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

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.

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.

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

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.)

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.

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.

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

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

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.)

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

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

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.)

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.

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.

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

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)

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)

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.

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.

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

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

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

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

Leakage monitoring and maintenance for main heat-exchanger of project 871

International Nuclear Information System (INIS)

Li Yanshui; Cao Zhibin

1995-01-01

Main heat-exchanger is one of the most important equipment. To prevent the occurrence of leakage accident, desalting and water treatment for secondary cooling water is conducted chemical except controlling the water quality of coolant. There is a small loop for leakage monitoring to find out the leakage immediately. In order to ensure the safety of operation, the treatment of eliminating water dirt and inspection for outside and inside must be done periodically, and the relevant means must be adopted if any abnormal phenomenon is found

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.

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

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.).

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

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

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

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

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.

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%

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

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)

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

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.

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

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...

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....

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.)

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)

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

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...

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

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.

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].

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

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

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)

Core failure accident pathways and ways to control it

International Nuclear Information System (INIS)

Mayinger, F.

1982-01-01

In the German Risk Study accidents are assumed to result in core meltdown whenever the criteria spelt out in the guidelines of the Advisory Committee on Reactor Safeguards are no longer met. This assumption must be seen in the light of an earlier state of the art in which no detailed information could be obtained about intermediate stages in emergency core cooling systems working according to permit up to the complete failure of all heat removal systems. However, experimental studies and theoretical analyses conducted over the past few years have advanced the state of the art such that it is now possible to predict with considerably more physical reality the behavior of a core in a loss-of-coolant accident. These findings are not only based on calculations, but also on the results of experiments in large facilities allowing direct comparisons to be made with conditions in nuclear power plants. Studies of the effects of systems failures both in major leakages and in the small leakages regarded to be much more dangerous show much more favorable conditions with respect to core coolability than had to be anticipated on the basis of earlier assumptions. This also implies that it would neither be necessary nor meaningful to reinforce emergency core cooling systems. Instead, it is much more important, besides having technically highly qualified and thoroughly trained operating crews, to inform those crews reliably of the hydrodynamic and thermodynamic state of the primary system, especially the core. (orig.) [de

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.

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

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

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)

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

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.

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.

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

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.)

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.

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.)

Radiation doses in accidents at sea-transportation of spent fuel

International Nuclear Information System (INIS)

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

1978-01-01

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

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.

Perspectives on phenomenology and simulation of severe accident in light water reactors

International Nuclear Information System (INIS)

Sugimoto, Jun

2014-01-01

Severe accident phenomena in light water reactors (LWRs) are generally characterized by their physically and chemically complex processes involved with high temperature core melt, multi-component and multi-phase flows, transport of radioactive materials and sometimes highly non-equilibrium state. Severe accident phenomenology is usually categorized into four phases; (1) fuel degradation, (2) in-vessel phenomena, (3) ex-vessel phenomena and (4) fission product release and transport. Among these, ex-vessel phenomena consist of five subcategories; 1) direct containment heating, 2) fuel coolant interaction (steam explosion), 3) molten core concrete interaction, 4) hydrogen behaviour and control and 5) containment failure/leakage. In the field of simulation of severe accident, severe accident analytical codes have been developed in the United States, EU and Japan, such as MAAP, MELCOR, ASTEC, THALES and SAMPSON. Many different kinds of analytical codes for the specific severe accident phenomena have also been developed worldwide. After the accident at Fukushima Daiichi Nuclear Power Station, review of severe accident research issues has been conducted and several issues are reconsidered, such as effects of BWR core degradation behaviors, sea water injection, pool scrubbing under rapid depressurization, containment failure/leakage and re-criticality. Some new experimental and analytical efforts have been started after the Fukushima accident. The present paper describes the perspectives on phenomenology and simulation of severe accident in LWRs, with the emphasis of insights obtained in the review of Fukushima accident. (author)

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

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

Analyzing containment leakage from a sodium fire by the response surface method

International Nuclear Information System (INIS)

Person, L.W.

1978-01-01

The SPOOL-FIRE code has been used with the response surface method and a Monte Carlo simulation to study sodium fire accidents. The study provides a simple method of estimating the radioactive release via containment leakage; the sensitivity of the output consequences to the variations in the input parameters is also presented

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)

Waste Evaporator Accident Simulation Using RELAP5 Computer Code

International Nuclear Information System (INIS)

POLIZZI, L.M.

2004-01-01

An evaporator is used on liquid waste from processing facilities to reduce the volume of the waste through heating the waste and allowing some of the water to be separated from the waste through boiling. This separation process allows for more efficient processing and storage of liquid waste. Commonly, the liquid waste consists of an aqueous solution of chemicals that over time could induce corrosion, and in turn weaken the tubes in the steam tube bundle of the waste evaporator that are used to heat the waste. This chemically induced corrosion could escalate into a possible tube leakage and/or the severance of a tube(s) in the tube bundle. In this paper, analyses of a waste evaporator system for the processing of liquid waste containing corrosive chemicals are presented to assess the system response to this accident scenario. This accident scenario is evaluated since its consequences can propagate to a release of hazardous material to the outside environment. It is therefore important to ensure that the evaporator system component structural integrity is not compromised, i.e. the design pressure and temperature of the system is not exceeded during the accident transient. The computer code used for the accident simulation is RELAP5-MOD31. The accident scenario analyzed includes a double-ended guillotine break of a tube in the tube bundle of the evaporator. A mitigated scenario is presented to evaluate the excursion of the peak pressure and temperature in the various components of the evaporator system to assess whether the protective actions and controls available are adequate to ensure that the structural integrity of the evaporator system is maintained and that no atmospheric release occurs

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

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)

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.

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...

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)

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.

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

Spent nuclear fuel transportation casks evaluation for water in-leakage

International Nuclear Information System (INIS)

Shah, M.J.; Huang, D.T.; Guttmann, J.; Klymyshyn, N.A.; Koeppel, B.J.; Adkins, H.E.

2004-01-01

The United States Nuclear Regulatory Commission (USNRC) is responsible for licensing commercial spent fuel storage and transportation systems. To ensure that the regulations are risk-informed, and do not place unnecessary regulatory burden on the industry, the USNRC has been examining its regulations that apply to spent fuel transportation casks for maintaining sub-criticality under hypothetical accident conditions. Code of Federal Regulations, Title 10, Part 71[1] (10 CFR 71), section 71.55(b) requires that, for evaluation of sub-criticality for fissile material packages, water moderation should be assumed to occur to the most reactive credible extent consistent with the chemical and physical form of the contents. This requirement is based on a defense-in-depth policy, and accounts for any possibility of water intrusion into the package. This program is designed to quantify the margins of safety of certified transportation casks to water intrusion following hypothetical accident conditions. This paper describes the current status of analytical work being performed to evaluate two USNRC-certified spent fuel transportation casks, HI-STAR 100[2] and TN-68[3]. The analytical work is performed using the ANSYS registered [4] and LS-DYNA trademark [5] finite element analysis (FEA) codes. The models are sufficiently detailed in the areas of bolt closure interfaces and containment boundaries to evaluate the likelihood water in-leakage under free drop hypothetical accident conditions of 10 CFR 71.73

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

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.

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

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

OpenAIRE

Shan, Xian; Liu, Kang; Sun, Pei-Liang

2017-01-01

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...

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

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

Revisiting Ulchin 4 SGTR Accident - Analysis for EOP Improvement

Energy Technology Data Exchange (ETDEWEB)

Lee, Eun-Hye; Lee, Wook-Jo; Jerng, Dong-Wook [Chung-Ang University, Seoul (Korea, Republic of)

2016-10-15

The Steam Generator Tube Ruputure (SGTR) is an accident that U-tube inside the SG is defected so that the reactor coolant releases through broken U-tube and this is one of design basis accidents. Operating the Nuclear Power Plants (NPP), maintaing the integrity of core and preventing radiation release are most important things. Because of risks, many researchers have studied scenarios, impacts and the ways to mitigate SGTR accidents. The study to provide an experimental database of aerosol particle retention and to develop models to support accident management interventions during SGTR was performed. The scaled-down models of NPP were used for experiments, also, MELCOR and SCDAP/RELAP5 were used to simulate a design basis SGTR accident. This study had a major role to resolve uncertainties of various physical models for aerosol mechanical resuspension. The other study which analyzed SGTR accident for System-integrated Modular Advanced Reactor (SMART) was performed. In this analysis, the amount of break flow was focused and TASS/SMRS code was used. It assumed that maximum leak was generated, and found that high RCS pressure, low core inlet coolant temperature, and low break location of the SG cassette contributed to leakage. Although the leakage was large, there was no direct release to atmosphere because the pressure of secondary loop was maintained below the safety relief valve set point. In this analysis, comparison of mitigating procedure when SGTR occurs between shutdown condition and full power condition was performed. In shutdown condition, the core uncovery would not take place in 16 hours whether the cooling procedures are performed or not. Therefore, the integrated amount of break flow should be considered only. In this point of view, cooling through intact SG only, case 3, is the best way to minimize the amount of break flow. In full power condition, the core water level is changed due to high reactor power. The important thing to protect NPP is to keep

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

Potential for containment leak paths through electrical penetration assemblies under severe accident conditions

International Nuclear Information System (INIS)

Sebrell, W.

1983-07-01

The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350 0 F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage

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

Safety during sea transport of radioactive materials. Probabilistic safety analysis of package fro sea surface fire accident

International Nuclear Information System (INIS)

Matsuoka, Takeshi; Obara, Isonori; Akutsu, Yukio; Aritomi, Masanori

2000-01-01

The ships carrying irradiated nuclear fuel, plutonium and high level radioactive wastes(INF materials) are designed to keep integrity of packaging based on the various safety and fireproof measures, even if the ship encounters a maritime fire accident. However, granted that the frequency is very low, realistic severe accidents should be evaluated. In this paper, probabilistic safety assessment method is applied to evaluate safety margin for severe sea fire accidents using event tree analysis. Based on our separate studies, the severest scenario was estimated as follows; an INF transport ship collides with oil tanker and induces a sea surface fire. Probability data such as ship's collision, oil leakage, ignition, escape from fire region, operations of cask cooling system and water flooding systems were also introduced from above mentioned studies. The results indicate that the probability of which packages cannot keep their integrity during the sea surface fire accident is very low and sea transport of INF materials is carried out very safely. (author)

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.

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

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)

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).

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)

Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (5) Identification of dominant factors in ex-vessel accident sequences

International Nuclear Information System (INIS)

Ohno, Shuji; Seino, Hiroshi; Miyahara, Shinya

2009-01-01

The evaluation of accident progression outside of a reactor vessel (ex-vessel) and subsequent transfer behavior of radioactive materials is of great importance from the viewpoint of Level 2 PSA. Hence typical ex-vessel accident sequences in the JAEA Sodium-cooled Fast Reactor are qualitatively discussed in this paper and dominant behaviors or factors in the sequences are investigated through parametric calculations using the CONTAIN/LMR code. Scenarios to be focused on are, 1) sodium vapor leakage from the reactor vessel and 2) sodium-concrete reaction, which are both to be considered in the accident category of LOHRS (loss of heat removal system) and might be followed by an early containment failure due to the thermal effect of sodium combustion and hydrogen burning respectively. The calculated results clarify that the sodium vapor leak rate and the scale of sodium-concrete reaction are the important factors to dominate the ex-vessel accident progression. In addition to the understandings of the dominant factors, the analyzed results also provide the specific information such as pressure loading value to the containment and the timing of pressurization, which is indispensable as technical base in Level 2 PSA for developing event trees and for quantifying the accident consequences. (author)

[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.

Study on confinement function of reactor containment during late phase severe accident

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

During a severe accident reactor containment integrity is maintained by accident management. However, gas leakage from containment is inevitable after the severe accident. A large amount of hydrogen and rare gases are produced due to core damage or melting. These non-condensable gases cause the containment pressure much higher than atmospheric pressure even after residual heat removal system recovery especially for BWR with smaller containment volume. Besides, iodine confined in water pool is re-evaporated under radiation field. The present study consists of realistic evaluation of fission products source term inside containment, quantitative evaluation of iodine re-evaporation effect and the experimental study of hydrogen treatment in BWR using ammonia production method by catalyst. Activities in fiscal year 2012 are that modification of MELCOR fission product chemical model was done and verified by experimental data, and that effects of CsI on ammonia production rate for Ru catalyst were conducted. (author)

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.

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)

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.

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.

An international co-ordinated research programme on nuclear accident dosimetry

International Nuclear Information System (INIS)

Flakus, F.N.

