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

International Nuclear Information System (INIS)

Rubin, A.M.

1986-01-01

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

Pressurized-water-reactor station blackout

International Nuclear Information System (INIS)

Dobbe, C.A.

1983-01-01

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

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

International Nuclear Information System (INIS)

Volkanovski, Andrija

2015-01-01

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

Accident Analysis of Chinese CPR1000 in Response to Station Blackout

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Estimation of station blackout frequency in FBTR

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

1991-04-01

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

Reactor coolant pump shaft seal behavior during station blackout

International Nuclear Information System (INIS)

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

1985-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-15

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

The Fukushima Accident: A Station Blackout and the Consequences

International Nuclear Information System (INIS)

Schäfer, F.; Tusheva, P.; Kliem, S.

2012-01-01

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

Probabilistic assessment of the Juragua NPP response under Station Blackout conditions

International Nuclear Information System (INIS)

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

1996-01-01

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

Station blackout with reactor coolant pump seal leakage

International Nuclear Information System (INIS)

Evinay, A.

1993-01-01

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

Design Provisions for Station Blackout at Nuclear Power Plants

International Nuclear Information System (INIS)

Duchac, Alexander

2015-01-01

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

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

International Nuclear Information System (INIS)

Baranowsky, P.W.

1985-05-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

Bang, Young Seok; Kim, Hho Jung

1992-01-01

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

Station blackout and public confidence: a cautionary tale

International Nuclear Information System (INIS)

Cave, L.

1990-01-01

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

International Nuclear Information System (INIS)

Valle Cepero, Reinaldo

2004-01-01

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

Severe accident sequences simulated at the Grand Gulf Nuclear Station

International Nuclear Information System (INIS)

Carbajo, J.J.

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1988-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-01

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

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

International Nuclear Information System (INIS)

Kaliatka, A.

2007-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-07-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Analysis of Peach Bottom station blackout with MELCOR

International Nuclear Information System (INIS)

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

1987-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-09-15

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

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

International Nuclear Information System (INIS)

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

2011-11-01

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

Benchmarking Simulation of Long Term Station Blackout Events

International Nuclear Information System (INIS)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-11-15

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Cooling methods of station blackout scenario for LWR plants

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

Cooling methods of station blackout scenario for LWR plants

International Nuclear Information System (INIS)

2012-01-01

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

Preliminary Evaluations of CSPACE for a Station Blackout Transient in APR1400

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

MELCOR simulation of long-term station blackout at Peach Bottom

International Nuclear Information System (INIS)

Madni, I.K.

1990-01-01

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

Analysis of station blackout accidents for the Bellefonte pressurized water reactor

International Nuclear Information System (INIS)

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

1986-09-01

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

Alcohol-Induced Blackout

Directory of Open Access Journals (Sweden)

Dai Jin Kim

2009-11-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

International Nuclear Information System (INIS)

Carbajo, J.J.

1995-06-01

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

Main considerations for modelling a station blackout scenario with trace

International Nuclear Information System (INIS)

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

2017-01-01

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

Main considerations for modelling a station blackout scenario with trace

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

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

International Nuclear Information System (INIS)

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

1993-11-01

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

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

International Nuclear Information System (INIS)

Barrett, A.J.; Marquino, W.

2013-01-01

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

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

Directory of Open Access Journals (Sweden)

Andrej Prošek

2013-01-01

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

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

International Nuclear Information System (INIS)

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

1989-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-08-01

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

Station Blackout Analysis of HTGR-Type Experimental Power Reactor

Science.gov (United States)

Syarip; Zuhdi, Aliq; Falah, Sabilul

2018-01-01

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

Extended Station Blackout Analysis for VVER-1000 MWe Reactor

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

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

International Nuclear Information System (INIS)

El-Sahlamy, N.M.

2017-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Study of a Station Blackout Event in the PWR Plant

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

Park, Sang Gil; Kim, Han Chul

2012-01-01

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

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

International Nuclear Information System (INIS)

2014-01-01

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

Reactor coolant pump shaft seal behavior during blackout conditions

International Nuclear Information System (INIS)

Mings, W.J.

1985-01-01

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

Station blackout with failure of wired shutdown system for AHWR

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Large blackouts in North America: Historical trends and policy implications

International Nuclear Information System (INIS)

Hines, Paul; Apt, Jay; Talukdar, Sarosh

2009-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

Bayless, P.D.

1988-10-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Investigation of a Station Blackout Scenario with the ATLAS Test

International Nuclear Information System (INIS)

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

2012-01-01

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

47 CFR 76.111 - Cable sports blackout.

Science.gov (United States)

2010-10-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

47 CFR 76.127 - Satellite sports blackout.

Science.gov (United States)

2010-10-01

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

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

International Nuclear Information System (INIS)

Prosek, A.

2016-01-01

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

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

International Nuclear Information System (INIS)

Krishnamoorthy, S; Close, S

2017-01-01

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

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

International Nuclear Information System (INIS)

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

1990-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Fractal Characteristics Analysis of Blackouts in Interconnected Power Grid

DEFF Research Database (Denmark)

Wang, Feng; Li, Lijuan; Li, Canbing

2018-01-01

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

Study on self organized criticality of China power grid blackouts

Energy Technology Data Exchange (ETDEWEB)

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

2009-03-15

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

Study on self organized criticality of China power grid blackouts

Energy Technology Data Exchange (ETDEWEB)

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

2009-03-15

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

Alcohol-induced blackout as a criminal defense or mitigating factor: an evidence-based review and admissibility as scientific evidence.

Science.gov (United States)

Pressman, Mark R; Caudill, David S

2013-07-01

Alcohol-related amnesia--alcohol blackout--is a common claim of criminal defendants. The generally held belief is that during an alcohol blackout, other cognitive functioning is severely impaired or absent. The presentation of alcohol blackout as scientific evidence in court requires that the science meets legal reliability standards (Frye, FRE702/Daubert). To determine whether "alcohol blackout" meets these standards, an evidence-based analysis of published scientific studies was conducted. A total of 26 empirical studies were identified including nine in which an alcohol blackout was induced and directly observed. No objective or scientific method to verify the presence of an alcoholic blackout while it is occurring or to confirm its presence retrospectively was identified. Only short-term memory is impaired and other cognitive functions--planning, attention, and social skills--are not impaired. Alcoholic blackouts would not appear to meet standards for scientific evidence and should not be admissible. © 2013 American Academy of Forensic Sciences.

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

Directory of Open Access Journals (Sweden)

Roman Mukin

2018-04-01

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

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

International Nuclear Information System (INIS)

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

1999-12-01

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

Calculation of spent fuel pool severe accident with MELCOR

International Nuclear Information System (INIS)

Deng Jian; Xiang Qing'an; Zhou Kefeng

2014-01-01

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

MELCOR DB Construction for the Severe Accident Analysis DB

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-03-01

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

Transnational infrastructure vulnerability : the historical shaping of the 2006 European 'blackout'

NARCIS (Netherlands)

Vleuten, van der E.B.A.; Lagendijk, V.C.

2010-01-01

The "European Blackout" of 4 November 2006 is a key reference in current debates on transnational electricity infrastructure vulnerability and governance. Several commentators have observed that to understand what happened, one must look at history. Our paper answers this call and demonstrates how

Radionuclide release calculations for selected severe accident scenarios

International Nuclear Information System (INIS)

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

1990-08-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Extended blackout mitigation strategy for PWR

International Nuclear Information System (INIS)

Prošek, Andrej; Volkanovski, Andrija

2015-01-01

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

Extended blackout mitigation strategy for PWR

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-15

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

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

Energy Technology Data Exchange (ETDEWEB)

Hakansson, K

2004-02-01

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

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

International Nuclear Information System (INIS)

Hakansson, K.

2004-02-01

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

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

Science.gov (United States)

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

2017-03-01

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

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

Science.gov (United States)

Manning, Robert M.

2017-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Sustainability from the Occurrence of Critical Dynamic Power System Blackout Determined by Using the Stochastic Event Tree Technique

Directory of Open Access Journals (Sweden)

Muhammad Murtadha Othman

2017-06-01

Full Text Available With the advent of advanced technology in smart grid, the implementation of renewable energy in a stressed and complicated power system operation, aggravated by a competitive electricity market and critical system contingencies, this will inflict higher probabilities of the occurrence of a severe dynamic power system blackout. This paper presents the proposed stochastic event tree technique used to assess the sustainability against the occurrence of dynamic power system blackout emanating from implication of critical system contingencies such as the rapid increase in total loading condition and sensitive initial transmission line tripping. An extensive analysis of dynamic power system blackout has been carried out in a case study of the following power systems: IEEE RTS-79 and IEEE RTS-96. The findings have shown that the total loading conditions and sensitive transmission lines need to be given full attention by the utility to prevent the occurrence of dynamic power system blackout.

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

Science.gov (United States)

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

2011-01-01

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

Computer chaos and the blackout

CERN Multimedia

Malik, Rex

1971-01-01

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

Mitigating reentry radio blackout by using a traveling magnetic field

Directory of Open Access Journals (Sweden)

Hui Zhou

2017-10-01

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

Mitigating reentry radio blackout by using a traveling magnetic field

Science.gov (United States)

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

2017-10-01

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

Impact assessment of the 1977 New York City blackout. Final report

Energy Technology Data Exchange (ETDEWEB)

Corwin, J. L.; Miles, W. T.

