian detection was modeled in order to identify the functionality of its different attributes in the timeline of each crash scenario. These attributes were assessed to determine their impact on pedestrian safety. The influence of the detection and the activation of the AEB system were explored by varying the field of view (FOV) of the sensor and the level of deceleration. A FOV of 35° was estimated to be required to detect and react to the majority of crash scenarios. For the reaction of a system (from hazard detection to triggering the brakes), between 0.5 and 1s appears necessary. Copyright © 2015 Elsevier Ltd. All rights reserved.

  20. Methodology for time-dependent reliability analysis of accident sequences and complex reactor systems

    International Nuclear Information System (INIS)

    Paula, H.M.

    1984-01-01

    The work presented here is of direct use in probabilistic risk assessment (PRA) and is of value to utilities as well as the Nuclear Regulatory Commission (NRC). Specifically, this report presents a methodology and a computer program to calculate the expected number of occurrences for each accident sequence in an event tree. The methodology evaluates the time-dependent (instantaneous) and the average behavior of the accident sequence. The methodology accounts for standby safety system and component failures that occur (a) before they are demanded, (b) upon demand, and (c) during the mission (system operation). With respect to failures that occur during the mission, this methodology is unique in the sense that it models components that can be repaired during the mission. The expected number of system failures during the mission provides an upper bound for the probability of a system failure to run - the mission unreliability. The basic event modeling includes components that are continuously monitored, periodically tested, and those that are not tested or are otherwise nonrepairable. The computer program ASA allows practical applications of the method developed. This work represents a required extension of the presently available methodology and allows a more realistic PRA of nuclear power plants

  1. The development of a nuclear accident risk information system(NARIS)

    International Nuclear Information System (INIS)

    Jeong, Jong Tae; Jung, Won Dea

    2001-03-01

    The computerized system, NARIS(Nuclear Accident Risk Information System) was developed in order to support the estimation of health effects and the establishment the effective risk reduction strategies. Using the system, we can analyze the distribution of health effects easily by displaying the results on the digital map of the site. Also, the thematic mapping allows the diverse analysis of the distribution of the health effects.The NARIS can be used in the emergency operation facilities in order to analyze the distribution of the health effects resulting from the severe accidents of a nuclear power plant. Also, the rapid analysis of the health effect is possible by storing the health effect results in the form of a database. Therefore, the staffs of the emergency operation facilities can establish the rapid and effective emergency response strategies. The module for the optimization of the costs and benefits and the decision making support will be added. The technical support for the establishment of the optimum and effective emergency response strategies will be possible using this system

  2. Modernization of the accident localization system and relevant dose exposure on Unit 4 of KNPP

    Energy Technology Data Exchange (ETDEWEB)

    Valtchev, G.; Neshkova, A.; Nikolov, M. [Nuclear Power Plant Kozloduy, 3321 Kozloduy (Bulgaria)

    2004-07-01

    In 2001 a modernization of the Accident Localization System (ALS) on unit 4 was accomplished. The outage duration was longer then usually and special dose budget was elaborated. All ALS work was performed by external organization. An ALARA implementation was recognized priority. The really accumulated collective doses were analyzed and conclusions drawn. A short film on CD was prepared. Two conclusions are drawn: 1. Good work management and a first attempt of effective empowerment of the workers gave satisfactory results; 2. Although the work was not typical, and performed for a first time, the ALARA implementation reduced the projected collective dose with 19%.

  3. Current statistical tools, systems and bodies concerned with safety and accident statistics. Contribution to the OECD seminar `International Road Traffic and Accident Databases IRTAD', Helsinki, Finland, September 19, 1995.

    NARCIS (Netherlands)

    Koornstra, M.J.

    1996-01-01

    The current use of road safety information systems and the few systems for international use are discussed. Recommendations are formulated for a more efficient, less costly and improved accident registration on the local, national and international levels.

  4. Modeling of severe accident sequences with the new modules CESAR and DIVA of ASTEC system code

    International Nuclear Information System (INIS)

    Pignet, Sophie; Guillard, Gaetan; Barre, Francois; Repetto, Georges

    2003-01-01

    Systems of computer codes, so-called 'integral' codes, are being developed to simulate the scenario of a hypothetical severe accident in a light water reactor, from the initial event until the possible radiological release of fission products out of the containment. They couple the predominant physical phenomena that occur in the different reactor zones and simulate the actuation of safety systems by procedures and by operators. In order to allow to study a great number of scenarios, a compromise must be found between precision of results and calculation time: one day of accident time should take less than one day of real time to simulate on a PC computer. This search of compromise is a real challenge for such integral codes. The development of the ASTEC integral code was initiated jointly by IRSN and GRS as an international reference code. The latest version 1.0 of ASTEC, including the new modules CESAR and DIVA which model the behaviour of the reactor cooling system and the core degradation, is presented here. Validation of the modules and one plant application are described

  5. Systems thinking, the Swiss Cheese Model and accident analysis: a comparative systemic analysis of the Grayrigg train derailment using the ATSB, AcciMap and STAMP models.

    Science.gov (United States)

    Underwood, Peter; Waterson, Patrick

    2014-07-01

    The Swiss Cheese Model (SCM) is the most popular accident causation model and is widely used throughout various industries. A debate exists in the research literature over whether the SCM remains a viable tool for accident analysis. Critics of the model suggest that it provides a sequential, oversimplified view of accidents. Conversely, proponents suggest that it embodies the concepts of systems theory, as per the contemporary systemic analysis techniques. The aim of this paper was to consider whether the SCM can provide a systems thinking approach and remain a viable option for accident analysis. To achieve this, the train derailment at Grayrigg was analysed with an SCM-based model (the ATSB accident investigation model) and two systemic accident analysis methods (AcciMap and STAMP). The analysis outputs and usage of the techniques were compared. The findings of the study showed that each model applied the systems thinking approach. However, the ATSB model and AcciMap graphically presented their findings in a more succinct manner, whereas STAMP more clearly embodied the concepts of systems theory. The study suggests that, whilst the selection of an analysis method is subject to trade-offs that practitioners and researchers must make, the SCM remains a viable model for accident analysis. Copyright © 2013 Elsevier Ltd. All rights reserved.

  6. Review of U.S. Army Unmanned Aerial Systems Accident Reports: Analysis of Human Error Contributions

    Science.gov (United States)

    2018-03-20

    within report documents. The information presented was obtained through a request to use the U.S. Army Combat Readiness Center’s Risk Management ...controlled flight into terrain (13 accidents), fueling errors by improper techniques (7 accidents), and a variety of maintenance errors (10 accidents). The...and 9 of the 10 maintenance accidents. Table 4. Frequencies Based on Source of Human Error Human error source Presence Poor Planning

  7. Having a New Pair of Glassess : Applying Systemic Accident Models on Road Safety

    OpenAIRE

    Huang, Yu-Hsing

    2007-01-01

    The main purpose of the thesis is to discuss the accident models which underlie accident prevention in general and road safety in particular, and the consequences of relying on a particular model have for actual preventive work. The discussion centres on two main topics. The first topic is whether the underlying accident model, or paradigm, of traditional road safety should be exchanged for a more complex accident model, and if so, which model(s) are appropriate. From a discussion of current ...