1977-01-01

Where fissile materials are being processed in quantities exceeding the minimum critical amounts, a radiation risk to workers arises from the possibility of criticality excursions. Despite the fact that techniques for preventing the occurende of such accidental excursions have reached very high standards it is generally agreed that the availability of suitable nuclear accident dosimetry (NAD) systems is very important. Following the recommendations of an Advisory Group meeting on NAD, the IAEA had established in 1969 an international coordinated research programme on NAD systems and elaborating standarized systems. A large number of research groups from 14 Member States throughout the world participated in this co-ordinated work. Since 1970 four international multilaboratory intercomparison experiments on NAD have been organized and the response of a variety of dosimeters examined in different neutron spectra under simulated accident conditions at Valduc (France), Oak Ridge (USA), Vinca (Yugoslavia) and Harwell (UK). The results achieved in these intercomparison studies show that NAD systems have been substantially improved and that several systems are available now in a number of laboratories throughout the world that perform within the criteria laid down by the initiating advisory group in 1969. A compendium of neutron leakage spectra has also been elaborated for facilitating the determination of dose from readings of detectors exposed to various neutron fields in criticality accidents

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.

Method of detecting water leakage in radioactive waste containing vessel

International Nuclear Information System (INIS)

Ishioka, Hitoshi; Takao, Yoshiaki; Hayakawa, Kiyoshige.

1989-01-01

Lower level radioactive wastes formed upon operation of nuclear facilities are processed by underground storage. In this case, a plurality of drum cans packed with radioactive wastes are contained in a vessel and a water soluble dye material is placed at the inside of the vessel. The method of placing the water soluble dye material at the inside of the vessel includes a method of coating the material on the inner surface of the vessel and a method of mixing the material in sands to be filled between each of the drum cans. Then, leakage of water soluble dye material is detected when water intruding from the outside into the vessel is again leached out of the vessel, to detect the water leakage from the inside of the vessel. In this way, it is possible to find a water-invaded vessel before corrosion of the drum can by water intruded into the vessel and leakage of nuclides in the drum can. Accordingly, it is possible to apply treatment such as repair before occurrence of accident and can maintain the safety of radioactive water processing facilities. (I.S.)

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.

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

Severe accident tests and development of domestic severe accident system codes

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

Severe accident tests and development of domestic severe accident system codes

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)

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)

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

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...

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

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)

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.

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

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...

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.

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

Air-steam leakage through cracks in concrete

International Nuclear Information System (INIS)

Georges Nahas; Helene Simon

2005-01-01

Full text of publication follows: In the context of a severe accident in a Pressurised Water nuclear plant, the evaluation of the leakage rate through the containment wall remains a key point of the safety analysis, because it influences directly the consequences on the environment. During a severe accident, large amounts of steam could be released in the containment; internal pressure could rise beyond design limits causing cracks to appear in the internal concrete wall of the double-wall containment and fission products to leak towards the containment annulus. A research program led by the French Institute for Radiological Protection and Nuclear Safety aims to estimate this leakage. In the presence of cracks, most of the leak flows through them. Hence, a first phase of the program was to build a two-phase homogeneous model for the flow of an air-steam mixture through a idealized traversing crack, taking into account condensation phenomena, and considering crack openings from 25 μm to several hundred μm. A numerical model for the flow, coupled with heat transfer in the wall, was implemented in the Finite Element code CAST3M. This model was validated on a small scaled experiment which was made of two parallel glass plates. Comparison of the numerical and experimental results in this 'channel case' has shown good results for the total mass flow rate for channel openings greater than 100 μm. For the 50 μm opening the calculation gave a 50 % estimate of the experimental total mass flow rate. The second phase of the program is now to validate the model on cracks performed in a concrete specimen. In order to do so, we have simulated the experiment VK2/2 described in the article named 'Investigation of the leakage behavior of reinforced concrete walls' by N. Herrmann, C. Niklasch, M. Stegemann, L. Stempniewski. The reinforced concrete slab, 2.7 m long in the reinforcement direction and 1.2 m thick in the cracking direction, is placed in a mechanical set-up and an

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

Potential for containment leak paths through electrical penetration assemblies under severe accident conditions. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Sebrell, W.

1983-07-01

The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident sequence analysis conducted on a Boiling Water Reactor (BWR), Mark I (MK I), indicated very high temperatures in the dry-well region, which is the location of the majority of electrical penetration assemblies. Because of the high temperatures, it was postulated in the ORNL study that the sealants would fail and all the electrical penetration assemblies would leak before structural failure would occur. Since other containments had similar electrical penetration assemblies, it was concluded that all containments would experience the same type of failure. The results of this study, however, show that this conclusion does not hold for PWRs because in the worst accident sequence, the long time containment gases stabilize to 350/sup 0/F. BWRs, on the other hand, do experience high dry-well temperatures and have a higher potential for leakage.

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.

Nuclear accidents and safety measures of domestic nuclear power plants

International Nuclear Information System (INIS)

Song Zurong; Che Shuwei; Pan Xiang

2012-01-01

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

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)

A concept of countermeasure against radioactive wastewater generated in disastrous nuclear accident such as Fukushima Daiichi site case

Energy Technology Data Exchange (ETDEWEB)

Kim, Kwang-Wook; Lee, Keun-Young; Lee, Eil-Hee; Baek, Yeji; So, Ji-Yang; Chung, Dong-Young; Moon, Jei-Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

Before the operation of initial wastewater treatment systems supplied by AREVA and Kurion companies, which were installed about 6 months after the accident, the contaminated water was accumulated in reactor and turbine buildings, then was moved and stored in many hurriedly-prepared storage tanks including even mega float barge. The wastewater treatment systems using Cs-adsorption columns and desalination equipment was not properly operated and there were several small and big leakages of contaminated water from the wastewater treatment system and storage tanks, so that tremendous wastewater had been accumulated during those periods. That thereafter led to many secondary problems in management and treatment of the wastewater. Since the disastrous accident at Fukushima, several measures to more enhance safety of nuclear power plants located on coastal area have been asked. As one of them, a countermeasure against generation of tremendous radioactive wastewater in disastrous nuclear accident like the Fukushima Daiichi station was asked to be prepared.

Effect of RCIC Operating Conditions on the Accident Scenario in Fukushima Unit 2

International Nuclear Information System (INIS)

Kim, Sung Il; Park, Jong Hwa; Ha, Kwang Soon

2015-01-01

This study was conducted by using MELCOR 1.8.6. Fukushima unit 2 accident was analyzed using MELCOR in this study, and best estimate scenario with considering RCIC operating conditions was presented. Researches on the boiling water reactor (BWR) plant with reactor core isolation cooling (RCIC) system have been conducted. Research on the RCIC operation in Fukushima unit 2 was also conducted by Sandia National Laboratory. MELCOR analysis of the Fukushima unit 2 accident was conducted in the report and energy balance in wetwell was described by considering RCIC operation. However, the effect of RCIC operation condition on the accident scenario has not been studied. The operating conditions of RCIC system affect the pressures in wetwell and drywell, and the high pressure can make leakage path of fission product from PCV to reactor building. Thus it can be directly related with the amount of fission product which released to environment. In this study, severe accident on Fukushima unit 2 was analyzed considering the operating condition of RCIC system, and best estimated scenario was presented. In addition, the effect of RCIC turbine efficiency on the accident progression was examined. Energy balance in suppression chamber was also considered with discussion on the effect of torus room flooding level. It was found that the operating condition of RCIC turbine not only affects the variation of drywell pressure but also the amount of released fission products to environment. It was also confirmed that the RCIC turbine efficiency in the accident would be less than normal operating condition

Analytical methods of leakage rate estimation from a containment under a LOCA

International Nuclear Information System (INIS)

Chun, M.H.

1981-01-01

Three most outstanding maximum flow rate formulas are identified from many existing models. Outlines of the three limiting mass flow rate models are given along with computational procedures to estimate approximate amount of fission products released from a containment to environment for a given characteristic hole size for containment-isolation failure and containment pressure and temperature under a loss of coolant accident. Sample calculations are performed using the critical ideal gas flow rate model and the Moody's graphs for the maximum two-phase flow rates, and the results are compared with the values obtained from then mass leakage rate formula of CONTEMPT-LT code for converging nozzle and sonic flow. It is shown that the critical ideal gas flow rate formula gives almost comparable results as one can obtain from the Moody's model. It is also found that a more conservative approach to estimate leakage rate from a containment under a LOCA is to use the maximum ideal gas flow rate equation rather than the mass leakage rate formula of CONTEMPT-LT. (author)

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)

Analysis of SBO accident for a swimming pool reactor

International Nuclear Information System (INIS)

Wang Guimin; Li Weiwei; Li Ning; Guo Wenhui

2015-01-01

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

[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.

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....

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

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...

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

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...

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,

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.)

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

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.

Detection device for pipeway water leakage in building

International Nuclear Information System (INIS)

Hanawa, Jun.

1988-01-01

Purpose: To rapidly detect pipeway leakage at predetermined areas over a wide range in a building. Constitution: If flooding should occur in a power plant building and left as it is, emergency core cooling system, as well as auxiliary equipments and electrical equipments of the system are flooded to make the safety shutdown of the plant impossible. The present invention copes with such a risk. That is, an inlet flow meter and as exit flow meter are disposed to the inlet and the exit of pipeways disposed in a predetermined region in the building and a flow rate difference detector between them is disposed. In this way, pipeway leakage is detected by detecting the flow rate difference between the inlet flow rate and the exit flow rate of the pipeway in the predetermined region. According to the present invention, if a pipeway in a predetermined region is raptured to cause water leakage, the pipeway leakage can rapidly be detected depending on the flow rate difference between the inlet flow rate and the exit flow rate. Further, the water leakage over the entire the predetermined region can be detected rapidly as compared with the conventional case of detecting the leakage at a restricted portion where the leakage detector is disposed. (Kamimura, M.)

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

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

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)

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)

Forest Carbon Leakage Quantification Methods and Their Suitability for Assessing Leakage in REDD

Directory of Open Access Journals (Sweden)

Sabine Henders

2012-01-01

Full Text Available This paper assesses quantification methods for carbon leakage from forestry activities for their suitability in leakage accounting in a future Reducing Emissions from Deforestation and Forest Degradation (REDD mechanism. To that end, we first conducted a literature review to identify specific pre-requisites for leakage assessment in REDD. We then analyzed a total of 34 quantification methods for leakage emissions from the Clean Development Mechanism (CDM, the Verified Carbon Standard (VCS, the Climate Action Reserve (CAR, the CarbonFix Standard (CFS, and from scientific literature sources. We screened these methods for the leakage aspects they address in terms of leakage type, tools used for quantification and the geographical scale covered. Results show that leakage methods can be grouped into nine main methodological approaches, six of which could fulfill the recommended REDD leakage requirements if approaches for primary and secondary leakage are combined. The majority of methods assessed, address either primary or secondary leakage; the former mostly on a local or regional and the latter on national scale. The VCS is found to be the only carbon accounting standard at present to fulfill all leakage quantification requisites in REDD. However, a lack of accounting methods was identified for international leakage, which was addressed by only two methods, both from scientific literature.

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

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

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)

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.

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.)

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

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

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

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.

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...

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)

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.

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

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...

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.

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)

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.

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)

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.

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

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...

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

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

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

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.)

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...

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)

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

Evaluation of special safety features of the SNR-300 in view of the Chernobyl accident

International Nuclear Information System (INIS)

Vossebrecker, H.

1987-03-01

A comparison of those characteristics, which decisively influenced the accident in the RMBK-1000 reactor, with the safety features of SNR-300 has been performed. The conclusions of this comparison are presented in the present report. The SNR-300 is characterized by a stable reactivity behaviour and good controllability, whereas RBMK-1000 has an instable behaviour and complex spatial dependencies in the core. Among other points, design deficiencies in the protection and emergency shutdown systems were responsible for the Chernobyl accident. The protection and scram systems of the SNR-300 are unquestionably superior to those of the RBMK-1000 with regard to redundancy, diversity, degree of automation, separation of operational and safety-relevant tasks, protection against inadmissible interventions, effectiveness and safety reserves. Therefore, excursion accidents can be classified as hypothetical for SNR-300. Due to elementary physical properties, possible energy releases during hypothetical excursions are substantially lower for SNR-300 and would be controlled by the design of the primary system and containment systems. No damage limiting measures are provided in the RBMK-100 for excursion accidents. Finally, exothermal processes augmented the consequences of the accident in the RBMK-1000 and the long-lasting graphite fire intensified the release of radioactivity. In the SNR-300, however, inertisation of the containment, the steel plate lining and the floor troughs ensure that activity enclosure inside the containment after leakage or hypothetical excursion accident is not endangered by exothermal reactions. Further safety aspects are presented in the report, which can be linked with the accident in Chernobyl. In summary, it is obvious that the disadvantageous physical and technical features of the RBMK-1000 do either not exist in the SNR-300 or are covered by the safety design

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)

Accident analysis of Fukushima Daiichi Nuclear Power Station unit 1

International Nuclear Information System (INIS)

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

2015-01-01

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

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)

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

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.

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.

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

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.

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

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

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.

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

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

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.