1978-07-01

This study was commissioned by the Division of Electric Energy Systems (EES), Department of Energy (DOE) shortly after the July 13, 1977 New York City Blackout. The objectives were two-fold: to assess the availability and collect, where practical, data pertaining to a wide variety of impacts occurring as a result of the blackout; and to broadly define a framework to assess the value of electric power reliability from consideration of the blackout and its effects on individuals, businesses, and institutions. The impacts were complex and included both economic and social costs. In order to systematically classify the most significant of these impacts and provide guidance for data collection, impact classification schemes were developed. Major economic impact categories examined are business; government; utilities (Consolidated Edison); insurance industry; public health services; and other public services. Impacts were classified as either direct or indirect depending upon whether the impact was due to a cessation of electricity or a response to that cessation. The principal economic costs of the blackout are shown. Social impacts, i.e., the changes in social activities and adaptations to these changes were particularly significant in New York due to its unique demographic and geographic characteristics. The looting and arson that accompanied the blackout set aside the NYC experience from other similar power failures. (MCW)

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

Science.gov (United States)

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

2018-06-01

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

Blackout cloth for dormancy induction

Science.gov (United States)

Tom Jopson

2007-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

Altkind, Franz; Schmid, Daniel

2015-01-01

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

Severe accident analysis for level 2 PSA of SMART reactor

Energy Technology Data Exchange (ETDEWEB)

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

2010-12-15

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

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

International Nuclear Information System (INIS)

Zhang Yanzhao; Zhang Fan; Zhao Xinwen; Zheng Yingfeng

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-15

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

Learning from the blackouts. Transmission system security in competitive electricity markets

Energy Technology Data Exchange (ETDEWEB)

none

2005-07-01

Electricity market reform has fundamentally changed the environment for maintaining reliable and secure power supplies. Growing inter-regional trade has placed new demands on transmission systems, creating a more integrated and dynamic network environment with new real-time challenges for reliable and secure transmission system operation. Despite these fundamental changes, system operating rules and practices remain largely unchanged. The major blackouts of 2003 and 2004 raised searching questions about the appropriateness of these arrangements. Management of system security needs to be transformed to maintain reliable electricity services in this more dynamic operating environment. These challenges raise fundamental issues for policymakers. This publication presents case studies drawn from recent large-scale blackouts in Europe, North America, and Australia. It concludes that a comprehensive, integrated policy response is required to avoid preventable large-scale blackouts in the future.

Control and prediction for blackouts caused by frequency collapse in smart grids.

Science.gov (United States)

Wang, Chengwei; Grebogi, Celso; Baptista, Murilo S

2016-09-01

The electric power system is one of the cornerstones of modern society. One of its most serious malfunctions is the blackout, a catastrophic event that may disrupt a substantial portion of the system, playing havoc to human life and causing great economic losses. Thus, understanding the mechanisms leading to blackouts and creating a reliable and resilient power grid has been a major issue, attracting the attention of scientists, engineers, and stakeholders. In this paper, we study the blackout problem in power grids by considering a practical phase-oscillator model. This model allows one to simultaneously consider different types of power sources (e.g., traditional AC power plants and renewable power sources connected by DC/AC inverters) and different types of loads (e.g., consumers connected to distribution networks and consumers directly connected to power plants). We propose two new control strategies based on our model, one for traditional power grids and another one for smart grids. The control strategies show the efficient function of the fast-response energy storage systems in preventing and predicting blackouts in smart grids. This work provides innovative ideas which help us to build up a robuster and more economic smart power system.

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-10-01

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

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

International Nuclear Information System (INIS)

Hodge, S.A.

1990-01-01

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

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

International Nuclear Information System (INIS)

Baron, J.; Bastianelli, B.

1997-01-01

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

Acute alcohol effects on narrative recall and contextual memory: an examination of fragmentary blackouts.

Science.gov (United States)

Wetherill, Reagan R; Fromme, Kim

2011-08-01

The present study examined the effects of alcohol consumption on narrative recall and contextual memory among individuals with and without a history of fragmentary blackouts in an attempt to better understand why some individuals experience alcohol-induced memory impairments whereas others do not, even at comparable blood alcohol concentrations (BACs). Standardized beverage (alcohol and no alcohol) administration procedures and neuropsychological assessments measured narrative recall and context memory performance before and after alcohol consumption in individuals with (n=44) and without (n=44) a history of fragmentary blackouts. Findings indicate that acute alcohol intoxication led to impairments in free recall, but not next-day cued recall. Further, participants showed similar memory performance when sober, but individuals who consumed alcohol and had a positive history of fragmentary blackouts showed greater contextual memory impairments than those who had not previously experienced a fragmentary blackout. Thus, it appears that some individuals may have an inherent vulnerability to alcohol-induced memory impairments due to alcohol's effects on contextual memory processes. Copyright © 2011 Elsevier Ltd. All rights reserved.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-08-15

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

Severe accident sequence assessment for boiling water reactors: program overview

International Nuclear Information System (INIS)

Fontana, M.H.

1980-10-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Evaluating the effect of a campus-wide social norms marketing intervention on alcohol-use perceptions, consumption, and blackouts.

Science.gov (United States)

Su, Jinni; Hancock, Linda; Wattenmaker McGann, Amanda; Alshagra, Mariam; Ericson, Rhianna; Niazi, Zackaria; Dick, Danielle M; Adkins, Amy

2018-04-01

To evaluate the effect of a campus-wide social norms marketing intervention on alcohol-use perceptions, consumption, and blackouts at a large, urban, public university. 4,172 college students (1,208 freshmen, 1,159 sophomores, 953 juniors, and 852 seniors) who completed surveys in Spring 2015 for the Spit for Science Study, a longitudinal study of students' substance use and emotional health. Participants were e-mailed an online survey that queried campaign readership, perception of peer alcohol use, alcohol consumption, frequency of consumption, and frequency of blackouts. Associations between variables were evaluated using path analysis. We found that campaign readership was associated with more accurate perceptions of peer alcohol use, which, in turn, was associated with self-reported lower number of drinks per sitting and experiencing fewer blackouts. This evaluation supports the use of social norms marketing as a population-level intervention to correct alcohol-use misperceptions and reduce blackouts.

The use of influence diagrams for evaluating severe accident management strategies

International Nuclear Information System (INIS)

Jae, M.; Apostolakis, G.E.

1992-01-01

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

Advanced validation of CFD-FDTD combined method using highly applicable solver for reentry blackout prediction

International Nuclear Information System (INIS)

Takahashi, Yusuke

2016-01-01

An analysis model of plasma flow and electromagnetic waves around a reentry vehicle for radio frequency blackout prediction during aerodynamic heating was developed in this study. The model was validated based on experimental results from the radio attenuation measurement program. The plasma flow properties, such as electron number density, in the shock layer and wake region were obtained using a newly developed unstructured grid solver that incorporated real gas effect models and could treat thermochemically non-equilibrium flow. To predict the electromagnetic waves in plasma, a frequency-dependent finite-difference time-domain method was used. Moreover, the complicated behaviour of electromagnetic waves in the plasma layer during atmospheric reentry was clarified at several altitudes. The prediction performance of the combined model was evaluated with profiles and peak values of the electron number density in the plasma layer. In addition, to validate the models, the signal losses measured during communication with the reentry vehicle were directly compared with the predicted results. Based on the study, it was suggested that the present analysis model accurately predicts the radio frequency blackout and plasma attenuation of electromagnetic waves in plasma in communication. (paper)

Exchange of pressurizer safeguarding system at Biblis nuclear power station

International Nuclear Information System (INIS)

Weber, D.; Hofbeck, W.

1991-01-01

Valves and piping of the pressurizer safeguarding system are exchanged and reset in such a way that they are suitable not only for discharging steam, but also for discharging a water-steam mixture and hot pressurized water; for the emergency measure of primary depressurization by hand (bleed) in the event of failure of the entire feedwater supply and station black-out, and in the event of operational transients with supposed failure of the reactor scram (ATWS). To achieve this, in addition to the requirements of the pressurizer discharging station, changes have to be made to the valve drive to dominate the water loads. During the 1990 inspection this exchange of the pressurizer discharging station was performed at the Biblis A unit as the first German plant. (orig.) [de

Cognitive Blackouts in Mild Cognitive Impairment and Alzheimer’s Dementia

Directory of Open Access Journals (Sweden)

Georg Adler

2018-02-01

Full Text Available Background: Cognitive blackouts, e.g. moments of amnesia, disorientation, or perplexity may be an early sign of incipient Alzheimer’s dementia (AD. A short questionnaire, the checklist for cognitive blackouts (CCB, was evaluated cross-sectionally in users of a memory clinic. Methods: The CCB was performed in 130 subjects, who further underwent a neuropsychological and clinical examination. Subjective memory impairment and depressive symptoms were assessed. Differences in the CCB score between diagnostic groups and relationships with cognitive performance, depression, and subjective memory impairment were analyzed. Results: The CCB score was increased in mild cognitive impairment of the amnestic type or mild AD and correctly predicted 69.2% of the respective subjects. It was negatively correlated with cognitive performance, positively correlated with depressive symptoms, and substantially increased in subjects who estimated their memory poorer than that of other persons of their age. Discussion: The CCB may be a helpful screening tool for the early recognition of AD.