  8. Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

    Science.gov (United States)

    Susyadi; Endiah, P. H.; Sukmanto, D.; Andi, S. E.; Syaiful, B.; Hendro, T.; Geni, R. S.

    2018-02-01

    RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

  9. Review of Atomic Energy Laws Related to Radiological Accidents and Methods of Improvement

    International Nuclear Information System (INIS)

    Chang, Gun Hyun; Kim, Sang Won; Yoo, Jeong; Ahn, Hyoung Jun; Park, Young Sik; Kim, Hong Suk; Kwon, Jeong Wan; Jang, Ki Won; Kim, Sok Chul

    2009-01-01

    Atomic energy-related laws in Korea have a two pronged management system for radiological accidents. To be specific, the Atomic Energy Act is applicable to all radiological accidents, i.e. accidents pertaining to nuclear facilities and radioactive materials while the Act for Physical Protection and Radiological Emergency ('APPRE') applies to accidents related to nuclear materials and large-scale nuclear facilities. The Atomic Energy Act contains three provisions directly related with radiological accidents (Articles 89, 98 and 102). Article 89 provides for the obligations of nuclear licensees or consigned transporters to institute safety measures and file a report to the head of the Ministry of Education, Science and Technology ('MEST') in the event of any radiological accident during transport or packing of radioactive materials, etc. Article 98 stipulates obligations of nuclear licensees to implement safety procedures and submit a report to the Minister of Education, Science and Technology concerning radiation hazards arising in the event a radiological accident occurs in connection with nuclear projects, as well as the Minister's requests to implement necessary measures. Article 102 explicitly provides for obligations to file a report to the Minister in the event of theft, loss, fire or other accidents involving radioactive materials, etc. in the possession of nuclear licensees. The APPRE classifies radiological accidents according to location and scale of the accidents. Based on location, accidents are divided into accidents inside or outside nuclear facilities. Accidents inside nuclear facilities refer to accidents that occur at nuclear reactors, nuclear fuel cycling facilities, radioactive waste storage, treatment and disposal facilities, facilities using nuclear materials and facilities related to radioisotopes of not lower than 18.5PBq (Subparagraph 2, Article 2 of the APPRE) while accidents outside nuclear facilities mean accidents that take place on

  10. The program system UFOMOD for assessing the consequences of nuclear accidents

    International Nuclear Information System (INIS)

    Ehrhardt, J.; Burkart, K.; Hasemann, I.; Matzerath, C.; Panitz, H.J.; Steinhauer, C.

    1988-10-01

    The programm system UFOMOD is a completely new accident consequence assessment (ACA) code. Its structure and modelling is based on the experience gained from applications of the old UFOMOD code during and after the German Risk Study - Phase A, the results of scientific investigations performed within the ongoing Phase B and the CEC-project MARIA, and the requirements resulting from the extended use of ACAs to help in decision-making. One of the most important improvements is the introduction of different trajecotry models for describing atmospheric dispersion in the near range and at larger distances. Emergency actions and countermeasures modelling takes into account recommendations of international commissions. The dosimetric models contain completely new age-, sex- and time-dependent data of dose-conversion factors for external and internal radiation; the ingestion pathway is modelled to consider seasonal dependencies. New dose-risk-relationships for stochastic and non-stochastic health effects are implemented; a special algorithm developed for ACA codes allows individual and collective leukemia and cancer risks to be presented as a function of time after the accident. According to the modular structure of the new program system UFOMOD, an easy access to parameter values and the results of the various submodels exists what facilitates sensitivity and uncertainty analyses. (orig.) [de

  11. Interfacing systems loss of coolant accident (ISLOCA) pressure capacity methodology and Davis-Besse results

    International Nuclear Information System (INIS)

    Wesley, D.A.

    1991-01-01

    A loss of coolant accident resulting from the overpressurization by reactor coolant fluid of a system designed for low-pressure, low-temperature service has been identified as a potential contributor to nuclear power plant risk. In this paper, the methodology developed to assess the probability of failure as a function of internal pressure is presented, and sample results developed for the controlling failure modes and locations of four fluid systems at the Davis-Besse Plant are shown. Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The variability in the probability of failure is included, and the estimated leak rates or leak areas are given for the controlling modes of failure. For this evaluation, all failures are based on quasistatic pressures since the probability of dynamic effects resulting from such causes as water hammer have been initially judged to be negligible for the Davis-Besse plant ISLOCA

  12. Comparison and verification of two computer programs used to analyze ventilation systems under accident conditions

    International Nuclear Information System (INIS)

    Hartig, S.H.; Wurz, D.E.; Arnitz, T.; Ruedinger, V.

    1985-01-01

    Two computer codes, TVENT and EVENT, which were developed at the Los Alamos National Laboratory (LANL) for the analysis of ventilation systems, have been modified to model air-cleaning systems that include active components with time-dependent flow-resistance characteristics. With both modified programs, fluid-dynamic transients were calculated for a test facility used to simulate accident conditions in air-cleaning systems. Experiments were performed in the test facility whereby flow and pressure transients were generated with the help of two quick-actuating air-stream control valves. The numerical calculations are compared with the test results. Although EVENT makes use of a more complex theoretical flow model than TVENT, the numerical simulations of both codes were found to be very similar for the flow conditions studied and to closely follow the experimental results

  13. Information systems in Chernobyl accident after-effect elimination: On the way from youth to maturity

    International Nuclear Information System (INIS)

    Chabanyuk, V.S.; Proskura, N.I.; Tabachny, L.Ya.

    1997-01-01

    10-years period of Information Systems in the Chernobyl Accident after-effects elimination (these systems we name Chernobyl Information Systems (ChIS) for simplicity of reference) creation is analyzed. It is claimed that ChIS are introducing into the maturity phase now. The paper consists of Introduction, four paragraphs and Conclusion. Short history of ChIS creation on the example of radioecological component is described in Introduction. Two phases: youth and maturity, are identified. The youth phase is divided on three periods: 1986-1988, 1988-1992, 1993-1995. The maturity phase has started in 1994 with accepting of new Conception of ChIS implementation. Main characteristics of each phase and period are described

  14. Fission products transport in CANDU Primary Heat Transport System in a severe accident

    International Nuclear Information System (INIS)

    Constantin, M.; Rizoiu, A.; Turcu, I.; Negut, Gh.

    2005-01-01

    Full text: The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) System by using the ASTEC code (Accident Source Term Evaluation Code). The complexity of the data required by ASTEC and the complexity of CANDU PHT were strong motivation to begin with a simplified geometry in order to avoid the introducing of unmanageable errors at the level of input deck. Thus only 1/4 of the PHT circuit was simulated, an simplified FPs inventory and some simplifications in the feeders geometry were also used. The circuit consists of 95 horizontal fuel channels connected to 95 horizontal out-feeders, then through vertical feeders to the outlet-header (a big pipe that collects the water from feeders); the circuit continues from the outlet-header with a riser and then with the steam generator and a pump. After this pump, the circuit was broken; in this point the FPs are transferred to the containment. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU NPP loss of coolant accident sequence. Temperature and pressure conditions in the time of the accident were calculated by CATHENA code and the source term of FPs introduced into the PHT was estimated by ORIGEN code. The results consist of mass distributions in the nodes of the circuit and the mass transfer to the containment through the break for different species (FPs and chemical species). The study is completed by sensitivity analysis for the parameters with important uncertainties. (authors)

  15. Applications of nano-fluids to enhance LWR accidents management in in-vessel retention and emergency core cooling systems

    International Nuclear Information System (INIS)

    Chupin, A.; Hu, L. W.; Buongiorno, J.