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

Condensate-polisher resin-leakage quantification and resin-transport studies

International Nuclear Information System (INIS)

Stauffer, C.C.; Doss, P.L.

1983-04-01

The objectives of this program were to: (1) determine the extent of resin leakage from current generation condensate polisher systems, both deep bed and powdered resin design, during cut-in, steady-state and flow transient operation, (2) analyze moisture separator drains and other secondary system samples for resin fragments and (3) document the level of organics in the secondary system. Resin leakage samples were obtained from nine-power stations that have either recirculating steam generators or once through steam generators. Secondary system samples were obtained from steam generator feedwater, recirculating steam generator blowdown and moisture separator drains. Analysis included ultraviolet light examination, SEM/EDX, resin quantification and infrared analysis. Data obtained from the various plants were compared and factors affecting resin leakage were summarized

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)

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...

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

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

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

Benchmark calculations for evaluation methods of gas volumetric leakage rate

International Nuclear Information System (INIS)

Asano, R.; Aritomi, M.; Matsuzaki, M.

1998-01-01

A containment function of radioactive materials transport casks is essential for safe transportation to prevent the radioactive materials from being released into environment. Regulations such as IAEA standard determined the limit of radioactivity to be released. Since is not practical for the leakage tests to measure directly the radioactivity release from a package, as gas volumetric leakages rates are proposed in ANSI N14.5 and ISO standards. In our previous works, gas volumetric leakage rates for several kinds of gas from various leaks were measured and two evaluation methods, 'a simple evaluation method' and 'a strict evaluation method', were proposed based on the results. The simple evaluation method considers the friction loss of laminar flow with expansion effect. The strict evaluating method considers an exit loss in addition to the friction loss. In this study, four worked examples were completed for on assumed large spent fuel transport cask (Type B Package) with wet or dry capacity and at three transport conditions; normal transport with intact fuels or failed fuels, and an accident in transport. The standard leakage rates and criteria for two kinds of leak test were calculated for each example by each evaluation method. The following observations are made based upon the calculations and evaluations: the choked flow model of ANSI method greatly overestimates the criteria for tests ; the laminar flow models of both ANSI and ISO methods slightly overestimate the criteria for tests; the above two results are within the design margin for ordinary transport condition and all methods are useful for the evaluation; for severe condition such as failed fuel transportation, it should pay attention to apply a choked flow model of ANSI method. (authors)

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

Improving the safety and reliability of Monju

International Nuclear Information System (INIS)

Itou, Kazumoto; Maeda, Hiroshi; Moriyama, Masatoshi

1998-01-01

Comprehensive safety review has been performed at Monju to determine why the Monju secondary sodium leakage accident occurred. We investigated how to improve the situation based on the results of the safety review. The safety review focused on five aspects of whether the facilities for dealing with the sodium leakage accident were adequate: the reliability of the detection method, the reliability of the method for preventing the spread of the sodium leakage accident, whether the documented operating procedures are adequate, whether the quality assurance system, program, and actions were properly performed and so on. As a result, we established for Monju a better method of dealing with sodium leakage accidents, rapid detection of sodium leakage, improvement of sodium drain facilities, and way to reduce damage to Monju systems after an accident. We also improve the operation procedures and quality assurance actions to increase the safety and reliability of Monju. (author)

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.

Control room unfiltered in-leakage limit analysis of design-basis LOCA for Lungmen ABWR plant

International Nuclear Information System (INIS)

Tsai Chihming; Chang Chinjang; Yuann Yngruey

2014-01-01

In USNRC's Generic Letter 2003-01, 'Control Room Habitability,' it requests utilities provide information to demonstrate that the control room at each of their respective facilities complies with the current licensing and design bases, and applicable regulatory requirements, and that suitable design, maintenance and testing control measures are in place for maintaining this compliance. In particular, each utility is required to perform the control room in-leakage test to demonstrate that the unfiltered in-leakage rate is within that assumed in the licensing analyses. It must be ensured that the control room envelope habitability, in terms of radiation dose, is maintained during normal operations as well as design basis accidents. In view of this, a dose analysis has been performed to establish the control room unfiltered in-leakage limit which can be used as an acceptance criterion for the in-leakage test. The analysis in this study is for Lungmen ABWR plant. The plant has twin units, with each unit having its own control room. The TID-4844 source terms and associated methodology are used. The USNRC RADTRAD v3.03 code is employed for the transport calculation of radioactive materials in different paths, including control room in-leakage path. The radiological criterion on protection of the operators specified in 10 CFR 50, Appendix A, General Design Criterion 19 is followed. It's demonstrated that the performance of Lungmen control room with 500 cfm unfiltered in-leakage air could meet the radiological habitability acceptance criteria in case of radiation hazards. (author)

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

Practical Leakage-Resilient Symmetric Cryptography

DEFF Research Database (Denmark)

Faust, Sebastian; Pietrzak, Krzysztof; Schipper, Joachim

2012-01-01

Leakage resilient cryptography attempts to incorporate side-channel leakage into the black-box security model and designs cryptographic schemes that are provably secure within it. Informally, a scheme is leakage-resilient if it remains secure even if an adversary learns a bounded amount of arbitr......Leakage resilient cryptography attempts to incorporate side-channel leakage into the black-box security model and designs cryptographic schemes that are provably secure within it. Informally, a scheme is leakage-resilient if it remains secure even if an adversary learns a bounded amount...

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

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...

Investigation of effect of stopping supply flow into the cell on the confinement of the radioactive materials under fire accident

International Nuclear Information System (INIS)

Abe, Hitoshi; Watanabe, Koji

1999-03-01

On November 20th 1997, a fire accident happened at Uranium Enrichment Research Laboratory, Tokai, Japan Atomic Energy Research Institute and ventilation filters in the laboratory clogged. When fire accident occurs in a controlled area, a large quantity of smoke generates in the area and dropping exhaust flow from the area by the clogging of ventilation filters and rising pressure in the area are caused. Moreover, leakage of smoke including radioactive materials from the area by the pressure rising is expected. To prevent the leakage, it is expected that stopping supply flow to the area during a fire accident is effective, however, quantitative evaluation about this effect has not been performed. By using CELVA-1D code, one-dimensional thermofluid analysis code, this effect is evaluated quantitatively by modeling the laboratory and estimating source terms released during the fire accident. As the results, it has been found that the efficiency of confinement of the radioactive materials into the area is preserved in the slightly long period of time in case of stopping supply flow to the area, however, this effect can be neglected in case that scale of fire accident is relatively large. (author)

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.

[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.

Dose assessment method for control room habitability in accident condition in nuclear power plants

International Nuclear Information System (INIS)

Yang Dong; Tang Shaohua; Wang Jianhua

2012-01-01

Based on the NRC. technical requirements on NPP control room habitability assessment, and considering the characteristics of the improved second generation NPPs in China, this paper developed a complete dose assessment model for control room habitability. Contrasting to the existing model in China, this model is applicable for DBA and sever accident, and the short term atmospheric diffusion factor can be calculated using the combined wake mode. By considering the zoning of habitable area and the design characteristics of the ventilation system, the effects of un-filtrated air leakage from the building and the ventilation system on the assessment calculation can be considered. (authors)

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.

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

[Bile leakage after liver resection: A retrospective cohort study].

Science.gov (United States)

Menclová, K; Bělina, F; Pudil, J; Langer, D; Ryska, M

2015-12-01

Many previous reports have focused on bile leakage after liver resection. Despite the improvements in surgical techniques and perioperative care the incidence of this complication rather keeps increasing. A number of predictive factors have been analyzed. There is still no consensus regarding their influence on the formation of bile leakage. The objective of our analysis was to evaluate the incidence of bile leakage, its impact on mortality and duration of hospitalization at our department. At the same time, we conducted an analysis of known predictive factors. The authors present a retrospective review of the set of 146 patients who underwent liver resection at the Department of Surgery of the 2nd Faculty of Medicine of the Charles University and Central Military Hospital Prague, performed between 20102013. We used the current ISGLS (International Study Group of Liver Surgery) classification to evaluate the bile leakage. The severity of this complication was determined according to the Clavien-Dindo classification system. Statistical significance of the predictive factors was determined using Fishers exact test and Students t-test. The incidence of bile leakage was 21%. According to ISGLS classification the A, B, and C rates were 6.5%, 61.2%, and 32.3%, respectively. The severity of bile leakage according to the Clavien-Dindo classification system - I-II, IIIa, IIIb, IV and V rates were 19.3%, 42%, 9.7%, 9.7%, and 19.3%, respectively. We determined the following predictive factors as statistically significant: surgery for malignancy (pBile leakage significantly prolonged hospitalization time (pbile leakage the perioperative mortality was 23 times higher (pBile leakage is one of the most serious complications of liver surgery. Most of the risk factors are not easily controllable and there is no clear consensus on their influence. Intraoperative leak tests could probably reduce the incidence of bile leakage. In the future, further studies will be required to improve

The COLIMA experiment on aerosol retention in containment leak paths under severe nuclear accidents

Energy Technology Data Exchange (ETDEWEB)

Parozzi, Flavio, E-mail: flavio.parozzi@rse-web.it [RSE, Power Generation Department, via Rubattino 54, I-20134 Milano (Italy); Caracciolo, Eduardo D.J., E-mail: eduardo.caracciolo@rse-web.it [RSE, Power Generation Department, via Rubattino 54, I-20134 Milano (Italy); Journeau, Christophe, E-mail: christophe.journeau@cea.fr [CEA Cadarache (France); Piluso, Pascal, E-mail: pascal.piluso@cea.fr [CEA Cadarache (France)

2013-08-15

Highlights: ► Experiment investigating aerosol retention within concrete containment cracks under nuclear severe accident conditions. ► Provided representative conditions of the aerosols suspended inside the containment of PWRs under a severe accident. ► Prototypical aerosol particles generated with a thermite reaction and transported through the crack sample reproducing surface characteristics, temperature, pressure drop and gas leakage. ► The results indicate the significant retention due to zig-zag path. -- Abstract: CEA and RSE managed an experimental research concerning the investigation of aerosol retention within concrete containment cracks under severe accident conditions. The main experiment was carried out in November 2008 with aerosol generated from the COLIMA facility and a sample of cracked concrete with defined geometric characteristics manufactured by RSE. The facility provided representative conditions of the aerosols suspended inside the containment of PWRs under a severe accident. Prototypical aerosol particles were generated with a thermite reaction and transported through the crack sample, where surface characteristics, temperature, pressure drop and gas leakage were properly reproduced. The paper describes the approach adopted for the preparation of the cracked concrete sample and the dimensioning of the experimental apparatus, the test procedure and the measured parameters. The preliminary results, obtained from this single test, are also discussed in the light of the present knowledge about aerosol phenomena and the theoretical analyses of particle behaviour with the crack path.

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

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

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.

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)

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.

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.

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

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 fluctuation of the calculated leak rate, it is sometimes hard to know the trend of the leakage as well as the instantaneous leak rate. Any fluctuation of operating conditions can results in unreliable leak rate. This study proposes a new way of determining the unidentified leak rate using a Kalman filter and smoother technique. The proposed algorithm enhances the accuracy of the leak rate calculation not only for the steady state operations but also for transients in a well timed manner

Highly Reliable Transformerless Photovoltaic Inverters With Leakage Current and Pulsating Power Elimination

DEFF Research Database (Denmark)

Tang, Yi; Yao, Wenli; Loh, Poh Chiang

2016-01-01

This paper presents a transformerless inverter topology, which is capable of simultaneously solving leakage current and pulsating power issues in grid-connected photovoltaic (PV) systems. Without adding any additional components to the system, the leakage current caused by the PV-to-ground parasi......This paper presents a transformerless inverter topology, which is capable of simultaneously solving leakage current and pulsating power issues in grid-connected photovoltaic (PV) systems. Without adding any additional components to the system, the leakage current caused by the PV......-to-ground parasitic capacitance can be bypassed by introducing a common-mode (CM) conducting path to the inverter. The resulting ground leakage current is therefore well controlled to be below the regulation limit. Furthermore, the proposed inverter can also eliminate the well-known double-line-frequency pulsating...... power that is inherent in single-phase PV systems. By properly injecting CM voltages to the output filter capacitors, the pulsating power can be decoupled from the dc-link. Therefore, it is possible to use long-lifetime film capacitors instead of electrolytic capacitors to improve the reliability...

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

[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...

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.

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

Long-term distribution of Cs-137 in freshwater ecosystem and the effect of bioturbation on leakage of Cs-137 from the sediments

International Nuclear Information System (INIS)

Aquilonius, K.