Emergency diesel generator reliability program

International Nuclear Information System (INIS)

Serkiz, A.W.

1989-01-01

The need for an emergency diesel generator (EDG) reliability program has been established by 10 CFR Part 50, Section 50.63, Loss of All Alternating Current Power, which requires that utilities assess their station blackout duration and recovery capability. EDGs are the principal emergency ac power sources for coping with a station blackout. Regulatory Guide 1.155, Station Blackout, identifies a need for (1) an EDG reliability equal to or greater than 0.95, and (2) an EDG reliability program to monitor and maintain the required levels. The resolution of Generic Safety Issue (GSI) B-56 embodies the identification of a suitable EDG reliability program structure, revision of pertinent regulatory guides and Tech Specs, and development of an Inspection Module. Resolution of B-56 is coupled to the resolution of Unresolved Safety Issue (USI) A-44, Station Blackout, which resulted in the station blackout rule, 10 CFR 50.63 and Regulatory Guide 1.155, Station Blackout. This paper discusses the principal elements of an EDG reliability program developed for resolving GSI B-56 and related matters

Europe's electrical vulnerability geography : historical interpretations of the 2006 'European blackout'

NARCIS (Netherlands)

Vleuten, van der E.B.A.; Lagendijk, V.C.

2009-01-01

The so-called "European Blackout" of 4 November 2006 counts as a key example of present day transnational infrastructure vulnerability and an important reference in current debates on transnational electricity infrastructure governance. This is best examplified by the debate itself, where proponents

Blackout sequence modeling for Atucha-I with MARCH3 code

International Nuclear Information System (INIS)

Baron, J.; Bastianelli, B.

1997-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-15

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

Optimization of station battery replacement

International Nuclear Information System (INIS)

Jancauskas, J.R.; Shook, D.A.

1994-01-01

During a loss of ac power at a nuclear generating station (including diesel generators), batteries provide the source of power which is required to operate safety-related components. Because traditional lead-acid batteries have a qualified life of 20 years, the batteries must be replaced a minimum of once during a station's lifetime, twice if license extension is pursued, and more often depending on actual in-service dates and the results of surveillance tests. Replacement of batteries often occurs prior to 20 years as a result of systems changes caused by factors such as Station Blackout Regulations, control system upgrades, incremental load growth, and changes in the operating times of existing equipment. Many of these replacement decisions are based on the predictive capabilities of manual design basis calculations. The inherent conservatism of manual calculations may result in battery replacements occurring before actually required. Computerized analysis of batteries can aid in optimizing the timing of replacements as well as in interpreting service test data. Computerized analysis also provides large benefits in maintaining the as-configured load profile and corresponding design margins, while also providing the capability of quickly analyze proposed modifications and response to internal and external audits

Blackout risk prevention in a smart grid based flexible optimal strategy using Grey Wolf-pattern search algorithms

International Nuclear Information System (INIS)

Mahdad, Belkacem; Srairi, K.

2015-01-01

Highlights: • A generalized optimal security power system planning strategy for blackout risk prevention is proposed. • A Grey Wolf Optimizer dynamically coordinated with Pattern Search algorithm is proposed. • A useful optimized database dynamically generated considering margin loading stability under severe faults. • The robustness and feasibility of the proposed strategy is validated in the standard IEEE 30 Bus system. • The proposed planning strategy will be useful for power system protection coordination and control. - Abstract: Developing a flexible and reliable power system planning strategy under critical situations is of great importance to experts and industrials to minimize the probability of blackouts occurrence. This paper introduces the first stage of this practical strategy by the application of Grey Wolf Optimizer coordinated with pattern search algorithm for solving the security smart grid power system management under critical situations. The main objective of this proposed planning strategy is to prevent the practical power system against blackout due to the apparition of faults in generating units or important transmission lines. At the first stage the system is pushed to its margin stability limit, the critical loads shedding are selected using voltage stability index. In the second stage the generator control variables, the reactive power of shunt and dynamic compensators are adjusted in coordination with minimization the active and reactive power at critical loads to maintain the system at security state to ensure service continuity. The feasibility and efficiency of the proposed strategy is applied to IEEE 30-Bus test system. Results are promising and prove the practical efficiency of the proposed strategy to ensure system security under critical situations

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

International Nuclear Information System (INIS)

Hodge, S.A.

1983-01-01

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

Severe Accident Test Station Activity Report

Energy Technology Data Exchange (ETDEWEB)

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

2015-06-01

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

Review of Leading Approaches for Mitigating Hypersonic Vehicle Communications Blackout and a Method of Ceramic Particulate Injection Via Cathode Spot Arcs for Blackout Mitigation

Science.gov (United States)

Gillman, Eric D.; Foster, John E.; Blankson, Isaiah M.

2010-01-01

Vehicles flying at hypersonic velocities within the atmosphere become enveloped in a "plasma sheath" that prevents radio communication, telemetry, and most importantly, GPS signal reception for navigation. This radio "blackout" period has been a problem since the dawn of the manned space program and was an especially significant hindrance during the days of the Apollo missions. An appropriate mitigation method must allow for spacecraft to ground control and ground control to spacecraft communications through the reentry plasma sheath. Many mitigation techniques have been proposed, including but not limited to, aerodynamic shaping, magnetic windows, and liquid injection. The research performed on these mitigation techniques over the years will be reviewed and summarized, along with the advantages and obstacles that each technique will need to overcome to be practically implemented. A unique approach for mitigating the blackout communications problem is presented herein along with research results associated with this method. The novel method involves the injection of ceramic metal-oxide particulate into a simulated reentry plasma to quench the reentry plasma. Injection of the solid ceramic particulates is achieved by entrainment within induced, energetic cathode spot flows.

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Emergency makeup of nuclear steam generators in blackout conditions

International Nuclear Information System (INIS)

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

2014-01-01

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

Modelling and analysis of severe accidents for VVER-1000 reactors

International Nuclear Information System (INIS)

Tusheva, Polina

2012-01-01

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

Observing power blackouts from space - A disaster related study

Science.gov (United States)

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

2009-04-01

capability of detecting power blackouts in OLS data have been identified (e.g. sunlight, heavy cloud cover and bright moonlight). Furthermore, the change detection procedure only works when power blackouts happen or still persist at night at the time of an OLS overpass. In some cases (e.g. Hurricane Katrina) it has been possible to track the gradual recovery of power by repeating the procedure on nights following a disaster event. In this paper several examples of successful power blackout detection following natural disasters including hurricanes (e.g. Isabel 2003 and Wilma 2005 in the USA) and earthquakes (e.g. Gujarat Earthquake 2001 in India) will be presented, whereas overlaid hurricane paths and earthquake epicenters serve as landmarks and indicate locations around the potential highest impact. Disaster impact assessment and post-disaster research is strongly related to impacts on population, related infrastructure and activities (Kerle et al. 2005, Zhang and Kerle 2008). In particular in the case of emergency management and response humans are the main actors and first-pass assessment of affected population and locations of affected areas are essential. Space-based power blackout detection, as described above, has the potential to delineate the spatial extent of the disaster impact. Overlaying the respective OLS data with regional population data such as LandScan (Dobson et al. 2000) or Gridded Population of the World (CIESIN and CIAT 2005) allows estimating a potential number of affected people. Without a doubt such estimates comprise a considerable number of uncertainties. However, the capability of providing the information in near-real time as offered by using DMSP-OLS makes the presented approach very valuable for emergency and disaster managers worldwide. REFERENCES Center for International Earth Science Information Network CIESIN at Columbia University, and Centro Internacional de Agricultura Tropical CIAT (2005). Gridded Population of the World Version 3 (GPWv

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-01

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

Recent SCDAP/RELAP5 improvements for BWR severe accident simulations

International Nuclear Information System (INIS)

Griffin, F.P.

1995-01-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Performance Evaluation of Target Detection with a Near-Space Vehicle-Borne Radar in Blackout Condition.

Science.gov (United States)

Li, Yanpeng; Li, Xiang; Wang, Hongqiang; Deng, Bin; Qin, Yuliang

2016-01-06

Radar is a very important sensor in surveillance applications. Near-space vehicle-borne radar (NSVBR) is a novel installation of a radar system, which offers many benefits, like being highly suited to the remote sensing of extremely large areas, having a rapidly deployable capability and having low vulnerability to electronic countermeasures. Unfortunately, a target detection challenge arises because of complicated scenarios, such as nuclear blackout, rain attenuation, etc. In these cases, extra care is needed to evaluate the detection performance in blackout situations, since this a classical problem along with the application of an NSVBR. However, the existing evaluation measures are the probability of detection and the receiver operating curve (ROC), which cannot offer detailed information in such a complicated application. This work focuses on such requirements. We first investigate the effect of blackout on an electromagnetic wave. Performance evaluation indexes are then built: three evaluation indexes on the detection capability and two evaluation indexes on the robustness of the detection process. Simulation results show that the proposed measure will offer information on the detailed performance of detection. These measures are therefore very useful in detecting the target of interest in a remote sensing system and are helpful for both the NSVBR designers and users.