    2008-01-01

    Water-based nano-fluid, colloidal dispersions of nano-particles in water; have been shown experimentally to increase the critical heat flux and surface wettability at very low concentrations. The use of nano-fluids to enhance accidents management would allow either to increase the safe margins in case of severe accidents or to upgrade the power of an existing power plant with constant margins. Building on the initial work, computational fluid dynamics simulations of the nano-fluid injection system have been performed to evaluate the feasibility of a nano-fluid injection system for in-vessel retention application. A preliminary assessment was also conducted on the emergency core cooling system of the European Pressurized Reactor (EPR) to implement a nano-fluid injection system for improving the management of loss of coolant accidents. Several design options were compared/or their respective merits and disadvantages based on criteria including time to injection, safety impact, and materials compatibility. (authors)

  16. Solving the problem of valve stem leakage

    International Nuclear Information System (INIS)

    Dixon, D.F.

    1976-01-01

    Engineering solutions to valve stem leakage, in systems carrying expensive heavy water under pressure, have progressed from changing packing brands (failure) to leak collection (partial success) to elimination of small packed valves and an improved valve packing strategy involving stable packing materials, live Belleville spring-loading of packing, and issuance of a detailed stuffing box specification (success). (E.C.B.)

  17. Diagnostic and prognostic system for identification of accident scenarios and prediction of 'source term' in nuclear power plants under accident conditions

    International Nuclear Information System (INIS)

    Santhosh; Gera, B.; Kumar, Mithilesh

    2014-01-01

    Nuclear power plant experiences a number of transients during its operations. These transients may be due to equipment failure, malfunctioning of process support systems etc. In such a situation, the plant may result in an abnormal state which is undesired. In case of such an undesired plant condition, the operator has to carry out diagnostic and corrective actions. When an event occurs starting from the steady state operation, instruments' readings develop a time dependent pattern and these patterns are unique with respect to the type of the particular event. Therefore, by properly selecting the plant process parameters, the transients can be distinguished. In this connection, a computer based tool known as Diagnostic and Prognostic System has been developed for identification of large pipe break scenarios in 220 MWe Pressurised Heavy Water Reactors (PHWRs) and for prediction of expected 'Source Term' and consequence for a situation where Emergency Core Cooling System (ECCS) is not available or partially available. Diagnostic and Prognostic System is essentially a transient identification and expected source term forecasting system. The system is based on Artificial Neural Networks (ANNs) that continuously monitors the plant conditions and identifies a Loss Of Coolant Accident (LOCA) scenario quickly based on the reactor process parameter values. The system further identifies the availability of injection of ECCS and in case non-availability of ECCS, it can forecast expected 'Source Term'. The system is a support to plant operators as well as for emergency preparedness. The ANN is trained with a process parameter database pertaining to accident conditions and tested against blind exercises. In order to see the feasibility of implementing in the plant for real-time diagnosis, this system has been set up on a high speed computing facility and has been demonstrated successfully for LOCA scenarios. (author)

  18. Particle and radiation leakage importance: definition, analysis, and interpretation

    International Nuclear Information System (INIS)

    Cacuci, D.G.; Wagschal, J.J.; Yaari, A.

    1982-01-01

    The concept of leakage importance function has been introduced and analyzed for physical systems governed by the Boltzmann transport equation. This leakage importance function represents a measure of the relative importance of source particles located at every point in phase space in contributing to the leakage and provides insight regarding the specific physical process that leads to leakage. The equation satisfied by the leakage importance function has been derived by using adjoint operators. It has been shown that procedures that are customarily used to derive an equation obeyed by an importance function suitable for an integral parameter such as a detector response or an eigenvalue lead to difficulties when directly applied to derive an equation obeyed by the leakage importance function. This is because, although leakage is also an integral parameter (i.e., a functional of the forward flux density), leakage is expressed in terms of a surface integral rather than in terms of volume integrals such as those appearing in expressions of detector responses or eigenvalues. Therefore, a procedure that departs from the customary course has been devised to derive the equation satisfied by the leakage importance function

  19. Valve packing leakage monitoring device

    International Nuclear Information System (INIS)

    Ezekoye, L.I.

    1985-01-01

    A device for monitoring leakage of fluid across a seal in a component connected to a pressurized fluid system including a housing having a chamber with an inlet for receiving fluid leaking across the seal and an outlet. A positioning means is connected to an orifice plug so as to move the plug for permitting the fluid to be discharged through the orifice at the same rate at which it enters the first chamber and means for detecting the movement of the plug is provided to produce and output signal corresponding to the distance moved by the plug and thereby indicate flow rate. The positioning means compromise a piston attached to the plug by a hollow tube and springs, which at low flow rates locate the piston. When flow increases sufficiently pressure increases and urges the piston upwards. A magnetic portion of tube actuates a succession of proximity switches to indicate flow rate. (author)

  20. A Novel Thermal-Mechanical Detection System for Reactor Pressure Vessel Bottom Failure Monitoring in Severe Accidents

    International Nuclear Information System (INIS)

    Bi, Daowei; Bu, Jiangtao; Xu, Dongling

    2013-06-01

    Following the Fukushima Daiichi nuclear accident in Japan, there is an increased need of enhanced capabilities for severe accident management (SAM) program. Among others, a reliable method for detecting reactor pressure vessel (RPV) bottom failure has been evaluated as imperative by many utility owners. Though radiation and/or temperature measurement are potential solutions by tradition, there are some limitations for them to function desirably in such severe accident as that in Japan. To provide reliable information for assessment of accident progress in SAM program, in this paper we propose a novel thermal-mechanical detection system (TMDS) for RPV bottom failure monitoring in severe accidents. The main components of TMDS include thermally sensitive element, metallic cables, tension controlled switch and main control room annunciation device. With TMDS installed, there shall be a reliable means of keeping SAM decision-makers informed whether the RPV bottom has indeed failed. Such assurance definitely guarantees enhancement of severe accident management performance and significantly improve nuclear safety and thus protect the society and people. (authors)

  1. Rade-aid a decision support system to evaluate countermeasures after a radiological accident

    International Nuclear Information System (INIS)

    Wagenaar, G.; Van Den Bosch, C.J.H.; Weger, D. de.