1995-06-01

The first aim of the project was to make a literature compilation regarding Cs-137 in freshwater systems. Secondly, in order to follow up the long-term behaviour of Cs-137 in Lake Hillesjoen, sediment and water was sampled with respect to Cs-137. Finally, an experiment was done with chironomids, with the aim to assess the impact of bioturbation on the leakage of Cs-137 from the sediment, since this is one of the factors contributing to the uncertainties in radioecological models. Some aspects of long-term distribution of Cs-137 in freshwater lakes is discussed. In lake Hillesjoen the pool of Cs-137 in sediments (225 kBq/m 2 , 1993) has gradually increased since the Chernobyl accident, and now exceeds the initial deposition (100 kBq/m 2 ). The effect of bioturbation on the leakage from the sediment to the water could not be clarified in this experiment, but is discussed. 40 refs

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.

Research on leakage detection of nuclear power station primary loop

International Nuclear Information System (INIS)

Cao Shengquan; Yang Jicai; Li Jianjun; Dong Zhijuan; Ren Juyan; Zhang Pinyuan

1993-01-01

The sensor of leak-detection system is a special humidity-sensitive element. In this system, the collection and treatment of detected data, and the leakage alarm are controlled by a simple board computer. The three parameters A, B and C are set up in this system, and the alarm will give out when signal surpasses values of B and C at the same time. The effects of environment humidity and temperature on collection of detected data are considered, and the leakage alarm is reliable. When the leakage rate of water at 320 degree C and 12.2 MPa is 0.3 g/min, the alarm signal will sent out within 5 s

Design features of ACR in severe accident mitigation

International Nuclear Information System (INIS)

Shapiro, H.; Krishnan, V.S.; Santamaura, P.; Lekakh, B.; Blahnik, C.

2007-01-01

New reactor designs require the evaluation of design alternatives to reduce the radiological risk by preventing severe accidents or by limiting releases from the plant in the event of such accidents. The Advanced CANDU Reactor TM (ACR TM ) design has provisions to prevent and mitigate severe accidents. This paper describes key ACR design features for severe accident mitigation. It provides a high-level overview of the findings to date. Several design provisions have not yet been finalized or decided, but the designers are keenly aware of the SAM concepts and their requirements. The active heat sinks for 'vessels' (i.e., the fuel channels, the calandria vessel, the calandria end-shields and the calandria vault) are all amply capable of dissipating the severe accident heat loads. These heat sinks are designed to be operable under severe accident environmental conditions; however, their operability is yet to be confirmed by assessments. The active heat sinks for the various process vessels are 'backed up' by passive heat sinks (i.e., steaming plus water make-up from the RWS). The supply side of passive heat sinks is simple, rugged, and not vulnerable to failures of plant systems. The importance of the steam relief side is recognized, and the adequate relief capacity will be provided. The passive heat sinks will give the SAM more than 1 day (likely several days) to diagnose the accident and to establish the ultimate heat sinks. The spray system for containment pressure suppression is designed for high reliability and has ample capacity to ensure low containment leakage without external intervention, after which time alternative supply to the sprays can be brought on line manually. The sprays are backed up by the LACs which are assessed for operability following a severe accident. The strong ACR containment will provide a long time of completely passive protection for any severe accident at decay power. Its characteristics are not prone to catastrophic failures. The

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

LMFBR technology. FFTF cover-gas leakage calculation

International Nuclear Information System (INIS)

Deboi, H.

1974-01-01

The FFTF LMFBR is intended to have a near zero release of radioactive gases during normal reactor operation with 1% failed fuel. This report presents calculations which provide an approximation of these cover gas leakages. Data from ongoing static and dynamic seal leak tests at AI are utilized. Leakage through both elastomeric and metallic seals in all sub-assemblies and penetrations comprising the reactor cover gas containment during reactor operation system are included

Feasibility study of tank leakage mitigation using subsurface barriers

International Nuclear Information System (INIS)

Treat, R.L.; Peters, B.B.; Cameron, R.J.; McCormak, W.D.; Trenkler, T.; Walters, M.F.; Rouse, J.K.; McLaughlin, T.J.; Cruse, J.M.

1994-01-01

The US Department of Energy (DOE) has established the Tank Waste Remediation System (TWRS) to satisfy manage and dispose of the waste currently stored in the underground storage tanks. The retrieval element of TWRS includes a work scope to develop subsurface impermeable barriers beneath SSTs. The barriers could serve as a means to contain leakage that may result from waste retrieval operations and could also support site closure activities by facilitating cleanup. Three types of subsurface barrier systems have emerged for further consideration: (1) chemical grout, (2) freeze walls, and (3) desiccant, represented in this feasibility study as a circulating air barrier. This report contains analyses of the costs and relative risks associated with combinations retrieval technologies and barrier technologies that from 14 alternatives. Eight of the alternatives include the use of subsurface barriers; the remaining six nonbarrier alternative are included in order to compare the costs, relative risks and other values of retrieval with subsurface barriers. Each alternative includes various combinations of technologies that can impact the risks associated with future contamination of the groundwater beneath the Hanford Site to varying degrees. Other potential risks associated with these alternatives, such as those related to accidents and airborne contamination resulting from retrieval and barrier emplacement operations, are not quantitatively evaluated in this report

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.

[Preparation of the database and the homepage on chemical accidents relating to health hazard].

Science.gov (United States)

Yamamoto, M; Morita, M; Kaminuma, T

1998-01-01

We collected the data on accidents due to chemicals occurred in Japan, and prepared the database. We also set up the World Wide Web homepage containing the explanation on accidents due to chemicals and the retrieval page for the database. We designed the retrieval page so that users can search the data from keywords such as chemicals (e.g. chlorine gas, hydrogen sulfide, pesticides), places (e.g. home, factory, vehicles, tank), causes (e.g. reaction, leakage, exhaust gas) and others (e.g. cleaning, painting, transportation).

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.)

Transformerless photovoltaic inverters with leakage current and pulsating power elimination

DEFF Research Database (Denmark)

Tang, Yi; Yao, Wenli; Wang, H.

2015-01-01

This paper presents a transformerless inverter topology, which is capable of simultaneously solving leakage current and pulsating power issues in grid-connected photovoltaic (PV) systems. Without adding any additional components to the system, the leakage current caused by the PV......-to-ground parasitic capacitance can be bypassed by introducing a common mode (CM) conducting path to the inverter. The resulting ground leakage current is therefore well controlled to be below the regulation limit. Moreover, the proposed inverter can also eliminate the well-known double line frequency pulsating power....... The mechanism of leakage current suppression and the closed-loop control of pulsating power decoupling are discussed in the paper in details. A 500 W prototype was also built and tested in the laboratory, and both simulation and experimental results are finally presented to show the excellent performance...

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

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 ...

Options for management of containment integrity during severe accident in Indian PHWR

International Nuclear Information System (INIS)

Sharma, Sanjeev Kr.; Bhartia, D.K.; Mohan, N.; Nair, Suma R.

2015-01-01

Severe accident progressions have the potential to raise the containment pressure beyond the design pressure of the structure. Although the load withstanding capability of the containment structure has been assessed to be substantially higher than the design pressure of the structure (typically 2 times of design pressure), it is possible that a few components of Containment System may degrade leading to excessive release of radioactive fission gases at ground level. Additionally, possible cracks in the concrete of the containment at high pressure may aggravate the release at ground level. Over and above, maintaining high containment pressure high for a longer period increases the ground level release due to leakage from the containment, which effect on dose might be high. For maintaining the Integrity of the Containment, containment pressure can be reduced by either energy management system such as removing the heat from the calandria vault (CVWC) water by using CV water heat exchanger intermittently or reliving the containment atmosphere either through Primary Containment Controlled Discharge (PCCD) or Containment Filtered Venting System (CFVS). Further, it is necessary that these provisions must be initiated below design pressure. This paper presents the analysis for the containment depressurization by using CVWC system restored, manual opening of (PCCD) line and operation of CFVS during the progressions of the accident

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

Feasibility of Locating Leakages in Sewage Pressure Pipes Using the Distributed Temperature Sensing Technology.

Science.gov (United States)

Apperl, Benjamin; Pressl, Alexander; Schulz, Karsten

2017-01-01

The cost effective maintenance of underwater pressure pipes for sewage disposal in Austria requires the detection and localization of leakages. Extrusion of wastewater in lakes can heavily influence the water and bathing quality of surrounding waters. The Distributed Temperature Sensing (DTS) technology is a widely used technique for oil and gas pipeline leakage detection. While in pipeline leakage detection, fiber optic cables are installed permanently at the outside or within the protective sheathing of the pipe; this paper aims at testing the feasibility of detecting leakages with temporary introduced fiber optic cable inside the pipe. The detection and localization were tested in a laboratory experiment. The intrusion of water from leakages into the pipe, producing a local temperature drop, served as indicator for leakages. Measurements were taken under varying measurement conditions, including the number of leakages as well as the positioning of the fiber optic cable. Experiments showed that leakages could be detected accurately with the proposed methodology, when measuring resolution, temperature gradient and measurement time were properly selected. Despite the successful application of DTS for leakage detection in this lab environment, challenges in real system applications may arise from temperature gradients within the pipe system over longer distances and the placement of the cable into the real pipe system.

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.

Reanastomosis with Stapler in Duodenojejunal Junction Anastomosis Leakage: A Case Report

Directory of Open Access Journals (Sweden)

Ahmet Seker

2013-04-01

Full Text Available After anastomosis leakages, treatment of patient gets more difficult and mortality rates increase. At lower level gastrointestinal anastomosis leakages, because of always there is an ostomy alternative, digestion problems are seen lesser. But at upper level gastrointestinal system anastomosis leakages, when it is taken account of nutrient condition of patient, requirement of making anastomos increases. So moratlity rates increase. At this article we aimed to present a different technique that we administered on management of a patient who had duodenojejunal junction anastomosis leakage.

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.

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)

Severe accident testing of electrical penetration assemblies

International Nuclear Information System (INIS)

Clauss, D.B.

1989-11-01

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

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.

Functions and requirements for single-shell tank leakage mitigation

International Nuclear Information System (INIS)

Cruse, J.M.

1994-01-01

This document provides the initial functions and requirements for the leakage mitigation mission applicable to past and potential future leakage from the Hanford Site's 149 single-shell high-level waste tanks. This mission is a part of the overall mission of the Westinghouse Hanford Company Tank Waste Remediation System division to remediate the tank waste in a safe and acceptable manner. Systems engineering principles are being applied to this effort. A Mission Analysis has been completed, this document reflects the next step in the systems engineering approach to decompose the mission into primary functions and requirements. The functions and requirements in this document apply to mitigative actions to be taken regarding below ground leaks from SST containment boundaries and the resulting soil contamination. Leakage mitigation is invoked in the TWRS Program in three fourth level functions: (1) Store Waste, (2) Retrieve Waste, and (3) Disposition Excess Facilities

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.

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

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

International Nuclear Information System (INIS)

Hsii, Y.H.

1975-01-01

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

Emergency Evacuation of Hazardous Chemical Accidents Based on Diffusion Simulation

Directory of Open Access Journals (Sweden)

Jiang-Hua Zhang

2017-01-01

Full Text Available The recent rapid development of information technology, such as sensing technology, communications technology, and database, allows us to use simulation experiments for analyzing serious accidents caused by hazardous chemicals. Due to the toxicity and diffusion of hazardous chemicals, these accidents often lead to not only severe consequences and economic losses, but also traffic jams at the same time. Emergency evacuation after hazardous chemical accidents is an effective means to reduce the loss of life and property and to smoothly resume the transport network as soon as possible. This paper considers the dynamic changes of the hazardous chemicals’ concentration after their leakage and simulates the diffusion process. Based on the characteristics of emergency evacuation of hazardous chemical accidents, we build a mixed-integer programming model and design a heuristic algorithm using network optimization and diffusion simulation (hereafter NODS. We then verify the validity and feasibility of the algorithm using Jinan, China, as a computational example. In the end, we compare the results from different scenarios to explore the key factors affecting the effectiveness of the evacuation process.

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.

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.

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.

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

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.)

Protection of ground water at shallow geothermal power plants by means of an automatic leakage detection and liquid backwashing; Grundwasserschutz bei flachen Geothermieanlagen durch automatische Leckagenerkennung und Fluessigkeitsrueckspuelung

Energy Technology Data Exchange (ETDEWEB)

Wohnlich, Stefan; Scheliga, Roman [Bochum Univ. (Germany). Lehrstuhl fuer Angewandte Geologie; Bonin, Juergen [Umwelt und Technik, Xanten (Germany)

2011-10-24

A protective device is examined with which the contamination of groundwater in an accident can be reduced to a minimum. The proposed device (geo-protector) registered a leak in two stages of pressure sensors. If the value falls below the lower minimum pressure, the entire system is flushed with water. The brine can be collected in a separate vessel. Thus, only drinking water escapes from the leak. A further contamination of ground water with brine is avoided. In order to investigate the functionality and reliability of the geo-protectors, a model of a geothermal power plant was created. With this, leakages of varying sizes were simulated at different places of the geothermal probes. All measured leakage scenarios could be registered. The output amount of the brine during the flushing process was at most 5% of the total volume of the system. The output amount could be further minimized by means of a pressure reducer. The outflow quantity of the brine is reduced to a minimum by means of a shallow geothermal system. This significantly may contribute to the protection of groundwater.