Lightning rod ionizing natural ionca - Ionic electrode active trimetallictriac of grounding - Definitive and total solution against 'blackouts' and electrical faults generated by atmospheric charges (lightning)

Energy Technology Data Exchange (ETDEWEB)

Cabareda, Luis

2010-09-15

The Natural Ionizing System of Electrical Protection conformed by: Lightning Rod Ionizing Natural Ionca and Ionic Electrode Active Trimetallic Triac of Grounding offers Total Protection, Maximum Security and Zero Risk to Clinics, Hospitals, Integral Diagnostic Center, avoiding ''the burning'' of Electronics Cards; Refineries, Tanks and Stations of Fuel Provision; Electrical Substations, Towers and Transmission Lines with transformer protection, motors, elevators, A/C, mechanicals stairs, portable and cooling equipment, electrical plants, others. This New High Technology is the solution to the paradigm of Benjamin Franklin and it's the mechanism to end the 'Blackouts' that produces so many damages and losses throughout the world.

High-Tech, Low-Tech, No-Tech: Communications Strategies During Blackouts

Science.gov (United States)

2013-12-01

straight-line “ derecho ”1 windstorms hit the mid-Atlantic region of the United States. In the National Capital Region (NCR), many residents lost...flooding (Grand Forks, North Dakota, 1997), rolling blackouts (California, 2001), multi-state power outage (Ohio and seven other 1 “ Derecho ” is a...flawlessly throughout the 2012 “super- derecho ”20 storm event.21 During its 2012 typhoon, the government of the Philippines used Twitter to

Can the complex networks help us in the resolution of the problem of power outages (blackouts) in Brazil?

Energy Technology Data Exchange (ETDEWEB)

Castro, Paulo Alexandre de; Souza, Thaianne Lopes de [Universidade Federal de Goias (UFG), Catalao, GO (Brazil)

2011-07-01

Full text. What the Brazilian soccer championship, Hollywood actors, the network of the Internet, the spread of viruses and electric distribution network have in common? Until less than two decade ago, the answer would be 'nothing' or 'almost nothing'. However, the answer today to this same question is 'all' or 'almost all'. The answer to these questions and more can be found through a sub-area of statistical physics | called science of complex networks that has been used to approach and study the most diverse natural and non-natural systems, such as systems/social networks, information, technological or biological. In this work we study the distribution network of electric power in Brazil (DEEB), from a perspective of complex networks, where we associate stations and/or substations with a network of vertices and the links between the vertices we associate with the transmission lines. We are doing too a comparative study with the best-known models of complex networks, such as Erdoes-Renyi, Configuration Model and Barabasi-Albert, and then we compare with results obtained in real electrical distribution networks. Based on this information, we do a comparative analysis using the following variables: connectivity distribution, diameter, clustering coefficient, which are frequently used in studies of complex networks. We emphasize that the main objective of this study is to analyze the robustness of the network DEEB, and then propose alternatives for network connectivity, which may contribute to the increase of robustness in maintenance projects and/or expansion of the network, in other words our goal is to make the network to proof the blackouts or improve the endurance the network against the blackouts. For this purpose, we use information from the structural properties of networks, computer modeling and simulation. (author)

Conditions for oxygen-deficient combustion during accidents with severe core concrete thermal attack

International Nuclear Information System (INIS)

Luangdilok, W.; Elicson, G.T.; Berger, W.E. Jr.

1993-01-01

This paper addresses the interactions between MCCI (molten core-concrete interactions)-induced offgas releases, mostly the combustible gases, natural circulation between the cavity and the lower containment based on recent research developments in the area of mixed convection flow (Epstein, et al., 1989; Epstein, 1988; Epstein, 1992) between compartments, and their effects on combustion in PWR containments during prolonged severe accidents. Specifically, large dry PWR containments undergoing severe core-concrete attack during station blackouts where the containment atmosphere is expected to be inerted are objects of this analysis. The purpose of this paper, given the conditions that oxygen can be brought to the cavity, is to demonstrate that consumption of most oxygen present in the containment can be achieved in a reasonable time scale assuming that combustion is not subject to flammability limits due to the high cavity temperatures. The conditions for cavity combustion depend on several factors including good gas flowpaths between the cavity and other containment regions, and combustion processes within the cavity with the hot debris acting as the ignition source

Thermal Hydraulic design parameters study for severe accidents using neural networks

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

Thermal Hydraulic design parameters study for severe accidents using neural networks

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

MELCOR assessment of sequential severe accident mitigation actions under SGTR accident

International Nuclear Information System (INIS)

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

2017-01-01

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

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

International Nuclear Information System (INIS)

Leteurtrois, J.P.

2007-01-01

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

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

International Nuclear Information System (INIS)

Park, Rae-Joon; Hong, Seong-Wan

2011-01-01

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

AP1000{sup R} severe accident features and post-Fukushima considerations

Energy Technology Data Exchange (ETDEWEB)

Scobel, J. H.; Schulz, T. L.; Williams, M. G. [Westinghouse Electric Company, LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

2012-07-01

The AP1000{sup R} passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at Fukushima without compromising core and containment integrity. The AP1000 plant shuts down the reactor, cools the core, containment and spent fuel pool for more than 3 days using passive systems that do not require AC or DC power or operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. To provide defense-in-depth for design extension conditions, the AP1000 plant has engineered features that mitigate the effects of core damage. Engineered features retain damaged core debris within the reactor vessel as a key feature. Other aspects of the design protect containment integrity during severe accidents, including unique features of the AP1000 design relative to passive containment cooling with water and air, and hydrogen management. (authors)

Sinister synergies : how competition for unregulated profit causes blackouts

International Nuclear Information System (INIS)

Wilson, J.

2005-01-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

77 FR 16175 - Station Blackout

Science.gov (United States)

2012-03-20

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

Silencing Boko Haram: Mobile Phone Blackout and Counterinsurgency in Nigeria’s Northeast region

Directory of Open Access Journals (Sweden)

Jacob Udo-Udo Jacob

2015-03-01

Full Text Available In the summer of 2013, the Nigerian military, as part of its counterinsurgency operations against Boko Haram insurgents, shut down GSM mobile telephony in three northeast states – Adamawa, Borno and Yobe. This article explores the rationale, impact and citizens’ opinion of the mobile phone blackout. It draws on focus group discussions with local opinion leaders and in-depth personal interviews with military and security insiders, as well as data of Boko Haram incidences before, during and after the blackout from military sources and conflict databases. It argues that, although the mobile phone shutdown was ‘successful’ from a military- tactical point of view, it angered citizens and engendered negative opinions toward the state and new emergency policies. While citizens developed various coping and circumventing strategies, Boko Haram evolved from an open network model of insurgency to a closed centralized system, shifting the center of its operations to the Sambisa Forest. This fundamentally changed the dynamics of the conflict. The shutdown demonstrated, among others, that while ICTs serve various desirable purposes for developing states, they will be jettisoned when their use challenges the state’s legitimacy and raison d'être, but not without consequences.

Interpreting transnational infrastructure vulnerability: European blackout and the historical dynamics of transnational electricity governance

International Nuclear Information System (INIS)

Vleuten, Erik van der; Lagendijk, Vincent

2010-01-01

Recent transnational blackouts exposed two radically opposed interpretations of Europe's electricity infrastructure, which inform recent and ongoing negotiations on transnational electricity governance. To EU policy makers such blackouts revealed the fragility of Europe's power grids and the need of a more centralized form of governance, thus legitimizing recent EU interventions. Yet to power sector spokespersons, these events confirmed the reliability of transnational power grids and the traditional decentralized governance model: the disturbances were quickly contained and repaired. This paper inquires the historic legacies at work in these conflicting interpretations and associated transnational governance preferences. It traces the power sector's interpretation to its building of a secure transnational power grid from the 1950s through the era of neoliberalization. Next it places the EU interpretation and associated policy measures against the historical record of EU attempts at transnational infrastructure governance. Uncovering the historical roots and embedding of both interpretations, we conclude that their divergence is of a surprisingly recent date and relates to the current era of security thinking. Finally we recommend transnational, interpretative, and historical analysis to the field of critical infrastructure studies.

An adaptive reentry guidance method considering the influence of blackout zone

Science.gov (United States)

Wu, Yu; Yao, Jianyao; Qu, Xiangju

2018-01-01

Reentry guidance has been researched as a popular topic because it is critical for a successful flight. In view that the existing guidance methods do not take into account the accumulated navigation error of Inertial Navigation System (INS) in the blackout zone, in this paper, an adaptive reentry guidance method is proposed to obtain the optimal reentry trajectory quickly with the target of minimum aerodynamic heating rate. The terminal error in position and attitude can be also reduced with the proposed method. In this method, the whole reentry guidance task is divided into two phases, i.e., the trajectory updating phase and the trajectory planning phase. In the first phase, the idea of model predictive control (MPC) is used, and the receding optimization procedure ensures the optimal trajectory in the next few seconds. In the trajectory planning phase, after the vehicle has flown out of the blackout zone, the optimal reentry trajectory is obtained by online planning to adapt to the navigation information. An effective swarm intelligence algorithm, i.e. pigeon inspired optimization (PIO) algorithm, is applied to obtain the optimal reentry trajectory in both of the two phases. Compared to the trajectory updating method, the proposed method can reduce the terminal error by about 30% considering both the position and attitude, especially, the terminal error of height has almost been eliminated. Besides, the PIO algorithm performs better than the particle swarm optimization (PSO) algorithm both in the trajectory updating phase and the trajectory planning phases.