    1990-01-01

    After Chernobyl the authorities in many countries were overwhelmed by the enormous amount of information that was being generated by measuring and monitoring programs. In making decisions, this information had to be combined with the results of specific countermeasures, in order to determine the optimal strategy with respect to a large number of consequences. The development of RADE-AID, the Radiological Accident Decision AIDing system, is aimed at providing a powerful tool in the decision-making process. RADE-AID is developed by TNO (The Netherlands) in a joint contract with KfK (FRG) and NRPB (UK). In the first phase a demonstration system will be built, called RADE-AID/D. RADE-AID/D will be used as a decision support system in the intermediate and late phase after a radiological accident. RADE-AID/D will consider countermeasures with respect to external exposure and internal exposure by food ingestion. Countermeasures are evaluated considering reduction in doses and in numbers of health effects, costs, and social effects. The paper covers the structure of the program, presentation of data and results, and the decision analysis technique that is being used. This decision analysis part is an important feature of the system; an advanced decision analysis technique is used, that is able to compare data of varying nature. Furthermore the place of RADE-AID in the decision-making process will be treated. RADE-AID/D is an interactive computer program, that offers the user the possibility to enter relevant data and to have data and results displayed in a variety of ways. Furthermore the system contains an advanced decision analysis technique, that is able to compare data of varying nature. Input data for the decision analysis calculations are provided by models from UFOMOD and MARC-codes

  2. On the scaling of gas leakage from static seals

    International Nuclear Information System (INIS)

    Chivers, T.C.; Hunt, R.P.

    1977-01-01

    The interaction between gas leakage from static seals and eight potential variables is discussed. From a consideration of the interaction of these various parameters and the mechanical design of the seal system the importance of correctly interpreting leakage data is demonstrated. Given a situation where model experiments are necessary, this document forms a basis for the definition and interpretation of a test programme. (author)

  3. Hunting the unknown: White-box database leakage detection

    NARCIS (Netherlands)

    Costante, E.; Hartog, den J.I.; Petkovic, M.; Etalle, S.; Pechenizkiy, M.; Atluri, V.; Pernul, G.

    2014-01-01

    Data leakage causes significant losses and privacy breaches worldwide. In this paper we present a white-box data leakage detection system to spot anomalies in database transactions. We argue that our approach represents a major leap forward w.r.t. previous work because: i) it significantly decreases

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  5. Quality systems for radiotherapy: Impact by a central authority for improved accuracy, safety and accident prevention

    International Nuclear Information System (INIS)

    Jaervinen, H.; Sipilae, P.; Parkkinen, R.; Kosunen, A.; Jokelainen, I.

    2001-01-01

    High accuracy in radiotherapy is required for the good outcome of the treatments, which in turn implies the need to develop comprehensive Quality Systems for the operation of the clinic. The legal requirements as well as the recommendation by professional societies support this modern approach for improved accuracy, safety and accident prevention. The actions of a national radiation protection authority can play an important role in this development. In this paper, the actions of the authority in Finland (STUK) for the control of the implementation of the new requirements are reviewed. It is concluded that the role of the authorities should not be limited to simple control actions, but comprehensive practical support for the development of the Quality Systems should be provided. (author)

  6. NIRS external dose estimation system for Fukushima residents after the Fukushima Dai-ichi NPP accident

    Science.gov (United States)

    Akahane, Keiichi; Yonai, Shunsuke; Fukuda, Shigekazu; Miyahara, Nobuyuki; Yasuda, Hiroshi; Iwaoka, Kazuki; Matsumoto, Masaki; Fukumura, Akifumi; Akashi, Makoto

    2013-04-01

    The great east Japan earthquake and subsequent tsunamis caused Fukushima Dai-ichi Nuclear Power Plant (NPP) accident. National Institute of Radiological Sciences (NIRS) developed the external dose estimation system for Fukushima residents. The system is being used in the Fukushima health management survey. The doses can be obtained by superimposing the behavior data of the residents on the dose rate maps. For grasping the doses, 18 evacuation patterns of the residents were assumed by considering the actual evacuation information before using the survey data. The doses of the residents from the deliberate evacuation area were relatively higher than those from the area within 20 km radius. The estimated doses varied from around 1 to 6 mSv for the residents evacuated from the representative places in the deliberate evacuation area. The maximum dose in 18 evacuation patterns was estimated to be 19 mSv.

  7. Statistical method application to knowledge base building for reactor accident diagnostic system

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Yokobayashi, Masao; Matsumoto, Kiyoshi; Kohsaka, Atsuo

    1989-01-01

    In the development of a knowledge based expert system, one of key issues is how to build the knowledge base (KB) in an efficient way with keeping the objectivity of KB. In order to solve this issue, an approach has been proposed to build a prototype KB systematically by a statistical method, factor analysis. For the verification of this approach, factor analysis was applied to build a prototype KB for the JAERI expert system DISKET. To this end, alarm and process information was generated by a PWR simulator and the factor analysis was applied to this information to define taxonomy of accident hypotheses and to extract rules for each hypothesis. The prototype KB thus built was tested through inferring against several types of transients including double-failures. In each diagnosis, the transient type was well identified. Furthermore, newly introduced standards for rule extraction showed good effects on the enhancement of the performance of prototype KB. (author)

  8. Failure/leakage predictions of concrete structures containing cracks

    International Nuclear Information System (INIS)

    Pan, Y.C.; Marchertas, A.H.; Kennedy, J.M.

    1984-06-01

    An approach is presented for studying the cracking and radioactive release of a reactor containment during severe accidents and extreme environments. The cracking of concrete is modeled as the blunt crack. The initiation and propagation of a crack are determined by using the maximum strength and the J-integral criteria. Furthermore, the extent of cracking is related to the leakage calculation by using a model developed by Rizkalla, Lau and Simmonds. Numerical examples are given for a three-point bending problem and a hypothetical case of a concrete containment structure subjected to high internal pressure during an accident

  9. Safety assessment and improvement of Ignalina NPP against downcomer ruptures outside Accident Localisation System

    International Nuclear Information System (INIS)

    Rimkevicius, S.; Urbonavicius, E.

    2002-01-01

    Accident Localisation System (ALS) of Ignalina NPP is a pressure suppression type confinement, designed to prevent the release of contaminated steam-water mixture to the environment in case of Loss-of-Coolant Accident (LOCA). One of the peculiarities of Ignalina NPP with RBMK-1500 reactors is that not all of the reactor coolant circuit is enclosed within ALS. Some part of downcomers, that connect Drum Separator (DS) and suction header of main circulation pump is located outside ALS. In case of downcomer rupture in DS compartment the discharge is not confined, but flows to the environment through the safety panels installed in the ceiling of DS compartments. Numerous safety analyses were performed to assess the safety of Ignalina NPP against downcomer break outside ALS, and results were used for different applications in order to improve the safety of the plant. This paper presents the overview of the performed analyses, recommendations raised and safety improvements made to enhance the safety level of NPP. One of the applications is to present the recommendations for safety improvement if maximal allowable pressure limits are exceeded. The calculations results demonstrate that in the case of two downcomers rupture in drum separators compartment the maximum permissible pressure in the reactor hall could be exceeded. The knock-out panels from the reactor hall to the environment were recommended and installed for reactor hall overpressure protection. The evaluation of the drainage system efficiency from DS compartments was performed. In this case the especial attention was paid to analyse the water collection and drainage system behaviour in long term after postulated breaks. The analysis results showed that the modernization of the drainage system prevents the accumulation of the released water in the compartments even in the case of two downcomer pipes ruptures, and decreases the release of radioactive fission products (FP) to the environment.(author)