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.

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)

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

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

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)

Experimental Investigation of Operation of VVER Steam Generator in Condensation Mode in the Event of the Severe Accident

Energy Technology Data Exchange (ETDEWEB)

Morozov, Andrey [Institute for Physics and Power Engineering by A.I. Leypunsky, 1 Bondarenko sq. Obninsk, 249033 (Russian Federation)

2008-07-01

For new Russian nuclear power plants with VVER-1200 reactor in the event of a beyond design basis accident, provision is made for the use of passive safety systems for necessary core cooling. These safety systems include the passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam. As a result, the condensate from SG arrives at the core providing its additional cooling. To investigate the condensation mode of VVER SG operation, a large scale HA2M-SG test facility was constructed. The rig incorporates: buffer tank, SG model with scale is 1:46, PHRS heat exchanger. Experiments at the test facility have been performed to investigate condensation mode of operation of SG model at the pressure 0.4 MPa, correspond to VVER reactor pressure at the last stage of the beyond design basis accident. The report presents the test procedure and the basic obtained test results. (authors)

Experimental investigations relevant for hydrogen and fission product issues raised by the Fukushima accident

Directory of Open Access Journals (Sweden)

Sanjeev Gupta

2015-02-01

Full Text Available The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS, unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled

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.

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.

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

Detection of gaseous heavy water leakage points in CANDU 6 pressurized heavy water reactors

International Nuclear Information System (INIS)

Park, T-K.; Jung, S-H.

1996-01-01

During reactor operation, the heavy water filled primary coolant system in a CANDU 6 Pressurized Heavy Water (PHWR) may leak through routine operations of the plant via components, mechanical joints, and during inadvertent operations etc. Early detection of leak points is therefore important to maintain plant safety and economy. There are many independent systems to monitor and recover heavy water leakage in a CANDU 6 PHWR. Methodology for early detection based on operating experience from these systems, is investigated in this paper. In addition, the four symptoms of D 2 O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

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)

[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.

Evaluation of isolation valve leakage in alternate charging piping

International Nuclear Information System (INIS)

Strauch, P.L.; Roarty, D.H.; Brice-Nash, R.L.

1995-01-01

The chemical and volume control system (CVCS) alternate charging flow path at an operating pressurized water reactor (PWR) plant was determined to be susceptible to adverse stresses from isolation valve leakage. Isolation valve leakage had resulted in pipe cracks at several nuclear units worldwide, as described in United States Nuclear Regulatory Commission Bulletin 88-08 and its supplements. To provide for continuing assurance that cracks would not initiate over the plant life, the operators considered performing fatigue evaluation to demonstrate structural integrity of the system. This evaluation included heat transfer, stress and fatigue analysis, using methods described in Electric Power Research Institute Report ''Thermal Stratification, Cycling, and Striping (TASCS),'' March 1994. The evaluation concluded that the fatigue usage would be less than 1.0 under worst case isolation valve leakage conditions, and therefore a significant investment in permanent temperature monitoring was avoided

Nuclear Reactor RA Safety Report, Vol. 12, Accidents during reactor operation

International Nuclear Information System (INIS)

1986-11-01

This volume includes description and analysis of typical accidents occurred during operation of RA reactor in chronological order, as follows: contamination of primary coolant circuit; leakage of heavy water from the primary coolant loop; contamination of vertical experimental channel; air contamination in the reactor building and loss of circulation of the primary coolant; failures of the vacuum pump and spent fuel packaging device; rupture of the spent fuel element cladding; dethronement's of capsule for irradiation of fuel element; rupture of the vertical experimental channel and contamination of the surroundings; swelling of a fuel element; appearance of deposits on the surface of the fuel elements cladding. The last chapter describes similar accidents occurred on nuclear reactors in the world [sr

Accident sequences and causes analysis in a hydrogen production process

Energy Technology Data Exchange (ETDEWEB)

Jae, Moo Sung; Hwang, Seok Won; Kang, Kyong Min; Ryu, Jung Hyun; Kim, Min Soo; Cho, Nam Chul; Jeon, Ho Jun; Jung, Gun Hyo; Han, Kyu Min; Lee, Seng Woo [Hanyang Univ., Seoul (Korea, Republic of)

2006-03-15

Since hydrogen production facility using IS process requires high temperature of nuclear power plant, safety assessment should be performed to guarantee the safety of facility. First of all, accident cases of hydrogen production and utilization has been surveyed. Based on the results, risk factors which can be derived from hydrogen production facility were identified. Besides the correlation between risk factors are schematized using influence diagram. Also initiating events of hydrogen production facility were identified and accident scenario development and quantification were performed. PSA methodology was used for identification of initiating event and master logic diagram was used for selection method of initiating event. Event tree analysis was used for quantification of accident scenario. The sum of all the leakage frequencies is 1.22x10{sup -4} which is similar value (1.0x10{sup -4}) for core damage frequency that International Nuclear Safety Advisory Group of IAEA suggested as a criteria.

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

Impact of leakage delay on bifurcation in high-order fractional BAM neural networks.

Science.gov (United States)

Huang, Chengdai; Cao, Jinde

2018-02-01

The effects of leakage delay on the dynamics of neural networks with integer-order have lately been received considerable attention. It has been confirmed that fractional neural networks more appropriately uncover the dynamical properties of neural networks, but the results of fractional neural networks with leakage delay are relatively few. This paper primarily concentrates on the issue of bifurcation for high-order fractional bidirectional associative memory(BAM) neural networks involving leakage delay. The first attempt is made to tackle the stability and bifurcation of high-order fractional BAM neural networks with time delay in leakage terms in this paper. The conditions for the appearance of bifurcation for the proposed systems with leakage delay are firstly established by adopting time delay as a bifurcation parameter. Then, the bifurcation criteria of such system without leakage delay are successfully acquired. Comparative analysis wondrously detects that the stability performance of the proposed high-order fractional neural networks is critically weakened by leakage delay, they cannot be overlooked. Numerical examples are ultimately exhibited to attest the efficiency of the theoretical results. Copyright © 2017 Elsevier Ltd. All rights reserved.

Study on the Break Accidents of the HTR-PM Primary Loop

International Nuclear Information System (INIS)

Lang Minggang; Sun Ximing; Zheng Yanhua

2014-01-01

In thermal hydraulics design and safety analysis of the HTR-PM, the THERMIX code was used to study the behavior of the helium in the primary system. Once the helium leaks from the primary loop through a break or a relief valve, it is hard to simulate the states of the leakage room with THERMIX. In this paper, the latest version of RELAP5/MOD4, was used to simulate the behavior of the helium released to the containment rooms. A RELAP5/MOD4 model of the HTR-PM, including the core, the primary system, the secondary loop and the containment, were developed and evaluated in this paper. Based on the model, this paper studied the accidents consequences of a large break in the pressure relief room and a small break in the instrument room of the HTR-PM reactor building. The simulating results illustrate that the temperature in the pressure relief room was no more than 200℃ after a un-isolating large break, and the temperature in the instrument room is less than 130 ℃ after a small un-isolating break. The analysis shows that the scram function and the ability to monitor the reactor temperature and pressure after accidents would not be affected by the break. (author)

Leakage pattern of linear accelerator treatment heads from multiple vendors

International Nuclear Information System (INIS)

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

2011-01-01

Full text: Patient life expectancy post-radiotherapy is becoming longer. Therefore, secondary cancers caused by radiotherapy treatment have more time to develop. Increasing attention is being given to out-of-field dose resulting from scatter and accelerator head leakage. Dose leakage from equivalent positions on Varian600C, Varian21-X, Siemens Primus and Elekta Synergy-II linacs were measured with TLD 1 00 H dosimeter chips and compared. Treatment parameters such as field size and beam energy were altered. Leakage doses are presented as a percentage of the dose to isocentre (5 Gy). Results illustrate significant variations in leakage dose between linac models where no model emits consistently lower amounts of radiation leakage for all treatment parameters. Results are shown below. Leakage through the collimator assembly in different units is varying as a function of location and unit design by more than a factor of 10. Differences are more pronounced in comparing Varian or Elekta models, which are fitted with an additional collimator separate from the MLC leaves, to the Siemens model, which uses MLC leaves as its only secondary collimator. Further measurements are currently being taken at the patient plane with a directional detector system to determine the spatial distribution of high leakage sources.

Theory and Application of Magnetic Flux Leakage Pipeline Detection.

Science.gov (United States)

Shi, Yan; Zhang, Chao; Li, Rui; Cai, Maolin; Jia, Guanwei

2015-12-10

Magnetic flux leakage (MFL) detection is one of the most popular methods of pipeline inspection. It is a nondestructive testing technique which uses magnetic sensitive sensors to detect the magnetic leakage field of defects on both the internal and external surfaces of pipelines. This paper introduces the main principles, measurement and processing of MFL data. As the key point of a quantitative analysis of MFL detection, the identification of the leakage magnetic signal is also discussed. In addition, the advantages and disadvantages of different identification methods are analyzed. Then the paper briefly introduces the expert systems used. At the end of this paper, future developments in pipeline MFL detection are predicted.

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.

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

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.

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.

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

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)

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.

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

International Nuclear Information System (INIS)

Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu

2015-01-01

Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance

Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

Energy Technology Data Exchange (ETDEWEB)

Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu [KAERI, Daejeon (Korea, Republic of)

2015-05-15

Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance.

Predicting Envelope Leakage in Attached Dwellings

Energy Technology Data Exchange (ETDEWEB)

Faakye, O. [Consortium for Advanced Residential Buildings (CARB), Norwalk, CT (United States); Arena, L. [Consortium for Advanced Residential Buildings (CARB), Norwalk, CT (United States); Griffiths, D. [Consortium for Advanced Residential Buildings (CARB), Norwalk, CT (United States)

2013-07-01

The most common method for measuring air leakage is to use a single blower door to pressurize and/or depressurize the test unit. In detached housing, the test unit is the entire home and the single blower door measures air leakage to the outside. In attached housing, this 'single unit', 'total', or 'solo' test method measures both the air leakage between adjacent units through common surfaces as well air leakage to the outside. Measuring and minimizing this total leakage is recommended to avoid indoor air quality issues between units, reduce energy losses to the outside, reduce pressure differentials between units, and control stack effect. However, two significant limitations of the total leakage measurement in attached housing are: for retrofit work, if total leakage is assumed to be all to the outside, the energy benefits of air sealing can be significantly over predicted; for new construction, the total leakage values may result in failing to meet an energy-based house tightness program criterion. The scope of this research is to investigate an approach for developing a viable simplified algorithm that can be used by contractors to assess energy efficiency program qualification and/or compliance based upon solo test results.

Subsea Hydraulic Leakage Detection and Diagnosis

OpenAIRE

Stavenes, Thomas

2010-01-01

The motivation for this thesis is reduction of hydraulic emissions, minimizing of process emergency shutdowns, exploitation of intervention capacity, and reduction of costs. Today, monitoring of hydraulic leakages is scarce and the main way to detect leakage is the constant need for filling of hydraulic fluid to the Hydraulic Power Unit (HPU). Leakage detection and diagnosis has potential, which would be adressed in this thesis. A strategy towards leakage detection and diagnosis is given....

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)

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

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.

A method to investigate inter-aquifer leakage using hydraulics and multiple environmental tracers

Science.gov (United States)

Priestley, Stacey; Love, Andrew; Wohling, Daniel; Post, Vincent; Shand, Paul; Kipfer, Rolf; Tyroller, Lina

2016-04-01

Informed aquifer management decisions regarding sustainable yields or potential exploitation require an understanding of the groundwater system (Alley et al. 2002, Cherry and Parker 2004). Recently, the increase in coal seam gas (CSG) or shale gas production has highlighted the need for a better understanding of inter-aquifer leakage and contaminant migration. In most groundwater systems, the quantity or location of inter-aquifer leakage is unknown. Not taking into account leakage rates in the analysis of large scale flow systems can also lead to significant errors in the estimates of groundwater flow rates in aquifers (Love et al. 1993, Toth 2009). There is an urgent need for robust methods to investigate inter-aquifer leakage at a regional scale. This study builds on previous groundwater flow and inter-aquifer leakage studies to provide a methodology to investigate inter-aquifer leakage in a regional sedimentary basin using hydraulics and a multi-tracer approach. The methodology incorporates geological, hydrogeological and hydrochemical information in the basin to determine the likelihood and location of inter-aquifer leakage. Of particular benefit is the analysis of hydraulic heads and environmental tracers at nested piezometers, or where these are unavailable bore couplets comprising bores above and below the aquitard of interest within a localised geographical area. The proposed methodology has been successful in investigating inter-aquifer leakage in the Arckaringa Basin, South Australia. The suite of environmental tracers and isotopes used to analyse inter-aquifer leakage included the stable isotopes of water, radiocarbon, chloride-36, 87Sr/86Sr and helium isotopes. There is evidence for inter-aquifer leakage in the centre of the basin ~40 km along the regional flow path. This inter-aquifer leakage has been identified by a slight draw-down in the upper aquifer during pumping in the lower aquifer, overlap in Sr isotopes, δ2H, δ18O and chloride

Application of the accident management information needs methodology to a severe accident sequence

International Nuclear Information System (INIS)

Ward, L.W.; Hanson, D.J.; Nelson, W.R.; Solberg, D.E.