Characterization of weakly ionized argon flows for radio blackout mitigation experiments

Science.gov (United States)

Steffens, L.; Koch, U.; Esser, B.; Gülhan, A.

2017-06-01

For reproducing the so-called E × B communication blackout mitigation scheme inside the L2K arc heated facility of the DLR in weakly ionized argon §ows, a §at plate model has been equipped with a superconducting magnet, electrodes, and a setup comprising microwave plasma transmission spectroscopy (MPTS). A thorough characterization of the weakly ionized argon §ow has been performed including the use of microwave interferometry (MWI), Langmuir probe measurements, Pitot probe pro¦les, and spectroscopic methods like diode laser absorption spectroscopy (DLAS) and emission spectroscopy.

Air ingression calculations for selected plant transients using MELCOR

International Nuclear Information System (INIS)

Kmetyk, L.N.

1994-01-01

Two sets of MELCOR calculations have been completed studying the effects of air ingression on the consequences of various severe accident scenarios. One set of calculations analyzed a station blackout with surge line failure prior to vessel breach, starting from nominal operating conditions; the other set of calculations analyzed a station blackout occurring during shutdown (refueling) conditions. Both sets of analyses were for the Surry plant, a three-loop Westinghouse PWR. For both accident scenarios, a basecase calculation was done, and then repeated with air ingression from containment into the core region following core degradation and vessel failure. In addition to the two sets of analyses done for this program, a similar air-ingression sensitivity study was done as part of a low-power/shutdown PRA, with results summarized here; that PRA study also analyzed a station blackout occurring during shutdown (refueling) conditions, but for the Grand Gulf plant, a BWR/6 with Mark III containment. These studies help quantify the amount of air that would have to enter the core region to have a significant impact on the severe accident scenario, and demonstrate that one effect, of air ingression is substantial enhancement of ruthenium release. These calculations also show that, while the core clad temperatures rise more quickly due to oxidation with air rather than steam, the core also degrades and relocates more quickly, so that no sustained, enhanced core heatup is predicted to occur with air ingression

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-08-15

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

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

Directory of Open Access Journals (Sweden)

Afshin Hedayat

2017-08-01

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

Development of filtered containment venting system and application for Kashiwazaki-Kariwa Nuclear Power Station Unit 6, 7

International Nuclear Information System (INIS)

Murai, Soutarou; Hiranuma, Naoki; Kimura, Takeo; Omori, Shuichi; Watanabe, Fumitoshi; Sasa, Daisuke

2014-01-01

The Fukushima Dai-ichi Nuclear Power Station (1F) of Tokyo Electric Power Company (TEPCO) had experienced severe radio-active release to the environment in the Tohoku Region Pacific Coast Earthquake (alias: the Great East Japan Earthquake) in 2011. Under the Station Black-Out (SBO) conditions caused by tsunami with the earthquake, the 1F operators had tried to vent the gasses in the Primary Containment Vessels (PCVs) of the unit 1, 2 and 3 to the environment through the water pools in the suppression chambers of the PCVs. Its venting, however, was imperfect and, as a result, major direct radio-active release to the environment was caused. After this disaster, TEPCO launched a project to develop the Filtered Containment Venting System (FCVS), in which our very bitter experiences in the 1F accident as described above are reflected. One of the main purposes of the development of the FCVS is to enhance operability of venting under the severe plant conditions such as the SBO during progressing of severe core damage, and another is to enhance removal performance of radio-nuclides with the newly added filtering equipment, which is installed in the venting line from the PCV to the outer. The Kashiwazaki-Kariwa NPS unit 6 and 7 will be the first reactors applied the FCVSs. In this paper, we show the design concept of the TEPCO's FCVS, the brief overview of the system design and the summary of experiment which has been performed for getting the performance data of the FCVS such as decontamination factor in various conditions. (author)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

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

Acute Alcohol Effects on Narrative Recall and Contextual Memory: An Examination of Fragmentary Blackouts

OpenAIRE

Wetherill, Reagan R.; Fromme, Kim

2011-01-01

The present study examined the effects of alcohol consumption on narrative recall and contextual memory among individuals with and without a history of fragmentary blackouts in an attempt to better understand why some individuals experience alcohol-induced memory impairments whereas others do not, even at comparable blood alcohol concentrations (BACs). Standardized beverage (alcohol, no alcohol) administration procedures and neuropsychological assessments measured narrative recall and context...

Binge Drinking and the Young Brain: A Mini Review of the Neurobiological Underpinnings of Alcohol-Induced Blackout

Directory of Open Access Journals (Sweden)

Daniel F. Hermens

2018-01-01

Full Text Available Binge drinking has significant effects on memory, particularly with regards to the transfer of information to long-term storage. Partial or complete blocking of memory formation is known as blackout. Youth represents a critical period in brain development that is particularly vulnerable to alcohol misuse. Animal models show that the adolescent brain is more vulnerable to the acute and chronic effects of alcohol compared with the adult brain. This mini-review addresses the neurobiological underpinnings of binge drinking and associated memory loss (blackout in the adolescent and young adult period. Although the extent to which there are pre-existing versus alcohol-induced neurobiological changes remains unclear, it is likely that repetitive binge drinking in youth has detrimental effects on cognitive and social functioning. Given its role in learning and memory, the hippocampus is a critical region with neuroimaging research showing notable changes in this structure associated with alcohol misuse in young people. There is a great need for earlier identification of biological markers associated with alcohol-related brain damage. As a means to assess in vivo neurochemistry, magnetic resonance spectroscopy (MRS has emerged as a particularly promising technique since changes in neurometabolites often precede gross structural changes. Thus, the current paper addresses how MRS biomarkers of neurotransmission (glutamate, GABA and oxidative stress (indexed by depleted glutathione in the hippocampal region of young binge drinkers may underlie propensity for blackouts and other memory impairments. MRS biomarkers may have particular utility in determining the acute versus longer-term effects of binge drinking in young people.

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-03-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-08-01

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

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

International Nuclear Information System (INIS)

Hyman, C.R.

1988-01-01

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

Hot Cell Installation and Demonstration of the Severe Accident Test Station

Energy Technology Data Exchange (ETDEWEB)

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

2017-08-01

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

Analysis of Radio Frequency Blackout for a Blunt-Body Capsule in Atmospheric Reentry Missions

Directory of Open Access Journals (Sweden)

Yusuke Takahashi

2016-01-01

Full Text Available A numerical analysis of electromagnetic waves around the atmospheric reentry demonstrator (ARD of the European Space Agency (ESA in an atmospheric reentry mission was conducted. During the ARD mission, which involves a 70% scaled-down configuration capsule of the Apollo command module, radio frequency blackout and strong plasma attenuation of radio waves in communications with data relay satellites and air planes were observed. The electromagnetic interference was caused by highly dense plasma derived from a strong shock wave generated in front of the capsule because of orbital speed during reentry. In this study, the physical properties of the plasma flow in the shock layer and wake region of the ESA ARD were obtained using a computational fluid dynamics technique. Then, electromagnetic waves were expressed using a frequency-dependent finite-difference time-domain method using the plasma properties. The analysis model was validated based on experimental flight data. A comparison of the measured and predicted results showed good agreement. The distribution of charged particles around the ESA ARD and the complicated behavior of electromagnetic waves, with attenuation and reflection, are clarified in detail. It is suggested that the analysis model could be an effective tool for investigating radio frequency blackout and plasma attenuation in radio wave communication.

Selection, design, qualification, testing, and reliability of emergency diesel generator units used as Class 1E onsite electric power systems at nuclear power plants

International Nuclear Information System (INIS)

1992-04-01

This guide has been prepared for the resolution of Generic Safety Issue B-56, ''Diesel Generator Reliability,'' and is related to Unresolved Safety Issue (USI) A-44, ''Station Blackout.'' The resolution of USI A-44 established a need for an emergency diesel generator (EDG) reliability program that has the capability to achieve and maintain the emergency diesel generator reliability levels in the range of 0.95 per demand or better to cope with station blackout

Evaluating the Effect of a Campus-Wide Social Norms Marketing Intervention on Alcohol-Use Perceptions, Consumption, and Blackouts

Science.gov (United States)

Su, Jinni; Hancock, Linda; Wattenmaker McGann, Amanda; Alshagra, Mariam; Ericson, Rhianna; Niazi, Zackaria; Dick, Danielle M.; Adkins, Amy

2018-01-01

Objective: To evaluate the effect of a campus-wide social norms marketing intervention on alcohol-use perceptions, consumption, and blackouts at a large, urban, public university. Participants: 4,172 college students (1,208 freshmen, 1,159 sophomores, 953 juniors, and 852 seniors) who completed surveys in Spring 2015 for the Spit for Science…

Correlation of Steam Generator Mixing Parameters for Severe Accident Hot-Leg Natural Circulation

Energy Technology Data Exchange (ETDEWEB)

Liao, Yehong; Guentay, Salih [Paul Scherrer Institut, Villigen PSI, CH-5232 (Switzerland)

2008-07-01

Steam generator inlet plenum mixing phenomenon with hot-leg counter-current natural circulation during a PWR station blackout severe accident is one of the important processes governing which component will fail first as a result of thermal challenge from the circulating gas with high temperature and pressure. Since steam generator tube failure represents bypass release of fission product from the reactor to environment, study of inlet plenum mixing parameters is important to risk analysis. Probability distribution functions of individual mixing parameter should be obtained from experiments or calculated by analysis. In order to perform sensitivity studies of the synergetic effects of all mixing parameters on the severe accident-induced steam generator tube failure, the distribution and correlation of these mixing parameters must be known to remove undue conservatism in thermal-hydraulic calculations. This paper discusses physical laws governing three mixing parameters in a steady state and setups the correlation among these mixing parameters. The correlation is then applied to obtain the distribution of one of the mixing parameters that has not been given in the previous CFD analysis. Using the distributions and considering the inter-dependence of the three mixing parameters, three sensitivity cases enveloping the mixing parameter uncertainties are recommended for the plant analysis. (authors)

Severe damage analysis of VVER 1000 following large break LOCA using Astec code

International Nuclear Information System (INIS)

Chatterjee, B.; Mukhopadhyay, D.; Lele, H.G.; Ghosh, A.K.; Kushwaha, H.S.