  10. Construction of a technique plan repository and evaluation system based on AHP group decision-making for emergency treatment and disposal in chemical pollution accidents

    International Nuclear Information System (INIS)

    Shi, Shenggang; Cao, Jingcan; Feng, Li; Liang, Wenyan; Zhang, Liqiu

    2014-01-01

    Highlights: • Different chemical pollution accidents were simplified using the event tree analysis. • Emergency disposal technique plan repository of chemicals accidents was constructed. • The technique evaluation index system of chemicals accidents disposal was developed. • A combination of group decision and analytical hierarchy process (AHP) was employed. • Group decision introducing similarity and diversity factor was used for data analysis. - Abstract: The environmental pollution resulting from chemical accidents has caused increasingly serious concerns. Therefore, it is very important to be able to determine in advance the appropriate emergency treatment and disposal technology for different types of chemical accidents. However, the formulation of an emergency plan for chemical pollution accidents is considerably difficult due to the substantial uncertainty and complexity of such accidents. This paper explains how the event tree method was used to create 54 different scenarios for chemical pollution accidents, based on the polluted medium, dangerous characteristics and properties of chemicals involved. For each type of chemical accident, feasible emergency treatment and disposal technology schemes were established, considering the areas of pollution source control, pollutant non-proliferation, contaminant elimination and waste disposal. Meanwhile, in order to obtain the optimum emergency disposal technology schemes as soon as the chemical pollution accident occurs from the plan repository, the technique evaluation index system was developed based on group decision-improved analytical hierarchy process (AHP), and has been tested by using a sudden aniline pollution accident that occurred in a river in December 2012

  11. Construction of a technique plan repository and evaluation system based on AHP group decision-making for emergency treatment and disposal in chemical pollution accidents

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Shenggang [College of Environmental Science and Engineering, Beijing Forestry University, Beijing 100083 (China); College of Chemistry, Baotou Teachers’ College, Baotou 014030 (China); Cao, Jingcan; Feng, Li; Liang, Wenyan [College of Environmental Science and Engineering, Beijing Forestry University, Beijing 100083 (China); Zhang, Liqiu, E-mail: zhangliqiu@163.com [College of Environmental Science and Engineering, Beijing Forestry University, Beijing 100083 (China)

    2014-07-15

    Highlights: • Different chemical pollution accidents were simplified using the event tree analysis. • Emergency disposal technique plan repository of chemicals accidents was constructed. • The technique evaluation index system of chemicals accidents disposal was developed. • A combination of group decision and analytical hierarchy process (AHP) was employed. • Group decision introducing similarity and diversity factor was used for data analysis. - Abstract: The environmental pollution resulting from chemical accidents has caused increasingly serious concerns. Therefore, it is very important to be able to determine in advance the appropriate emergency treatment and disposal technology for different types of chemical accidents. However, the formulation of an emergency plan for chemical pollution accidents is considerably difficult due to the substantial uncertainty and complexity of such accidents. This paper explains how the event tree method was used to create 54 different scenarios for chemical pollution accidents, based on the polluted medium, dangerous characteristics and properties of chemicals involved. For each type of chemical accident, feasible emergency treatment and disposal technology schemes were established, considering the areas of pollution source control, pollutant non-proliferation, contaminant elimination and waste disposal. Meanwhile, in order to obtain the optimum emergency disposal technology schemes as soon as the chemical pollution accident occurs from the plan repository, the technique evaluation index system was developed based on group decision-improved analytical hierarchy process (AHP), and has been tested by using a sudden aniline pollution accident that occurred in a river in December 2012.

  12. A filter system for steam-gas mixture ejections from under a nuclear reactor containment following a severe accident

    International Nuclear Information System (INIS)

    Dulepov, Ju. N.; Sharygin, L. M.; Tretjakov, S. Ja.; Shtin, A.P.; Glushko, V. V.; Babenko, E. A.; Kurakov, Ju. A.

    1997-01-01

    In this paper newly built NPPs obligatory incorporate a containment having a filter system for removing radioactive materials ejections under severe accidents including nuclear fuel melting is described. The system prevents a containment failure and provides ejected radioactive materials decontamination to permissible levels. The physical-chemical and chemical characteristics of Termoxid-58 sorbent (TiO 5 based sorbent) are presented

  13. Development of the assessment of nuclear accident consequences and decision support system in China: status, requirement and recommendations

    International Nuclear Information System (INIS)

    Shi Zhongqi; Wang Xingyu

    2003-01-01

    This paper introduces the status of nuclear accident consequence assessment/development of decision-making support system in China. The basic functions and roles of the consequence assessment/decision-making support system for three levels of nuclear emergency response organization (i.e. national, local offsite and nuclear power plant operator) in China are presented in the paper

  14. Aviation Safety Program: Weather Accident Prevention (WxAP) Development of WxAP System Architecture And Concepts of Operation

    Science.gov (United States)

    Grantier, David

    2003-01-01

    This paper presents viewgraphs on the development of the Weather Accident Prevention (WxAP) System architecture and Concept of Operation (CONOPS) activities. The topics include: 1) Background Information on System Architecture/CONOPS Activity; 2) Activity Work in Progress; and 3) Anticipated By-Products.

  15. Diagnostic performance of a CT-based scoring system for diagnosis of anastomotic leakage after esophagectomy: comparison with subjective CT assessment

    Energy Technology Data Exchange (ETDEWEB)

    Goense, Lucas; Rossum, Peter S.N. van [University Medical Center Utrecht, Department of Surgery, Utrecht (Netherlands); University Medical Center Utrecht, Department of Radiation Oncology, Utrecht (Netherlands); Stassen, Pauline M.C.; Ruurda, Jelle P.; Hillegersberg, Richard van [University Medical Center Utrecht, Department of Surgery, Utrecht (Netherlands); Wessels, Frank J.; Leeuwen, Maarten S. van [University Medical Center Utrecht, Department of Radiology, Utrecht (Netherlands)

    2017-10-15

    To develop a CT-based prediction score for anastomotic leakage after esophagectomy and compare it to subjective CT interpretation. Consecutive patients who underwent a CT scan for a clinical suspicion of anastomotic leakage after esophagectomy with cervical anastomosis between 2003 and 2014 were analyzed. The CT scans were systematically re-evaluated by two radiologists for the presence of specific CT findings and presence of an anastomotic leak. Also, the original CT interpretations were acquired. These results were compared to patients with and without a clinical confirmed leak. Out of 122 patients that underwent CT for a clinical suspicion of anastomotic leakage; 54 had a confirmed leak. In multivariable analysis, anastomotic leakage was associated with mediastinal fluid (OR = 3.4), esophagogastric wall discontinuity (OR = 4.9), mediastinal air (OR = 6.6), and a fistula (OR = 7.2). Based on these criteria, a prediction score was developed resulting in an area-under-the-curve (AUC) of 0.86, sensitivity of 80%, and specificity of 84%. The original interpretation and the systematic subjective CT assessment by two radiologists resulted in AUCs of 0.68 and 0.75 with sensitivities of 52% and 69%, and specificities of 84% and 82%, respectively. This CT-based score may provide improved diagnostic performance for diagnosis of anastomotic leakage after esophagectomy. (orig.)