1989-01-01

The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel

Application of the accident management information needs methodology to a severe accident sequence

Energy Technology Data Exchange (ETDEWEB)

Ward, L.W.; Hanson, D.J.; Nelson, W.R. (Idaho National Engineering Laboratory, Idaho Falls (USA)); Solberg, D.E. (Nuclear Regulatory Commission, Washington, DC (USA))

1989-11-01

The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel.

Analysis of flammability in the attached buildings to containment under severe accident conditions

Energy Technology Data Exchange (ETDEWEB)

Rosa, J.C. de la, E-mail: juan-carlos.de-la-rosa-blul@ec.europa.eu [European Commission Joint Research Centre (Netherlands); Fornós, Joan, E-mail: jfornosh@anacnv.com [Asociación Nuclear Ascó-Vandellós (Spain)

2016-11-15

Highlights: • Analysis of flammability conditions in buildings outside containment. • Stepwise approach easily applicable for any kind of containment and attached buildings layout. • Detailed application for real plant conditions has been included. - Abstract: Right after the events unfolded in Fukushima Daiichi, the European Union countries agreed in subjecting Nuclear Power Plants to Stress Tests as developed by WENRA and ENSREG organizations. One of the results as implemented in many European countries derived from such tests consisted of mandatory technical instructions issued by nuclear regulatory bodies on the analysis of potential risk of flammable gases in attached buildings to containment. The current study addresses the key aspects of the analysis of flammable gases leaking to auxiliary buildings attached to Westinghouse large-dry PWR containment for the specific situation where mitigating systems to prevent flammable gases to grow up inside containment are available, and containment integrity is preserved – hence avoiding isolation system failure. It also provides a full practical exercise where lessons learned derived from the current study – hence limited to the imposed boundary conditions – are applied. The leakage of gas from the containment to the support buildings is based on separate calculations using the EPRI-owned Modular Accident Analysis Program, MAAP4.07. The FATE™ code (facility Flow, Aerosol, Thermal, and Explosion) was used to model the transport and distribution of leaked flammable gas (H{sub 2} and CO) in the penetration buildings. FATE models the significant mixing (dilution) which occurs as the released buoyant gas rises and entrains air. Also, FATE accounts for the condensation of steam on room surfaces, an effect which acts to concentrate flammable gas. The results of the analysis show that during a severe accident, flammable conditions are unlikely to occur in compartmentalized buildings such as the one used in the

Analysis of flammability in the attached buildings to containment under severe accident conditions

International Nuclear Information System (INIS)

Rosa, J.C. de la; Fornós, Joan

2016-01-01

Highlights: • Analysis of flammability conditions in buildings outside containment. • Stepwise approach easily applicable for any kind of containment and attached buildings layout. • Detailed application for real plant conditions has been included. - Abstract: Right after the events unfolded in Fukushima Daiichi, the European Union countries agreed in subjecting Nuclear Power Plants to Stress Tests as developed by WENRA and ENSREG organizations. One of the results as implemented in many European countries derived from such tests consisted of mandatory technical instructions issued by nuclear regulatory bodies on the analysis of potential risk of flammable gases in attached buildings to containment. The current study addresses the key aspects of the analysis of flammable gases leaking to auxiliary buildings attached to Westinghouse large-dry PWR containment for the specific situation where mitigating systems to prevent flammable gases to grow up inside containment are available, and containment integrity is preserved – hence avoiding isolation system failure. It also provides a full practical exercise where lessons learned derived from the current study – hence limited to the imposed boundary conditions – are applied. The leakage of gas from the containment to the support buildings is based on separate calculations using the EPRI-owned Modular Accident Analysis Program, MAAP4.07. The FATE™ code (facility Flow, Aerosol, Thermal, and Explosion) was used to model the transport and distribution of leaked flammable gas (H_2 and CO) in the penetration buildings. FATE models the significant mixing (dilution) which occurs as the released buoyant gas rises and entrains air. Also, FATE accounts for the condensation of steam on room surfaces, an effect which acts to concentrate flammable gas. The results of the analysis show that during a severe accident, flammable conditions are unlikely to occur in compartmentalized buildings such as the one used in the

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.

International Clean Energy System Using Hydrogen Conversion (WE-NET). subtask 3. Conceptual design of the total system (safety measures and evaluation technology); Suiso riyo kokusai clean energy system gijutsu (WE-NET). subtask 3. Zentai system gainen sekkei (anzen taisaku hyoka gijutsu)

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-03-01

Safety measures and assessment technology were studied for the WE-NET project. As the study result in fiscal 1996, the information on safety design, anomaly and accident was collected and arranged. The information on safety measures, ideology and criterion was also collected by visiting some domestic and overseas organizations experienced about handling of liquid hydrogen (LH). The initial survey was made for the safety design ideology, analytical technique and disaster preventive measures of LNG systems as the similar cold liquid system. Accidents and explosion accident of a hydrogen production plant (water electrolysis) in Germany were analyzed. Events on storage tanks and leakage around the tanks were studied as typical risk of LH considering temporary and LNG system design information. The model based on the LH spillage test result and 3-D dispersion of vapor cloud were prepared by modifying a simulation code. The model allowed evaluation of the effect of explosion and fire accidents of compressed hydrogen gas and flying fragments on structures and people, and visual display of distances from a tank and damage conditions. 19 refs., 29 figs., 18 tabs.

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.

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.

Reactor coolant pump seal leakage monitoring

International Nuclear Information System (INIS)

Stevens, D.M.; Spencer, J.W.; Morris, D.J.; James, W.; Shugars, H.G.

1986-01-01

Problems with reactor coolant pump seals have historically accounted for a large percentage of unscheduled outages. Studies performed for the Electric Power Research Institute (EPRI) have shown that the replacement of coolant pump seals has been one of the leading causes of nuclear plant unavailability over the last ten years. Failures of coolant pump seals can lead to primary coolant leakage rates of 200-500 gallons per minute into the reactor building. Airborne activity and high surface contamination levels following these failures require a major cleanup effort and increases the time and personnel exposure required to refurbish the pump seals. One of the problems in assessing seal integrity is the inability to accurately measure seal leakage. Because seal leakage flow is normally very small, it cannot be sensed directly with normal flow instrumentation, but must be inferred from several other temperature and flow measurements. In operating plants the leakage rate has been quantified with a tipping-bucket gauge, a device which indicates when one quart of water has been accumulated. The tipping-bucket gauge has been used for most rainfall-intensity monitoring. The need for a more accurate and less expensive gauge has been addressed. They have developed a drop-counter precipitation sensor has been developed and optimized. The applicability of the drop-counter device to the problem of measuring seal leakage is being investigated. If a review of system specification and known drop-counter performance indicates that this method is feasible for measuring seal leak rates, a drop-counter gauge will be fabricated and tested in the laboratory. If laboratory tests are successful the gauge will be demonstrated in a pump test loop at Ontario Hydro and evaluated under simulated plant conditions. 3 references, 2 figures

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)

Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

International Nuclear Information System (INIS)

Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won

2015-01-01

In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario

Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

Energy Technology Data Exchange (ETDEWEB)

Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario.

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.

Insights into the behavior of LWR steel containment buildings during severe accidents

International Nuclear Information System (INIS)

Clauss, D.B.; Horschel, D.S.; Blejwas, T.E.

1987-01-01

Investigations into the performance of steel containment subject to pressure and temperature greater than their design basis loads are discussed. The timing, mechanism, and location of a containment failure, i.e., release of radioactive materials, have an important impact on the consequences of a severe accident. We review the results of experiments on steel containment models pressurized to failure, on aged and unaged seals subjected to elevated temperature and pressure, and on electrical penetration assemblies tested for leakage. Based on the results, the important features and details of analytical methods that can be used to predict containment performance are identified. Finally, we speculate on the performance of steel containments in severe accident conditions. (orig.)

Biomembrane Permeabilization: Statistics of Individual Leakage Events Harmonize the Interpretation of Vesicle Leakage.

Science.gov (United States)

Braun, Stefan; Pokorná, Šárka; Šachl, Radek; Hof, Martin; Heerklotz, Heiko; Hoernke, Maria

2018-01-23

The mode of action of membrane-active molecules, such as antimicrobial, anticancer, cell penetrating, and fusion peptides and their synthetic mimics, transfection agents, drug permeation enhancers, and biological signaling molecules (e.g., quorum sensing), involves either the general or local destabilization of the target membrane or the formation of defined, rather stable pores. Some effects aim at killing the cell, while others need to be limited in space and time to avoid serious damage. Biological tests reveal translocation of compounds and cell death but do not provide a detailed, mechanistic, and quantitative understanding of the modes of action and their molecular basis. Model membrane studies of membrane leakage have been used for decades to tackle this issue, but their interpretation in terms of biology has remained challenging and often quite limited. Here we compare two recent, powerful protocols to study model membrane leakage: the microscopic detection of dye influx into giant liposomes and time-correlated single photon counting experiments to characterize dye efflux from large unilamellar vesicles. A statistical treatment of both data sets does not only harmonize apparent discrepancies but also makes us aware of principal issues that have been confusing the interpretation of model membrane leakage data so far. Moreover, our study reveals a fundamental difference between nano- and microscale systems that needs to be taken into account when conclusions about microscale objects, such as cells, are drawn from nanoscale models.

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.

A Minimum Leakage Quasi-Static RAM Bitcell

Directory of Open Access Journals (Sweden)

Adam Teman

2011-05-01

Full Text Available As SRAMs continue to grow and comprise larger percentages of the area and power consumption in advanced systems, the need to minimize static currents becomes essential. This brief presents a novel 9T Quasi-Static RAM Bitcell that provides aggressive leakage reduction and high write margins. The quasi-static operation method of this cell, based on internal feedback and leakage ratios, minimizes static power while maintaining sufficient, albeit depleted, noise margins. This paper presents the concept of the novel cell, and discusses the stability of the cell under hold, read and write operations. The cell was implemented in a low-power 40 nm TSMC process, showing as much as a 12× reduction in leakage current at typical conditions, as compared to a standard 6T or 8T bitcell at the same supply voltage. The implemented cell showed full functionality under global and local process variations at nominal and low voltages, as low as 300 mV.

Report on the water leakage from instrumentation pipe in JMTR

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-03-01

On December 10, 2002, the leakage was found at the pressure instrumentation pipe attached to the exit pipe of No.1 charging pump of the purification system of a primary cooling system at JMTR in the Oarai Research establishment, JAERI. The Investigation Committee for Water Leakage from Instrumentation Pipe in JMTR was established and organized by specialists from inside and outside JAERI on December 16 and its meeting was held in public 3 times by 6th January, 2003. They found out the cause and countermeasures of cracks, and also investigated enhancement of safety management. As the result, it was considered that the leakage started around the 6th of December 2002 and the cause of the cracks was due to fatigue by vibration of the charging pump during operation. The committee discovered following incorrect actions in the safety management. First, operation of JMTR was continued without keeping careful watch in spite of occurrence of leakage detector alarm. Second, every time when the alarm range for the reasons other than the leakage, appropriate investigation and countermeasure were not taken. Third, the manager in charge didn't have a fair understanding of the situation and didn't give an appropriate direction. This is the report on the cause and countermeasures of cracks and enhancement of safety management. (author)

Issues behind Competitiveness and Carbon Leakage

Energy Technology Data Exchange (ETDEWEB)

NONE

2008-07-01

This report explores the vulnerability of heavy industry to carbon leakage and competitiveness loss. It reviews the existing literature on competitiveness and carbon leakage under uneven climate policies. It also suggests a statistical method to track carbon leakage, and applies this methodology to Phase I of the EU emissions trading scheme, for various industrial activities: iron and steel, cement, aluminium and refineries. Finally, it reviews measures to mitigate carbon leakage, as discussed in Australia, Canada, Europe, New Zealand and the US.

Radioactive particulate release associated with the DOT specification 6M container under hypothetical accident conditions

International Nuclear Information System (INIS)

Taylor, J.M.; Raney, P.J.