2007-01-01

Severe accident analysis of a reactor is an important aspect in the evaluation of source term. This in turn helps in emergency planning. An analysis has been carried out for VVER-1000 (V320) reactor following Large Break LOCA (loss of coolant accident) along with Station Blackout (SBO). Computer code ASTEC (jointly developed by IRSN, France, and GRS, Germany) is used for analyzing the transient. This integral code has been designed to be used as reference code for PSA2 studies. Severe accident analysis is carried out for an accident initiated by Large break LOCA along with SBO. Two cases have been analysed with the version ASTEC V1.2-rev1. In the first case hydro-accumulators are considered not available while the second case has been analysed with hydro accumulators. In this paper, ASTEC predictions have been studied for the in-vessel phase of the accident till vessel failure. The vessel failure was observed at 6979 s when accumulators were assumed not available. The vessel failure was quite delayed (19294 s) with operating accumulators. The hydrogen production was found to be very large (22% of total Zr inventory) in the case with accumulators compared to the case without accumulators (1.5% of total Zr inventory)

Severe Accident Test Station Design Document

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-01

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

Severe Accident Test Station Design Document

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-15

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

Real-time stability in power systems techniques for early detection of the risk of blackout

CERN Document Server

Savulescu, Savu

2014-01-01

This pioneering volume has been updated and enriched to reflect the state-of-the-art in blackout prediction and prevention. It documents and explains background and algorithmic aspects of the most successful steady-state, transient and voltage stability solutions available today in real-time. It also describes new, cutting-edge stability applications of synchrophasor technology, and captures industry acceptance of metrics and visualization tools that quantify and monitor the distance to instability. Expert contributors review a broad spectrum of additionally available techniques, such as traje

Finite element analysis of an underground protective test station subjected to severe ground motion

International Nuclear Information System (INIS)

Burchett, S.N.; Milloy, J.A.; Von Riesemann, W.A.

1974-01-01

A recoverable test station, to be used at a location very close to the detonation point of an underground nuclear test, was designed and tested. The design required that the station survive the very severe free-field stress and that the acceleration of the station be limited to a tolerable level. The predicted magnitude and time history of the free-field stress and acceleration indicated that the volcanic tuff medium and the materials of the proposed structure would exhibit nonlinear behavior, so that a transient dynamic analysis of a composite type structure involving nonlinear materials and large deformations was necessary. Parameter studies using a linear elastic dynamic finite element technique were first done to gain an understanding of the effect of the material properties and to study the response of various configurations of the protective structure to the transient loading. Based upon the results of these parameter studies and upon fielding considerations, a tentative design was chosen. This design was then evaluated using a newly developed finite element computer program. The results obtained from this analysis are compared to the field test results. (U.S.)

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

International Nuclear Information System (INIS)

Sadek, Sinisa; Amizic, Milan; Grgic, Davor

2013-01-01

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

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

Directory of Open Access Journals (Sweden)

SUNGMIN KIM

2013-08-01

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

Investigation on accident management measures for VVER-1000 reactors

International Nuclear Information System (INIS)

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

2009-01-01

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

Resolution of GSI B-56 - Emergency diesel generator reliability

International Nuclear Information System (INIS)

Serkiz, A.W.

1989-01-01

The need for an emergency diesel generator (EDG) reliability program has been established by 10 CFR Part 50, Section 50.63, Loss of All Alternating Current Power, which requires that licensees assess their station blackout coping and recovery capability. EDGs are the principal emergency ac power sources for avoiding a station blackout. Regulatory Guide 1.155, Station Blackout, identifies a need for (1) a nuclear unit EDG reliability level of at least 0.95, and (2) an EDG reliability program to monitor and maintain the required EDG reliability levels. NUMARC-8700, Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors, also provides guidance on such needs. The resolution of GSI B-56, Diesel Reliability will be accomplished by issuing Regulatory Guide 1.9, Rev. 3, Selection, Design, Qualification, Testing, and Reliability of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Plants. This revision will integrate into a single regulatory guide pertinent guidance previously addressed in R.G. 1.9, Rev. 2, R.G. 1.108, and Generic Letter 84-15. R.G. 1.9 has been expanded to define the principal elements of an EDG reliability program for monitoring and maintaining EDG reliability levels selected for SBO. In addition, alert levels and corrective actions have been defined to detect a deteriorating situation for all EDGs assigned to a particular nuclear unit, as well as an individual problem EDG

Overview of the Nuclear Regulatory Commission's safety research program

International Nuclear Information System (INIS)

Beckjord, E.S.

1989-01-01

Accomplishments during 1988 of the Office of Nuclear Regulatory Research and the program of safety research are highlighted, and plans, expections, and needs of the next year and beyond are discussed. Topics discussed include: ECCS Appendix K Revision; pressurized thermal shock; NUREG-1150, or the PRA method performance document; resolution of station blackout; severe accident integration plan; nuclear safety research review committee; and program management

MANAGING A PROLONGED STATION BLACKOUT CONDITION IN AHWR BY PASSIVE MEANS

Directory of Open Access Journals (Sweden)

MUKESH KUMAR

2013-10-01

In view of this, an analysis has been performed for decay heat removal characteristics over several days of an AHWR by ICs. The computer code RELAP5/MOD3.2 was used for this purpose. Results indicate that the ICs can remove the decay heat for more than 10 days without causing any bulk boiling in the GDWP. After that, decay heat can be removed for more than 40 days by boiling off the pool inventory. The pressure inside the containment does not exceed the design pressure even after 10 days by condensation of steam generated from the GDWP on the walls of containment and on the Passive Containment Cooling System (PCCS tubes. If venting is carried out after this period, the decay heat can be removed for more than 50 days without exceeding the design limits.

Steam Oxidation Testing in the Severe Accident Test Station

Energy Technology Data Exchange (ETDEWEB)

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

2017-08-01

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

Utilities must leverage existing resources and upgrade technology to avoid future blackouts

International Nuclear Information System (INIS)

Masiello, R.

2004-01-01

The blackout of August 14, 2003 is used as the incentive to examine transmission grid reliability, expose its deficiencies in terms of technology, standards and processes used to manage reliability, and to make recommendations for leveraging existing resources and upgrading both the technology and procedures to avoid similar breakdowns in the future. It is recommended that in the area of monitoring transmission grid reliability utilities should borrow a page from the Enterprise Risk Management Practices of the financial industry by adopting a system which looks beyond the first 'credible' contingency (the current system) and examine many more 'incredible' contingencies, and underlying events that can trigger multiple contingencies, and plan for them in their operations. Utility companies are also urged to upgrade their energy management systems (EMS) technology to be able to deal with the kinds of severely depressed voltages and overloaded circuits that many grids experience today. Investment in new capabilities in control rooms, more and better communications will be essential. EMS algorithms and models must be upgraded to operate under a broader spectrum of grid conditions and to simulate the once-in-a-lifetime outage scenarios that most operators believe could never strike their utility. Expert opinion strongly suggests that as part of this process of upgrading, contingency analysis should shift from the 'N-1' model of the present to a stochastic model that considers a wider range of possible events in a probabilistic framework

Utilities must leverage existing resources and upgrade technology to avoid future blackouts

Energy Technology Data Exchange (ETDEWEB)

Masiello, R. [KEMA Inc. Burlington, MA (United States)

2004-06-01

The blackout of August 14, 2003 is used as the incentive to examine transmission grid reliability, expose its deficiencies in terms of technology, standards and processes used to manage reliability, and to make recommendations for leveraging existing resources and upgrading both the technology and procedures to avoid similar breakdowns in the future. It is recommended that in the area of monitoring transmission grid reliability utilities should borrow a page from the Enterprise Risk Management Practices of the financial industry by adopting a system which looks beyond the first 'credible' contingency (the current system) and examine many more 'incredible' contingencies, and underlying events that can trigger multiple contingencies, and plan for them in their operations. Utility companies are also urged to upgrade their energy management systems (EMS) technology to be able to deal with the kinds of severely depressed voltages and overloaded circuits that many grids experience today. Investment in new capabilities in control rooms, more and better communications will be essential. EMS algorithms and models must be upgraded to operate under a broader spectrum of grid conditions and to simulate the once-in-a-lifetime outage scenarios that most operators believe could never strike their utility. Expert opinion strongly suggests that as part of this process of upgrading, contingency analysis should shift from the 'N-1' model of the present to a stochastic model that considers a wider range of possible events in a probabilistic framework.