  16. Learning lessons from accidents with a human and organisational factors perspective: deficiencies and failures of operating experience feedback systems

    International Nuclear Information System (INIS)

    Dechy, N.; Rousseau, J.M.; Jeffroy, F.

    2012-01-01

    This paper aims at reminding the failures of operating experience feedback (OEF) systems through the lessons of accidents and provides a framework for improving the efficiency of OEF processes. The risk is for example to miss lessons from other companies and industrial sectors, or to miss the implementation of adequate corrective actions with the risk to repeat accidents. Most of major accidents have been caused by a learning failure or other organisational factors as a contributing cause among several root causes. Some of the recurring organisational factors are: -) poor recognition of critical components, of critical activities or deficiency in anticipation and detection of errors, -) excessive production pressure, -) deficiency of communication or lack of quality of dialogue, -) Excessive formalism, -) organisational complexity, -) learning deficiencies (OEF, closing feedback loops, lack of listening of whistle-blowers). Some major accidents occurred in the nuclear industry. Although the Three Mile Island accident has multiple causes, in particular, an inappropriate design of the man-machine interface, it is a striking example of the loss of external lessons from incidents. As for Fukushima it is too early to have established evidence on learning failures. The systematic study and organisational analysis of OEF failures in industrial accidents whatever their sector has enabled us to provide a framework for OEF improvements. Five key OEF issues to improve in priority: 1) human and organisational factors analysis of the root causes of the events, 2) listening to the field staff, dissenting voices and whistle-blowers, 3) monitoring of the external events that provide generic lessons, 4) building an alive memory through a culture of accidents with people who become experiences pillars, and 5) the setting of external audit or organisational analysis of the OEF system by independent experts. The paper is followed by the slides of the presentation

  17. Loss-of-Coolant and Loss-of-Flow Accidents in the SEAFP first wall/blanket cooling system

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1995-01-01

    This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

  18. Loss-of-coolant and loss-of-flow accidents in the SEAFP first wall/blanket cooling system

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1994-07-01

    This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

  19. A systemic analysis of patterns of organizational breakdowns in accidents: A case from Helicopter Emergency Medical Service (HEMS) operations

    International Nuclear Information System (INIS)

    Kontogiannis, Tom; Malakis, Stathis

    2012-01-01

    In recent years, many accident models and techniques have shifted their focus from shortfalls in the actions of practitioners to systemic causes in the organization. Accident investigation techniques (e.g., STAMP) have been developed that looked into the flaws of control processes in the organization. Organizational models have looked into general patterns of breakdown related to structural vulnerabilities and gradual degradation of performance. Although some degree of cross-fertilization has been developed between these two trends, safety analysts are left on their own to integrate this gap between control flaws and patterns of organizational breakdown in accident investigation. This article attempts to elaborate the control dynamics of the Systems Theoretic Accident Model and Process (STAMP) technique on the basis of a theoretical model of organizational viability (i.e., the Viable Systems Model). The joint STAMP–VSM framework is applied to an accident from a Helicopter Emergency Medical Service (HEMS) organization to help analysts progress from the analysis of control flaws to the underlying patterns of breakdown. The joint framework may help analysts to rethink the safety organization, model new information loops and constraints, look at the adaptation and steering functions of the organization and finally, develop high leverage interventions. - Highlights: ► This article bridges the gap between two parallel trends in systemic accident models. ► Investigation techniques (i.e., STAMP) have looked into the flaws of safety management processes. ► The literature has highlighted many patterns (or archetypes) of organizational breakdowns. ► The Viable System Model is used with STAMP to link control flaws and organizational breakdowns.

  20. Developement of leakage localization technique by using acoustic signal

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. H.; Jeon, J. H.; Seo, D. H.; Kim, K. W. [KAIST, Daejeon (Korea, Republic of)

    2010-07-15

    The objective of this research is to develop a leakage monitoring system for pipelines or valves in the secondary water system of a nuclear power plant. The system aims to detect the existence of leakage and to estimate the leak location, especially by utilizing the noise generated from the leak. It is safe, precise real-time alert system compared with the previous monitoring methods and tools such as the visual test and the thermal imaging camera. When there exists leakage in the pipeline or valves of nuclear power plant, the noise due to gas flow is radiated through leak region. That is, the secondary water system with leakage generates different noise from the system without leakage. This motivates us to measure and analyze the noise generated from the secondary water system, so as firstly to detect the existence of leakage, and secondly to estimate the leak location by using the noise source identification technique such as beamforming and acoustic holography. Especially the beamforming method models the signal from the noise source to estimate the location of source. Therefore, it is necessary to model the noise due to leakage which is dependent upon parameters. In the process of leak localization, the reflected wave due to interior walls and the measurement noise should be removed for the precise estimation. Therefore, we attempt to characterize the reflected wave and the measurement noise by modeling the interior sound field, thus to remove them and to localize the leak location with high precision

  1. Accident management

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  2. A review of accidents, prevention and mitigation options related to hazardous gases

    International Nuclear Information System (INIS)

    Fthenakis, V.M.

    1993-05-01

    Statistics on industrial accidents are incomplete due to lack of specific criteria on what constitutes a release or accident. In this country, most major industrial accidents were related to explosions and fires of flammable materials, not to releases of chemicals into the environment. The EPA in a study of 6,928 accidental releases of toxic chemicals revealed that accidents at stationary facilities accounted for 75% of the total number of releases, and transportation accidents for the other 25%. About 7% of all reported accidents (468 cases) resulted in 138 deaths and 4,717 injuries ranging from temporary respiratory problems to critical injuries. In-plant accidents accounted for 65% of the casualties. The most efficient strategy to reduce hazards is to choose technologies which do not require the use of large quantities of hazardous gases. For new technologies this approach can be implemented early in development, before large financial resources and efforts are committed to specific options. Once specific materials and options have been selected, strategies to prevent accident initiating events need to be evaluated and implemented. The next step is to implement safety options which suppress a hazard when an accident initiating event occurs. Releases can be prevented or reduced with fail-safe equipment and valves, adequate warning systems and controls to reduce and interrupt gas leakage. If an accident occurs and safety systems fail to contain a hazardous gas release, then engineering control systems will be relied on to reduce/minimize environmental releases. As a final defensive barrier, the prevention of human exposure is needed if a hazardous gas is released, in spite of previous strategies. Prevention of consequences forms the final defensive barrier. Medical facilities close by that can accommodate victims of the worst accident can reduce the consequences of personnel exposure to hazardous gases

  3. Accident loads for a VVER-440/213 containment

    Energy Technology Data Exchange (ETDEWEB)

    Techy, Z. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Lajtha, G. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Taubner, R. [Institute for Electric Power Research (VEIKI), Budapest (Hungary)

    1995-08-01

    Specific features of the VVER-440/213 containment are the subdivided rectangular building and the localization system including the bubbler trays and air traps. Accident loads are calculated for a large break loss of coolant accident (LBLOCA). The maximum pressure and temperature loads are calculated with different codes during the blowdown phase of the LBLOCA. The uncertainty margins of the maximum pressure are given in this case. Sensitivity studies are performed for different leakage rates and hydraulic data of the containment. The effects of the active and passive spray systems on the depressurization are presented in this paper. The maximum pressure loads are also examined in case of degraded conditions of the localization system. (orig.).