1986-02-01

A testing program was conducted to determine the leakage of depleted uranium dioxide powder (DUO) from the inner containment components of the US Department of Transportation's (DOT) specification 6M container under hypothetical accident conditions. Depleted uranium dioxide was selected as a surrogate for plutonium oxide because of the similarities in the powder characteristics, density and particle size, and because of the special handling and special facilities required for plutonium oxide. The DUO was packaged inside food pack cans in three different configurations inside the 2R vessel of the 6M container. The amount of DUO powder leakage ranged from none detectable ( -7 g) to a high of 1 x 10 -3 g. The combination of gravity, vibration and pressure produced the highest leakage of DUO. Containers that had hermetic seals (leak rates -4 atm cc/min) did not leak any detectable amount ( -7 g) of DUO under the test conditions. Impact forces had no effect on the leakage of particles with the packaging configurations used. 23 refs., 24 figs., 3 tabs

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.)

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)

An investigation of awareness on the Fukushima nuclear accident and Radioactive contamination

Energy Technology Data Exchange (ETDEWEB)

Ha, Jeong Chul; Song, Young Ju [Dept. of Consumer Safety, Korea Consumer Agency, Eumseong (Korea, Republic of)

2016-05-15

The objective of this study was to investigate Korean people's awareness about impact of the Fukushima nuclear power plant accident in Japan and radioactive contamination caused by it. The respondents of the survey were 600 adults who resided in the Seoul metropolitan area. The survey results show that the majority of respondents were concerned about impact of radiation leakage that might have an effect on our environment. They were worried about radioactive contamination of foodstuffs, particularly fishery products and preferred to acquire information through TV(49.8%) or the Internet(31.3%). Meanwhile, respondents mentioned that the information on the Fukushima nuclear accident and radioactive contamination had not been sufficient and they didn't know well about the follow-up measures of the government on the accident. Most respondents answered that information on radioactive contamination levels and safety of foods and environment was most needed. The results of this study could be useful to enhance awareness on radioactivity and improve risk communication on nuclear power plant accidents.

Pancreatic anastomosis leakage management following pancreaticoduodenectomy how could be manage the anastomosis leakage after pancreaticoduodenectomy?

Directory of Open Access Journals (Sweden)

Seyed Abbas Tabatabei

2015-01-01

Full Text Available Background: Pancreatic anastomosis leakage and fistula formation following pancreaticoduodenectomy (Whipple′s procedure is a common complication. Delay in timely diagnosis and proper management is associated with high morbidity and mortality. To report our experience with management of pancreatic fistula following Whipple′s procedure. Materials and Methods: In this retrospective study, medical records of 90 patients who underwent Whipple′s procedure from 2009 to 2013 at our medical center were reviewed for documents about pancreatic anastomosis leakage and fistula formation. Results: There were 15 patients who developed pancreatico-jejunal anastomosis leakage. In 6 patients (3 males and 3 females the leakage was mild (conservative therapy was administered, but in 9 patients (6 males and 3 females, there was severe leakage. For the latter group, surgical intervention was done (2 cases underwent re-anastomosis and for 7 cases pancreatico-jejunal stump ligation was done along with drainage of the location. Conclusion: In severe pancreatic anastomotic leakage, it is better to intervene surgically as soon as possible by debridement of the distal part of the pancreas and ligation of the stump with nonabsorbable suture. Furthermore, debridement of the jejunum should be done, and the stump should be ligated thoroughly along with drainage.

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.

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

Risk analysis of emergent water pollution accidents based on a Bayesian Network.

Science.gov (United States)

Tang, Caihong; Yi, Yujun; Yang, Zhifeng; Sun, Jie

2016-01-01

To guarantee the security of water quality in water transfer channels, especially in open channels, analysis of potential emergent pollution sources in the water transfer process is critical. It is also indispensable for forewarnings and protection from emergent pollution accidents. Bridges above open channels with large amounts of truck traffic are the main locations where emergent accidents could occur. A Bayesian Network model, which consists of six root nodes and three middle layer nodes, was developed in this paper, and was employed to identify the possibility of potential pollution risk. Dianbei Bridge is reviewed as a typical bridge on an open channel of the Middle Route of the South to North Water Transfer Project where emergent traffic accidents could occur. Risk of water pollutions caused by leakage of pollutants into water is focused in this study. The risk for potential traffic accidents at the Dianbei Bridge implies a risk for water pollution in the canal. Based on survey data, statistical analysis, and domain specialist knowledge, a Bayesian Network model was established. The human factor of emergent accidents has been considered in this model. Additionally, this model has been employed to describe the probability of accidents and the risk level. The sensitive reasons for pollution accidents have been deduced. The case has also been simulated that sensitive factors are in a state of most likely to lead to accidents. Copyright © 2015 Elsevier Ltd. All rights reserved.

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)

Methane leakage in natural gas operations

International Nuclear Information System (INIS)

Jennervik, A.

1992-01-01

The world gas industry is efficient in conservation of natural gas within its systems. As the influence of methane as an infra-red absorbent gas has been more widely recognized, the considerations of methane's greenhouse effect has become vitally important to gas companies around the world. The industry is universally environmentally conscious. natural gas transmission and distribution companies want to maintain their image as suppliers of clean fuel. Further reductions in methane leakage --- particularly in older distribution systems --- can, should and will be pursued. Unfortunately, there has been little exchange of views on methane leakages between commentators on environmental matters and gas companies and organizations. There is absolutely no need for the industry to avoid the issue of greenhouse gases. Without industry involvement, the environmental debate concerning fossil fuels could lead to selective interpretation of scientific views and available evidence. Companies and authorities would be presented with confusing, contradictory evidence on which to base policy approaches and regulations

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)

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.

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

Cooperation of nuclear, thermal and hydroelectric power plants in the power system

International Nuclear Information System (INIS)

1984-01-01

The conference heard 36 papers of which 23 were incorporated in INIS. The subjects discussed were: the development of power industry in Czechoslovakia, methods of statistical analysis of data regarding nuclear power plant operation, the incorporation of WWER nuclear power plants in the power supply system, the standardization of nuclear power plants, the service life of components, use of nuclear energy sources, performance of the reactor accident protection system, the use of nuclear power and heating plants in Hungary, risk analysis, optimization of nuclear power plants, accidents caused by leakage of the primary and secondary circuit. (J.P.)

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)

Method for detecting resin leakage in LWR coolant

International Nuclear Information System (INIS)

Girard, J.E.

1988-05-01

Resin leakage from condensate polishing units can result in steam generator corrosion. This report describes the development of a resin leakage detection method based in analyzing the organic breakdown products released from resin on heating. The breakdown products are analyzed using high performance liquid chromatography (HPLC) with fluorescence detection. Some of the organic products formed have been identified. A design for a resin monitoring unit, suitable for incorporation into the IONTRAC system, is presented. Theoretically, detection of ppB levels of resin by processing about one liter of water, is possible

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

International Nuclear Information System (INIS)

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

1990-06-01

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

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.

Control room habitability during severe accidents

International Nuclear Information System (INIS)

Siu, R.P.

1989-01-01

The requirements for protection of control room personnel against radiation hazards are specified in 10CFR50, Appendix A, GDC 19. The conventional approach involves a mechanistic evaluation of the radiation doses to control room personnel during design-basis accidents. In this study, an assessment of control room habitability during severe accidents is conducted. The potential levels of radiation hazards to control room personnel are evaluated in terms of both magnitude and probability of occurrence. The expected values for the probabilities of exceeding GDC-19 limits and the cumulative probability distributions of control room doses are determined. In this study, a pressurized water reactor with a large dry containment has been selected for analysis. The types of control rooms evaluated in this study include designs with: (a) filtered local intakes only, (b) filtered recirculation only, (c) filtered local intakes and recirculation, and (d) filtered dual remote intakes and recirculation. From the observations, it is concluded that, except for control room D, all other control room designs may require improvements in order to provide adequate radiation protection during severe accidents, particularly in terms of reducing whole-body gamma doses and skin doses. Potential design improvements include reduction of intake flows for concepts relying on pressurization, reduction in overall leakages, and control room pressurization through the use of bottled air supply

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

ERROR REDUCTION IN DUCT LEAKAGE TESTING THROUGH DATA CROSS-CHECKS

Energy Technology Data Exchange (ETDEWEB)

ANDREWS, J.W.

1998-12-31

One way to reduce uncertainty in scientific measurement is to devise a protocol in which more quantities are measured than are absolutely required, so that the result is over constrained. This report develops a method for so combining data from two different tests for air leakage in residential duct systems. An algorithm, which depends on the uncertainty estimates for the measured quantities, optimizes the use of the excess data. In many cases it can significantly reduce the error bar on at least one of the two measured duct leakage rates (supply or return), and it provides a rational method of reconciling any conflicting results from the two leakage tests.

Pulse sequences for suppressing leakage in single-qubit gate operations

Science.gov (United States)

Ghosh, Joydip; Coppersmith, S. N.; Friesen, Mark

2017-06-01

Many realizations of solid-state qubits involve couplings to leakage states lying outside the computational subspace, posing a threat to high-fidelity quantum gate operations. Mitigating leakage errors is especially challenging when the coupling strength is unknown, e.g., when it is caused by noise. Here we show that simple pulse sequences can be used to strongly suppress leakage errors for a qubit embedded in a three-level system. As an example, we apply our scheme to the recently proposed charge quadrupole (CQ) qubit for quantum dots. These results provide a solution to a key challenge for fault-tolerant quantum computing with solid-state elements.

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.

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

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

Review on water leakage control in distribution networks and the associated environmental benefits.

Science.gov (United States)

Xu, Qiang; Liu, Ruiping; Chen, Qiuwen; Li, Ruonan

2014-05-01

Water supply is the primary element of an urban system. Due to rapid urbanization and water scarcity, maintaining a stable and safe water supply has become a challenge to many cities, whereas a large amount of water is lost from the pipes of distribution systems. Water leakage is not only a waste of water resources, but also incurs great socio-economic costs. This article presents a comprehensive review on the potential water leakage control approaches and specifically discusses the benefits of each to environmental conservation. It is concluded that water leakage could be further reduced by improving leakage detection capability through a combination of predictive modeling and monitoring instruments, optimizing pipe maintenance strategy, and developing an instant pressure regulation system. The environment could benefit from these actions because of water savings and the reduction of energy consumption as well as greenhouse gas emissions. Copyright © 2014 The Research Centre for Eco-Environmental Sciences, Chinese Academy of Sciences. Published by Elsevier B.V. All rights reserved.

Radionuclide leakage monitoring during hyperthermic isolated limb perfusion for treatment of local melanoma metastasis in an extremity

DEFF Research Database (Denmark)

Paulsen, Ida F; Chakera, Annette Hougaard; Schmidt, Grethe

2015-01-01

(99m) Tc-labelled tracer infused into the isolated limb circulation. RESULTS: One hundred and sixteen of 131 procedures were completed. In 13%, a leakage of ≥10% was detected; in 6% (n = 8), the cytotoxic drug was never infused because of constant leakage; in 7% (n = 9), leakage ≥10% was measured......INTRODUCTION: The aim is to describe the importance of leakage monitoring in hyperthermic isolated limb perfusion (ILP). It is generally recommended that leakage should not exceed 10% because of risk of systemic toxicity. MATERIAL AND METHODS: Data retrieved by retrospective analysis of 131...... were observed in 14%. Three of the patients with leakage ≥10% were successfully treated in a repeated procedure. CONCLUSION: Leakage monitoring using a threshold of 10% during ILP saves the patients from systemic toxicity, however, at the expense of early termination or cancellation of ILP treatment...

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

Leakage and sweet spots in triple-quantum-dot spin qubits: A molecular-orbital study

Science.gov (United States)

Zhang, Chengxian; Yang, Xu-Chen; Wang, Xin

2018-04-01

A triple-quantum-dot system can be operated as either an exchange-only qubit or a resonant-exchange qubit. While it is generally believed that the decisive advantage of the resonant-exchange qubit is the suppression of charge noise because it is operated at a sweet spot, we show that the leakage is also an important factor. Through molecular-orbital-theoretic calculations, we show that when the system is operated in the exchange-only scheme, the leakage to states with double electron occupancy in quantum dots is severe when rotations around the axis 120∘ from z ̂ is performed. While this leakage can be reduced by either shrinking the dots or separating them further, the exchange interactions are also suppressed at the same time, making the gate operations unfavorably slow. When the system is operated as a resonant-exchange qubit, the leakage is three to five orders of magnitude smaller. We have also calculated the optimal detuning point which minimizes the leakage for the resonant-exchange qubit, and have found that although it does not coincide with the double sweet spot for the charge noise, they are rather close. Our results suggest that the resonant-exchange qubit has another advantage, that leakage can be greatly suppressed compared to the exchange-only qubit, and operating at the double sweet spot point should be optimal both for reducing charge noise and suppressing leakage.