Resonant interaction of electromagnetic wave with plasma layer and overcoming the radiocommunication blackout problem

Science.gov (United States)

Bogatskaya, A. V.; Klenov, N. V.; Tereshonok, M. V.; Adjemov, S. S.; Popov, A. M.

2018-05-01

We present an analysis of the possibility of penetrating electromagnetic waves through opaque media using an optical-mechanical analogy. As an example, we consider the plasma sheath surrounding the vehicle as a potential barrier and analyze the overcoming of radiocommunication blackout problem. The idea is to embed a «resonator» between the surface on the vehicle and plasma sheath which is supposed to provide an effective tunneling of the signal to the receiving antenna. We discuss the peculiarities of optical mechanical analogy applicability and analyze the radio frequency wave tunneling regime in detail. The cases of normal and oblique incidence of radiofrequency waves on the vehicle surface are studied.

The Threshold of the State: Civil Defence, the Blackout and the Home in Second World War Britain.

Science.gov (United States)

Greenhalgh, James

2017-06-01

This article reconsiders the way that the British state extended its control of the home during the Second World War, using the implementation of air raid precautions and the blackout as a lens through which to view the state's developing attitudes to domestic space. Presented here is not the familiar story of pitch-dark, dangerous streets or altered cityscapes of fear and destruction; instead, by examining personal testimony the article inverts traditional treatments of the blackout to look at the interior of dwellings, demonstrating how the realities of total warfare impinged upon the psychological elements that constituted the home. What emerges not only expands historical understandings of the wartime experience of civilians, it also shows civil defence measures as highly visible points on an often antagonistic trajectory of state interactions with citizens concerning the privacy and security of the dwelling in the modern city. The requirements of civil defence, I argue, were not merely the product of exceptional wartime circumstances, but symptomatic of long-standing attempts to open up dwellings to state scrutiny. These attempts had both a significant pre-war lineage and, crucially, implications beyond the end of the war in private homes and on social housing estates. © The Author [2017]. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

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

Insights from Severe Accident Analyses for Verification of VVER SAMG

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Nuclear energy + solar energy, why not?

International Nuclear Information System (INIS)

Hernandez C, I.; Nelson E, P.

2016-09-01

Clean energies such as nuclear and solar are part of the solution to the energy dependence that we face today and also help us to reduce the greenhouse gas emissions, thus avoiding a global average temperature increase that is irreversible and harmful to all living beings on the planet. Independently the nuclear and solar energies have had a great development in recent years, so in this work we set ourselves the task of creating a synergy between them. First, we conducted a survey of different people involved in the area of energy (energy efficiency, clean energy and renewable sources) in order to know if the area of which they are part influences with respect to the impression that they have of safety in terms of supply, return on investment and safety to the health and environment of another energy source for which we use a correlation analysis. With the results obtained we propose to use photo thermic solar energy as a support to reduce the frequency of accidents by station blackout and we perform the analysis of the combination using the methodology of Probabilistic Analysis of Security with the help of SAPHIRE 7 software to realize the event trees by station blackout of a nuclear power plant and faults for a photo-thermal solar plant. Finally, the decrease in the probability of station blackout from the proposed combination is quantified. The results were favorable to indicate that the probability of station blackout is reduced in half and that is why is suggested to continue studying the combination. (Author)

Nuclear energy + solar energy, why not?; Energia nuclear + energia solar, por que no?

Energy Technology Data Exchange (ETDEWEB)

Hernandez C, I.; Nelson E, P., E-mail: ihernandezc91@hotmail.com [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

2016-09-15

Clean energies such as nuclear and solar are part of the solution to the energy dependence that we face today and also help us to reduce the greenhouse gas emissions, thus avoiding a global average temperature increase that is irreversible and harmful to all living beings on the planet. Independently the nuclear and solar energies have had a great development in recent years, so in this work we set ourselves the task of creating a synergy between them. First, we conducted a survey of different people involved in the area of energy (energy efficiency, clean energy and renewable sources) in order to know if the area of which they are part influences with respect to the impression that they have of safety in terms of supply, return on investment and safety to the health and environment of another energy source for which we use a correlation analysis. With the results obtained we propose to use photo thermic solar energy as a support to reduce the frequency of accidents by station blackout and we perform the analysis of the combination using the methodology of Probabilistic Analysis of Security with the help of SAPHIRE 7 software to realize the event trees by station blackout of a nuclear power plant and faults for a photo-thermal solar plant. Finally, the decrease in the probability of station blackout from the proposed combination is quantified. The results were favorable to indicate that the probability of station blackout is reduced in half and that is why is suggested to continue studying the combination. (Author)

Severe weather data near nuclear power station and reprocessing fuel facility in Japan

International Nuclear Information System (INIS)

Nagata, Tadahisa

2017-01-01

The main weather data are updated at any time. The strong wind and tornado (strong wind/tornado) data are opened until March 2016 in Japan. The main weather and the strong wind/tornado data near the nuclear power station (NPS) were investigated. The earthquake, Tunami and volcano were not mentioned on this report. The main weather data might not be severe. The maximum temperature had not been considered in the safety analysis of NPS. The weather data of some small observation posts near NPSs had not been considered. The unusual high temperature and the local severe rain near NPS by the global warming may be considered in future. The maximum intensities of the strong wind/tornado in Japan and near NPS were Fujita-scale 3 and 2, respectively. The maximum intensities of almost half NPSs were Fujita-scale 1. The intensity and the number of the strong winds/tornados differed depending on NPS. The Japanese main weather and strong wind/tornado might not be severe compared with other country. (author)

MELRPI - development and use

International Nuclear Information System (INIS)

Sozer, A.

1985-01-01

The MELRPI computer code has been developed by Rensselaer Polytechnic Institute under the sponsorship of Oak Ridge National Laboratory (ORNL) and, more recently, the Empire State Electrical Energy Research Corporation (ESEERCO). The code was developed especially for severe accident analyses concerning BWRs and is not applicable to PWRs. MELRPI.MOD2, one of the ORNL-severe accident analysis codes, has been applied for the first time to station blackout transient analysis for the Browns Ferry nuclear power plant in order to estimate the progression of core degradation

Severity Classification of a Seismic Event based on the Magnitude-Distance Ratio Using Only One Seismological Station

Directory of Open Access Journals (Sweden)

Luis Hernán Ochoa Gutiérrez

2014-07-01

Full Text Available Seismic event characterization is often accomplished using algorithms based only on information received at seismological stations located closest to the particular event, while ignoring historical data received at those stations. These historical data are stored and unseen at this stage. This characterization process can delay the emergency response, costing valuable time in the mitigation of the adverse effects on the affected population. Seismological stations have recorded data during many events that have been characterized by classical methods, and these data can be used as previous "knowledge" to train such stations to recognize patterns. This knowledge can be used to make faster characterizations using only one three-component broadband station by applying bio-inspired algorithms or recently developed stochastic methods, such as kernel methods. We trained a Support Vector Machine (SVM algorithm with seismograph data recorded by INGEOMINAS's National Seismological Network at a three-component station located near Bogota, Colombia. As input model descriptors, we used the following: (1 the integral of the Fourier transform/power spectrum for each component, divided into 7 windows of 2 seconds and beginning at the P onset time, and (2 the ratio between the calculated logarithm of magnitude (Mb and epicentral distance. We used 986 events with magnitudes greater than 3 recorded from late 2003 to 2008. The algorithm classifies events with magnitude-distance ratios (a measure of the severity of possible damage caused by an earthquake greater than a background value. This value can be used to estimate the magnitude based on a known epicentral distance, which is calculated from the difference between P and S onset times. This rapid (< 20 seconds magnitude estimate can be used for rapid response strategies. The results obtained in this work confirm that many hypocentral parameters and a rapid location of a seismic event can be obtained using a few

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-15

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

MAAP4 CANDU analysis of a generic CANDU-6 plant: preliminary results

Energy Technology Data Exchange (ETDEWEB)

Petoukhov, S.M.; Mathew, P.M

2001-10-01

To support the generic probabilistic safety analysis (PSA) program at AECL, in particular to conduct Level 2 PSA analysis of a CANDU 6 plant undergoing a postulated severe accident, the capability to conduct severe accident consequence analysis for a CANDU plant is required. For this purpose, AECL selected MAAP4 CANDU from a number of other severe accident codes. The necessary models for a generic CANDU 6 station have been implemented in the code, and the code version 0.2 beta was tested using station data, which were assembled for a generic CANDU 6 station. This paper describes the preliminary results of the consequence analysis using MAAP4 CANDU for a generic CANDU 6 station, when it undergoes a station blackout and a large loss-of-coolant accident scenario. The analysis results show that the plant response is consistent with the physical phenomena modeled and the failure criteria used. The results also confirm that the CANDU design is robust with respect to severe accidents, which is reflected in the calculated long times that are available for administering accident management measures to arrest the accident progression before the calandria vessel or containment become at risk. (author)

The Rising Frequency of IT Blackouts Indicates the Increasing Relevance of IT Emergency Concepts to Ensure Patient Safety.