  4. Treatment of sodium spills and leakage detection at loop-type fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Foerster, K; Fortmann, M; Lang, H; Moellerfeld, H [Interatom, Bergisch Gladbach (Germany)

    1979-03-01

    Sodium spills are of great importance in the safety analysis for sodium cooled nuclear plants. Large leakages can lead to a depletion of the heat transfer system and cause the loss of cooling of the reactor. Further the hot sodium may attack structural materials. In areas with air atmosphere large amounts of sodium can burn and cause great damages. Therefore the control of large leakages is an indispensable task in design and construction of sodium cooled reactor systems. Because of the typical arrangement of widespread long pipe systems loop type plants are subject to a gradually greater risk of damage than pool type plants. The sodium catching devices of the SNR-300 are described and their function is illustrated as an example for the treatment of large spills. Since the equipment for the control of large amounts of leaking sodium is very expensive, great efforts are made in order to save costs and to decrease safety problems. It is aimed to minimize the probability of such events to a degree that they no longer are to be considered realistic. The advantageous operating conditions and the favourable material properties support this aim. Under the well known keyword 'leak-before-rupture' criterion this task is pursued. Crack growth measurements are made at structural materials under LMFBR conditions, and leakage detecting systems are being developed. Some test results concerning this task are described. Despite the fact that there are good chances to verify the leak-before-rupture criterion it is assumed that certain hypothetical accidents occur, which are to be considered in the design of the reactor plant. The extremely improbable Bethe-Tait-accident (HCDA) is such an event. It would lead to a super spill, that means to the complete depletion of the reactor tank. For the SNR-300 plant a system is provided that is able to catch this super spill and the core melt. This core catcher must withstand the high temperatures and remove the decay heat. The purpose of this

  5. Treatment of sodium spills and leakage detection at loop-type fast reactors

    International Nuclear Information System (INIS)

    Foerster, K.; Fortmann, M.; Lang, H.; Moellerfeld, H.

    1979-01-01

    Sodium spills are of great importance in the safety analysis for sodium cooled nuclear plants. Large leakages can lead to a depletion of the heat transfer system and cause the loss of cooling of the reactor. Further the hot sodium may attack structural materials. In areas with air atmosphere large amounts of sodium can burn and cause great damages. Therefore the control of large leakages is an indispensable task in design and construction of sodium cooled reactor systems. Because of the typical arrangement of widespread long pipe systems loop type plants are subject to a gradually greater risk of damage than pool type plants. The sodium catching devices of the SNR-300 are described and their function is illustrated as an example for the treatment of large spills. Since the equipment for the control of large amounts of leaking sodium is very expensive, great efforts are made in order to save costs and to decrease safety problems. It is aimed to minimize the probability of such events to a degree that they no longer are to be considered realistic. The advantageous operating conditions and the favourable material properties support this aim. Under the well known keyword 'leak-before-rupture' criterion this task is pursued. Crack growth measurements are made at structural materials under LMFBR conditions, and leakage detecting systems are being developed. Some test results concerning this task are described. Despite the fact that there are good chances to verify the leak-before-rupture criterion it is assumed that certain hypothetical accidents occur, which are to be considered in the design of the reactor plant. The extremely improbable Bethe-Tait-accident (HCDA) is such an event. It would lead to a super spill, that means to the complete depletion of the reactor tank. For the SNR-300 plant a system is provided that is able to catch this super spill and the core melt. This core catcher must withstand the high temperatures and remove the decay heat. The purpose of this

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    Science.gov (United States)

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

    2004-01-01

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

  8. Fission products distributions in Candu primary heat transport and Candu containment systems during a severe accident

    International Nuclear Information System (INIS)

    Constantin, Marin; Rizoiu, Andrei

    2005-01-01

    The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) and CANDU Containment Systems by using the ASTEC code (Accident Source Term Evaluation Code). The complexity of the data required by ASTEC and the complexity both of CANDU PHT and Containment System were strong motivations to begin with a simplified geometry in order to avoid the introducing of unmanageable errors at the level of input deck. Thus only 1/4 of the PHT circuit was simulated and a simplified FPs inventory, some simplifications in the feeders geometry and containment were used. The circuit consists of 95 horizontal fuel channels connected to 95 horizontal out-feeders, then through vertical feeders to the outlet-header (a big pipe that collects the water from feeders); the circuit continues from the outlet-header with a riser and then with the steam generator and a pump. After this pump, the circuit was broken; in this point the FPs are transferred to the containment. The containment model consists of 4 rooms connected between by 6 links. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU NPP loss of coolant accident sequence. Temperature and pressure conditions in the time of the accident were calculated by the CATHENA code and the source term of FPs introduced into the PHT was estimated by the ORIGEN code. The FPs distribution in the nodes of the circuit and the FPs mass transfer per isotope and chemical species are obtained by using SOPHAEROS module of ASTEC code. The distributions into the containment are obtained by the CPA module of ASTEC code (thermalhydraulics calculations in the containment and FPs aerosol transport). The results consist of mass distributions in the nodes of the circuit and the transferred mass to the containment through the break for different species (FPs and chemical species) and mass distributions in the different parts and

  9. Unavoidable Accident

    OpenAIRE

    Grady, Mark F.

    2009-01-01

    In negligence law, "unavoidable accident" is the risk that remains when an actor has used due care. The counterpart of unavoidable accident is "negligent harm." Negligence law makes parties immune for unavoidable accident even when they have used less than due care. Courts have developed a number of methods by which they "sort" accidents to unavoidable accident or to negligent harm, holding parties liable only for the latter. These sorting techniques are interesting in their own right and als...

  10. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, R.; Harrell, J.

    1996-12-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

  11. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    International Nuclear Information System (INIS)

    Fuller, R.; Harrell, J.

    1996-01-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves

  12. Improved atmospheric dispersion modelling in the new program system UFOMOD for accident consequence assessments

    International Nuclear Information System (INIS)

    Panitz, H.J.