Radiological environment within an NPP after a severe nuclear accident

Science.gov (United States)

Andgren, Karin; Fritioff, Karin; Buhr, Anna Maria Blixt; Huutoniemi, Tommi

2017-09-01

The radiological environment following a severe nuclear accident can be visualised on building layouts. The direct radiation in an area (or room) can be visualized on the layout by a colouring scheme depending on the dose rate level (for example orange for high gamma dose rate level and purple for an intermediate gamma dose rate level). Following the Fukushima accident, a need for update of these layouts has been identified at the Swedish nuclear power plant of Forsmark. Shielding calculations for areas where access is desired for severe accident management have been performed. Many different sources of radiation together with different types of shielding material contribute to the dose that would be received by a person entering the area. External radiation from radioactivity within e.g. pipes and components is considered and also external radiation from radioactivity in the air (originating from diffuse leakage of the containment atmosphere). Results are presented as dose rates for relevant dose points together with a method for estimating the dose rate levels for each of the rooms of the reactor building.

Radiological environment within an NPP after a severe nuclear accident

Directory of Open Access Journals (Sweden)

Andgren Karin

2017-01-01

Full Text Available The radiological environment following a severe nuclear accident can be visualised on building layouts. The direct radiation in an area (or room can be visualized on the layout by a colouring scheme depending on the dose rate level (for example orange for high gamma dose rate level and purple for an intermediate gamma dose rate level. Following the Fukushima accident, a need for update of these layouts has been identified at the Swedish nuclear power plant of Forsmark. Shielding calculations for areas where access is desired for severe accident management have been performed. Many different sources of radiation together with different types of shielding material contribute to the dose that would be received by a person entering the area. External radiation from radioactivity within e.g. pipes and components is considered and also external radiation from radioactivity in the air (originating from diffuse leakage of the containment atmosphere. Results are presented as dose rates for relevant dose points together with a method for estimating the dose rate levels for each of the rooms of the reactor building.

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)

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

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

Differences in Ostomy Pouch Seal Leakage Occurrences Between North American and European Residents.

Science.gov (United States)

Fellows, Jane; Forest Lalande, Louise; Martins, Lina; Steen, Anne; Størling, Zenia M

The purpose of this study was to compare experiences and concerns about pouch seal leakage between persons with ostomies residing in North America (Canada and the United States) and Europe (United Kingdom, Netherlands, Sweden, Germany, Belgium, France, and Italy). Differences in reported pouch wear time and accessories used between the 2 groups were also examined. Secondary analysis of data from a cross-sectional study (Ostomy Life Study). Responses from persons residing in European countries (n = 1939) were compared with responses of 1387 individuals residing in North American countries. Persons with an ostomy completed a questionnaire that focused on 4 topics related to the daily use of an ostomy pouching system (pouch seal leakage, ballooning, appearance of pouching system such as color and size of the pouch and whether it is discrete under clothing, and coupling failure of 2-piece pouching systems). Pouch seal leakage was defined as stomal effluent seeping between the skin and the wafer of the ostomy pouching system. Statistical analysis was performed using a proportional odds model including various variable effects. Special attention was given to frequency of pouch seal leakage occurrences. All tests were 2-sided; P values ≤.05 were deemed statistically significant. Participants living in the North American countries indicated they were more likely to experience leakage from the ostomy (odds ratio = 2.610, 95% CI 2.187-3.115; P < .0001). Findings also indicated they were more likely to worry about pouch seal leakage than those in the European countries' data set (odds ratio = 2.722, 95% CI 2.283-3.246; P < .0001). Participants residing in the North American countries had significantly longer wear times than those participants in the European countries (P < .0001, χ test). The use of accessories was associated with a longer pouching system wear time. Study results suggest that participants from the North American countries indicated significantly more

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

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

Mission analysis report for single-shell tank leakage mitigation

International Nuclear Information System (INIS)

Cruse, J.M.

1994-01-01

This document provides an analysis of the leakage mitigation mission applicable to past and potential future leakage from the Hanford Site's 149 single-shell high-level waste tanks. This mission is a part of the overall missions of the Westinghouse Hanford Company Tank Waste Remediation System division to remediate the tank waste in a safe and acceptable manner. Systems engineers principles are being applied to this effort. Mission analysis supports early decision making by clearly defining program objectives. This documents identifies the initial conditions and acceptable final conditions, defines the programmatic and physical interfaces and constraints, estimates the resources to carry out the mission, and establishes measures of success. The results of the mission analysis provide a consistent basis for subsequent systems engineering work

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...

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

How to attribute market leakage to CDM projects

NARCIS (Netherlands)

Vöhringer, F.; Kuosmanen, T.K.; Dellink, R.B.

2006-01-01

Economic studies suggest that market leakage rates of greenhouse gas abatement can reach the two-digit percentage range. Although the Marrakesh Accords require Clean Development Mechanism (CDM) projects to account for leakage, most projects neglect market leakage. Insufficient leakage accounting is

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

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

Review of Atomic Energy Laws Related to Radiological Accidents and Methods of Improvement

Energy Technology Data Exchange (ETDEWEB)

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 [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2009-05-15

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

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

Preventive testing and leakage detection in pipe-lines of steam condensers and generators of a PWR type reactor

International Nuclear Information System (INIS)

Canalini, A.; Carvalho, N.C. de

1985-01-01

The non-destructive methods: Spum, Helium and Hydrostatic used in leakage detection in condenser pipelines for PWR type reactors are presented. The time, costs, sensitivity, resources necessary and personnel development factors are considered to choose adequated method, in function of nuclear power plant conditions. The leakage tests are applied in pressurized systems or vacuum. Eddy Current testing is used in condensers and steam generators aiming to avoid leakage in these equipments. The spume testing for leakage detection in condenser pipelines - which operation - and hydrostatic testing for leakage detection through reaming with shutdown - were most efficients. The Helium testing applied in pressurized systems or submitted to vacuum systems presented satisfactory results. The Eddy Current testing in condenser and steam generator pipelines reached desired objective, reducing leakage in the first and preserving the integrity in the second. (M.C.K.) [pt

PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation

Energy Technology Data Exchange (ETDEWEB)

Hadad, Kamal; Esmaeili-Sanjavanmareh, Mansour [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering

2017-05-15

PCTRAN capability to simulate a large break loss of coolant accident concurrent with the loss of offsite power in Bushehr Nuclear Power Plant is enhanced and investigated. Following the correction of the accident scenario for Bushehr nuclear power plant in PCTRAN, simulation results are compared with the final safety assessment report of that plant. As a result, the primary loop thermal hydraulics parameters including pressure, total flow rates, leakage flow rates and reactor power are in a good agreement with the reference data. Hot and cold leg temperature variations have the same trends as reference data but have a maximum of 80 C disagreement at the transient initiation. The reason for this disagreement is explained and its adjustment is discussed. Improvements of PCTRAN simulator are mainly due to enhancing user control for atmospheric steam dump valve, containment pressure and emergency core cooling systems which are thoroughly described in this paper.

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)

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

Excessive leakage measurement using pressure decay method in containment building local leakage rate test at nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Kyu; Kim, Chang Soo; Kim, Wang Bae [KHNP, Central Research Institute, Daejeon (Korea, Republic of)

2016-06-15

There are two methods for conducting the containment local leakage rate test (LLRT) in nuclear power plants: the make-up flow rate method and the pressure decay method. The make-up flow rate method is applied first in most power plants. In this method, the leakage rate is measured by checking the flow rate of the make-up flow. However, when it is difficult to maintain the test pressure because of excessive leakage, the pressure decay method can be used as a complementary method, as the leakage rates at pressures lower than normal can be measured using this method. We studied the method of measuring over leakage using the pressure decay method for conducting the LLRT for the containment building at a nuclear power plant. We performed experiments under conditions similar to those during an LLRT conducted on-site. We measured the characteristics of the leakage rate under varies pressure decay conditions, and calculated the compensation ratio based on these data.

Calculation of Leakage Inductance for High Frequency Transformers

DEFF Research Database (Denmark)

Ouyang, Ziwei; Jun, Zhang; Hurley, William Gerard

2015-01-01

Frequency dependent leakage inductance is often observed. High frequency eddy current effects cause a reduction in leakage inductance. The proximity effect between adjacent layers is responsible for the reduction of leakage inductance. This paper gives a detailed analysis of high frequency leakag...

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.)

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

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)

Root cause evaluation of pressurizer relief valve leakage

International Nuclear Information System (INIS)

Olson, D.E.; Voll, B.J.

1996-01-01

Pressurizer relief valves at two pressurized water reactor units experienced unacceptable leakage during plant heatup. The leakage was suspected to be caused by excessive pipe loads on the valves. This paper describes how monitoring via hard-wired transducers and a digital data acquisition system was used to quantify the pipe loads on the valves, and assist in determining the root cause of the pipe loads and appropriate corrective actions. The selection of the parameters monitored, how the monitoring was accomplished and interpretation of the results is discussed. The corrective actions implemented based on the monitoring results are also discussed

Analytical validation of operator actions in case of primary to secondary leakage for VVER-1000/V320

Energy Technology Data Exchange (ETDEWEB)

Andreeva, M., E-mail: m_andreeva@inrne.bas.bg; Groudev, P., E-mail: pavlinpg@inrne.bas.bg; Pavlova, M., E-mail: pavlova@inrne.bas.bg

2015-12-15

Highlights: • We validate operator actions in case of primary to secondary leakage. • We perform four scenarios related to SGTR accident for VVER-1000/V320. • The reference power plant for the analyses is Unit 6 at Kozloduy NPP. • The RELAP5/MOD 3.2 computer code is used in performing the analyses. • The analyses confirm the effectiveness of operator actions during PRISE. - Abstract: This paper presents the results of analytical validation of operator actions in case of “Steam Generator Tube Rupture” (SGTR) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP), done during the development of Symptom Based Emergency Operating Procedures (SB EOPs) for this plant. The purpose of the analyses is to demonstrate the ability to terminate primary to secondary leakage and to indicate an effective strategy for preventing secondary leakage to the environment and in this way to prevent radiological release to the environment. Following depressurization and cooldown of reactor coolant system (RCS) with isolation of the affected steam generator (SG), in these analyses are validated options for post-SGTR cooldown by: • back up filling the ruptured SG; • using letdown system in the affected SG and • by opening Fast Acting Isolation Valve (FAIV) and using Steam Dump Facility to the Condenser (BRU-K). The results of the thermal-hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs. The RELAP5/MOD3.2 computer code has been used for the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the thermal-hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS). This paper is possible through the participation of leading specialists from KNPP.

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.

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

International Nuclear Information System (INIS)

Sharafutdinov, Rachet; Lankin, Michail; Kharitonova, Nataliya

2014-01-01

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

Exploitation of Unintentional Information Leakage from Integrated Circuits

Science.gov (United States)

Cobb, William E.

2011-01-01

The information leakage of electronic devices, especially those used in cryptographic or other vital applications, represents a serious practical threat to secure systems. While physical implementation attacks have evolved rapidly over the last decade, relatively little work has been done to allow system designers to effectively counter the…

Control room habitability assessment and in-leakage test for Korean NPP - 15510

International Nuclear Information System (INIS)

Song, D.S.; Lee, J.B.; Ha, S.J.; Seong, J.J.

2015-01-01

The assessment of control room habitability for Wolsung unit 1 was performed based on GL 2003-01 and Reg. Guide 1.197. The control room habitability program including Control Room Envelope (CRE) in-leakage test procedures, self assessment guidance, CRE boundary control program, CRE maintenance/sealing program was developed for Wolsung unit 1. The integrated CRE test was performed utilizing ASTM E741. There are two operating modes of pressurization and isolation for CRE ventilation test, and four tests were performed using each of the control room HVAC sub-trains. The control room HVAC system lineup of pressurization mode test was based upon a lineup that encompassed the design basis radiological analyses. The other control room HVAC system lineup of isolation mode test was based on an operation mode that considers toxic gas. The in-leakage testing was performed in accordance with CRE in-leakage test procedures. In the pressurization mode, measured unfiltered in-leakage rates for train A and train B were 0 CFM and 405 CFM respectively. In the isolation mode, measured unfiltered in-leakage rates for train A and train B were 1,739 CFM and 1,502 CFM, respectively. Maximum concentration of ammonia at the control room HVAC intake is calculated to be 0.027 g/m 3 (39 ppm), and satisfied the toxicity limit of 300 ppm. The test result shows that the measured unfiltered in-leakage is bounded by the regulatory criteria assumed in the design basis radiological analyses. (authors)

Simulation of leakage through mechanical sealing device

Science.gov (United States)

Tikhomorov, V. P.; Gorlenko, O. A.; Izmerov, M. A.

2018-03-01

The procedure of mathematical modeling of leakage through the mechanical seal taking into account waviness and roughness is considered. The percolation process is represented as the sum of leakages through a gap between wavy surfaces and percolation through gaps formed by fractal roughness, i.e. the total leakage is determined by the slot model and filtration leakage. Dependences of leaks on the contact pressure of corrugated and rough surfaces of the mechanical seal elements are presented.

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

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)