    1988-01-01

    An essential aim of the improvements of the new program system UFOMOD for Accident Consequence Assessments (ACAs) was to substitute the straightline Gaussian plume model conventionally used in ACA models by more realistic atmospheric dispersion models. To identify improved models which can be applied in ACA codes and to quantify the implications of different concepts of dispersion modelling on the results of an ACA, probabilistic comparative calculations with different atmospheric dispersion models have been carried out. The study showed that there are trajectory models available which can be applied in ACAs and that these trajectory models provide more realistic results of ACAs than straight-line Gaussian models. This led to a completly novel concept of atmospheric dispersion modelling which distinguish between two different distance ranges of validity: the near range ( 50 km). The two ranges are assigned to respective trajectory models

  13. Analysis of risk reduction methods for interfacing system LOCAs [loss-of-coolant accidents] at PWRs

    International Nuclear Information System (INIS)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1988-01-01

    The Reactor Safety Study (WASH-1400) predicted that Interfacing System Loss-of-Coolant Accidents (ISL) events were significant contributors to risk even though they were calculated to be relatively low frequency events. However, there are substantial uncertainties involved in determining the probability and consequences of the ISL sequences. For example, the assumed valve failure modes, common cause contributions and the location of the break/leak are all uncertain and can significantly influence the predicted risk from ISL events. In order to provide more realistic estimates for the core damage frequencies (CDFs) and a reduction in the magnitude of the uncertainties, a reexamination of ISL scenarios at PWRs has been performed by Brookhaven National Laboratory. The objective of this study was to investigate the vulnerability of pressurized water reactor designs to ISLs and identify any improvements that could significantly reduce the frequency/risk of these events

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  15. Study on radioactive fallout from Fukushima nuclear accident by plant samples using an imaging plate system

    International Nuclear Information System (INIS)

    Minowa, Haruka

    2011-01-01

    The radioactive fallout from the Fukushima nuclear accident was investigated by the radiation images of plant samples using an Imaging Plate System. Plant samples exposed by an imaging plate BASIII 2040 (Fujifilm, Japan) in overnight to one week, and radiation images were read by Typhoon FLA7000 (GE Healthcare Japan Corp.). Identifying and quantitative analysis of radionuclides were measured by Auto Well Gamma System ARC-380CL (Aloha Co. Ltd., Tokyo, Japan). In the cross-sectional images of the bamboo shoot, the radioactive material is shown in heterogeneous distribution, it was found that it concentrated on the tip of the edible portion, and thin skin. These radionuclides were identified as "1"3"7Cs, "1"3"4Cs, and "4"0K. "4"0K is a natural radionuclide, on the other hand "1"3"7Cs and "1"3"4Cs would be derived from the accident of the Fukushima Daiichi Nuclear Power Plant. A high concentration of "1"3"4Cs was shown at the distance of 150 mm from the base of the bamboo shoot by cross-sectional cutting into the width of about 1 mm. It was estimated about 1 kBq of "1"3"4Cs would be included in about 400 g (wet weight) of this one bamboo shoot in an edible part. Imaging data suggests that the contamination of radioactive cesium in this bamboo shoot was caused not by the extraneous attachment but by the absorption from roots. Because bamboo is gather water from extensive area, bamboo shoot concentrates the radioactive material contained in the rain even at low concentrations of radioactive materials in soil. (author)

  16. 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.

  17. Leakage monitoring device and method

    International Nuclear Information System (INIS)

    Yamada, Izumi; Matsui, Yuji; Fujimori, Haruo.

    1995-01-01

    In a water leakage monitor for a steam generator, output signals from an acoustic sensor disposed in the vicinity of a region to be monitored is subjected to phasing calculation (beam forming calculation) to determine the distribution of a sound source intensity distribution. A peak is retrieved based on the distribution of the sound source intensity distribution. A correction coefficient depending on the position of the peak is multiplied to the sound source intensity. The presence or absence of leakage is determined based on the degree of the sound source intensity after the completion of correction. Namely, a relative value of sound source intensity for each of the portions in the region to be monitored is determined, and the point of the greatest sound source intensity is assumed as a leaking point, to determine the position of the leakage. An absolute value of the sound source intensity at the leaking point is determined by such a constitution that a correction coefficient depending on the position is multiplied to the intensity of the position of the peak in the distribution of the sound intensity. A threshold value for the determination of the presence or absence of the leakage can be set if a relation between an amount of the leakage previously determined experimentally and the intensity of the sound source. Then, a countermeasure can easily be taken after the detection of the leakage and a restoring operation can be carried out rapidly after the occurrence of leakage while avoiding unnecessary shutdown. (N.H.)

  18. MDEP Common Position CP-EPRWG-04. Common position on EPR containment heat removal system in accident conditions

    International Nuclear Information System (INIS)

    2015-01-01

    The importance of the integrity of the containment as a fundamental barrier to protect the people and environment against the effects of a nuclear accident is well established. In this regard, an essential objective is that the necessity for off-site counter-measures to reduce radiological consequences be limited or even eliminated. The design should provide engineering means to address those sequences which would otherwise lead to large or early releases, even in case of severe external hazards. The plant shall be designed so that it can be brought into a controlled and stable state and the containment function can be maintained, under accident conditions in which there is a significant amount of radioactive material in the containment, i.e. resulting from severe degradation of the reactor core. It is expected that due consideration to these requirements is to be given while tailoring long term loss of electrical power mitigation strategies. In order to reliably maintain the containment barrier, the regulators believe that: - safety features specifically designed for fulfilling safety functions required in core melt accidents shall be independent to the extent reasonably practicable from the Systems, Structures and Components (SSC) of the other levels of defense; - safety features specifically designed for fulfilling safety functions required in core melt accidents shall be safety classified and adequately qualified for the core melt accident environmental conditions for the time frame for which they are required to operate. In the light of the Fukushima Daiichi accident, the regulators believe that those safety features shall be designed with an adequate margin as compared to the levels of natural hazards considered for the site hazard evaluation; - the systems and components necessary for ensuring the containment function in a core melt accident shall have reliability commensurate with the function that they are required to fulfil. This may require redundancy of

  19. Distributed public key schemes secure against continual leakage

    DEFF Research Database (Denmark)

    Akavia, Adi; Goldwasser, Shafi; Hazay, Carmit

    2012-01-01

    -secure against continual memory leakage. Our DPKE scheme also implies a secure storage system on leaky devices, where a value s can be secretely stored on devices that continually leak information about their internal state to an external attacker. The devices go through a periodic refresh protocol......In this work we study distributed public key schemes secure against continual memory leakage. The secret key will be shared among two computing devices communicating over a public channel, and the decryption operation will be computed by a simple 2-party protocol between the devices. Similarly...... against continual memory leakage, under the Bilinear Decisional Diffie-Hellman and $2$-linear assumptions. Our schemes have the following properties: 1. Our DPKE and DIBE schemes tolerate leakage at all times, including during refresh. During refresh the tolerated leakage is a (1/2-o (1),1)-fraction...

  20. Determination of Unidentified Leakage Using a Kalman Smoother

    International Nuclear Information System (INIS)

    Jang, Soek Bo; Heo, Gyunyoung; Ra, Insik; Han, Jeonghyun; Lee, Seon Woo

    2008-01-01

    Since the safety significance of leaks from the RCS can widely vary depending on the source of the leak as well as the leak rate, the detection of the leakage is an important issue. The leakage is classified into 1) identified leakage which is defined as leakage into closed systems such as pump seal or valve packing leaks that can be captured, and 2) unidentified leakage which is all other leakage. The unidentified leakage is typically determined by the RCS inventory balance method which is based on NUREG-1107. Since the accuracy of leak rate calculation is dependent of the plant operating condition, the change in the RCS temperature, inventory, and the transient operating condition should be avoided during the measurement period. Nevertheless, the operation of the makeup of the borated water into the RCS and the diversion of the inventory to the outside of the RCS boundary makes it difficult to maintain the plant stable over an hour. Due to the large fluctuatio