Science.gov (United States)

Sax, Ulrich; Lipprandt, M; Röhrig, R

2016-11-10

As many medical workflows depend vastly on IT support, great demands are placed on the availability and accuracy of the applications involved. The cases of IT failure through ransomware at the beginning of 2016 are impressive examples of the dependence of clinical processes on IT. Although IT risk management attempts to reduce the risk of IT blackouts, the probability of partial/total data loss, or even worse, data falsification, is not zero. The objective of this paper is to present the state of the art with respect to strategies, processes, and governance to deal with the failure of IT systems. This article is conducted as a narrative review. Worst case scenarios are needed, dealing with methods as to how to survive the downtime of clinical systems, for example through alternative workflows. These workflows have to be trained regularly. We categorize the most important types of IT system failure, assess the usefulness of classic counter measures, and state that most risk management approaches fall short on exactly this matter. To ensure that continuous, evidence-based improvements to the recommendations for IT emergency concepts are made, it is essential that IT blackouts and IT disasters are reported, analyzed, and critically discussed. This requires changing from a culture of shame and blame to one of error and safety in healthcare IT. This change is finding its way into other disciplines in medicine. In addition, systematically planned and analyzed simulations of IT disaster may assist in IT emergency concept development.

OSSA. A second generation of severe accident management

International Nuclear Information System (INIS)

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

2009-01-01

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

Out of the Blackout and into the Light: How the Arts Survived Pinochet’s Dictatorship

Directory of Open Access Journals (Sweden)

Paula Thorrington Cronovich

2014-06-01

This article shows how various artists segued out of the cultural blackout of the late seventies and into a phase of surprising artistic production during the military regime in Chile. At a time when political parties were banned and public gatherings considered illegal, Chileans found alternative ways to oppose the military government. In this climate, I argue that artistic expression took on political meaning. The fact that the “No” Campaign of 1988 was able to oust the dictator with an optimistic message of joy and hope, attests to the point that Chileans were able to shed their fears and change their outlook. Throughout the decade, the arts—innovations in poetry, music, theater, narrative and the audiovisual media—had offered people a much-needed forum for expression.

Seismic PRA of a BWR plant

International Nuclear Information System (INIS)

Nishio, Masahide; Fujimoto, Haruo

2014-01-01

Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

Station blackout: Deterministic and probabilistic approach in the field of electrical supply losses by EDF

International Nuclear Information System (INIS)

Meslin, T.; Carnino, A.

1986-01-01

This example shows the thoroughness of EDF's approach in processing the difficult problems of the loss of electrical power supplies. Efforts are continuing in several directions: continued revision and improvement of operating procedures in the event of loss of electrical power supplies, PWR plant operator training courses devoted to the problems of power supply losses, and continued testing on simulators, and particularly testing under real conditions, including tests lasting several hours made possible by the performance of the new EDF simulators (two-phase code and taking all power losses into account)

Power and performance. Y2K challenges for electricity grids in Eastern Europe

International Nuclear Information System (INIS)

Kossilov, A.; Gueorguiev, B.; Ianev, I.; Purvis, E.

1999-01-01

The Year 2000 problem can directly affect the safety of nuclear power plants through interfaces with electric power and telecommunication systems. Recently, probabilistic safety assessments have made it clear that a 'station blackout' at a nuclear power plant is a major contributor to the sequence of events that could cause severe accidents. Within the IAEA actions concerned with Y2K problem, particular focus was on countries in eastern Europe, where here were delays in taking Y2K corrective actions

Contributions of the restructuring of the electric power industry to the August 14, 2003 blackout

International Nuclear Information System (INIS)

Casazza, J.; Delea, F.; Loehr, G.

2005-01-01

A review of the roles of industry and government in the 2003 blackout was presented. This white paper was prepared by a group of engineers with high level experience in the electric power industry who are concerned that deregulation of the industry has led to a significant decrease in reliability. It was noted that post-blackout reviews have focused on technical failures instead of examining the responsibilities and failures of the National Electric Reliability Council (NERC). Deficiencies in the analytical capabilities of control centres were discussed, as well as issues concerning communication protocols and training. Deregulation and the concomitant restructuring of the electric power industry has led to a shift from long term optimization, inter-system coordination and reliability towards dependence on immediate profits. In addition, there have been significant reductions in personnel at electric power organizations and companies, as well as increasing complexity in operations. Increased complexity has resulted in a dilution of management responsibility, as well as over-reliance on markets to solve scientifically complex problems. There have also been cutbacks in training and research. The functional separation of generation and transmission within companies has contributed to the diffusion of best technical knowledge. Many private utilities have divested their generation resources in response to regulatory pressures. The entrance of merchant power plants in the power system has led to the establishment of new market areas that are inconsistent with the boundaries of responsible operating entities. It was concluded that all these changes have created a more complicated and compartmentalized industry structure. Decisions are now made by a large number of entities, most of which are competitors and each of which has more interest in profit than in bulk power system reliability. Procedural rules established between and among the various parties are no longer

Analysis of some antecipated transients without scram for PWR type reactors by coupling of the CORAN code to the ALMOD code system

International Nuclear Information System (INIS)

Carvalho, F. de A.T. de.

1985-01-01

This study investigates some antecipated transients without scram for a pressurized water cooled reactor, using coupling of the containment CORAN code to the ALMOD code system, under severe random conditions. This coupling has the objective of including containment model as part of an unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle, a failure in the closure of the pressurizer relief valve was also investigated. (Author) [pt

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

International Nuclear Information System (INIS)

Ko Yuchih; Lee Min

2005-01-01

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

Utilization of Relap 5 computer code for analyzing thermohydraulic projects

International Nuclear Information System (INIS)

Silva Filho, E.

1987-01-01

This work deals with the design of a scaled test facility of a typical pressurized water reactor plant of the 1300 MW (electric) class. A station blackout has been choosen to investigate the thermohydraulic behaviour of the the test facility in comparison to the reactor plant. The computer code RELAPS/MOD1 has been utilized to simulate the blackout and to compare the test facility behaviour with the reactor plant one. The results demonstrate similar thermohydraulic behaviours of the two systems. (author) [pt

Evaluation of a severe accident management strategy for boiling water reactors -- Drywell flooding

International Nuclear Information System (INIS)

Yu, D.; Xing, L.; Kastenberg, W.E.; Okrent, D.

1994-01-01

Flooding of the drywell has been suggested as a strategy to prevent reactor vessel and containment failure in boiling water reactors. To evaluate the candidate strategy, this study considers accident management as a decision problem (''drywell flooding'' versus ''do nothing'') and develops a decision-oriented framework, namely, the influence diagram approach. This analysis chooses the long-term station blackout sequence for a Mark 1 nuclear power plant (Peach Bottom), and an influence diagram with a single decision node is constructed. The node probabilities in the influence diagram are obtained from US Nuclear Regulatory Commission reports or estimated by probabilistic risk assessment methodology. In assessing potential benefits compared with adverse effects, this analysis uses two consequence measures, i.e., early and late fatalities, as decision criteria. The analysis concludes that even though potential adverse effects exist, such as ex-vessel steam explosions and containment isolation failure, the drywell flooding strategy is preferred to ''do nothing'' when evaluated in terms of these consequence measures

47 CFR 76.128 - Application of sports blackout rules.

Science.gov (United States)

2010-10-01

... stations within whose specified zone the community of the community unit or the community within which the... other than major markets as defined in § 76.51, television broadcast stations within whose Grade B... specified zone. [65 FR 68101, Nov. 14, 2000, as amended at 67 FR 68951, Nov. 14, 2002] ...

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Re-evaluation of Station Blackout in Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

Effectiveness of containment sprays in containment management

International Nuclear Information System (INIS)

Nourbakhsh, H.P.; Perez, S.E.; Lehner, J.R.

1993-05-01

A limited study has been performed assessing the effectiveness of containment sprays-to mitigate particular challenges which may occur during a severe accident. Certain aspects of three specific topics related to using sprays under severe accident conditions were investigated. The first was the effectiveness of sprays connected to an alternate water supple and pumping source because the actual containment spray pumps are inoperable. This situation could occur during a station blackout. The second topic concerned the adverse as well as beneficial effects of using containment sprays during severe accident scenario where the containment atmosphere contains substantial quantities of hydrogen along with steam. The third topic was the feasibility of using containment sprays to moderate the consequences of DCH

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

Energy Technology Data Exchange (ETDEWEB)

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

1987-04-01

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

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

International Nuclear Information System (INIS)

Nomoto, Tsutomu

2015-01-01

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

Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

Energy Technology Data Exchange (ETDEWEB)

Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

2015-09-01

The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

Core damage frequency prespectives for BWR 3/4 and Westinghouse 4-loop plants based on IPE results

International Nuclear Information System (INIS)

Dingman, S.; Camp, S.; LaChance, J.; Mary Drouin

1995-01-01

This paper discusses the core damage frequency (CDF) insights gained by analyzing the results of the Individual Plant Examinations (IPES) for two groups of plants: boiling water reactor (BWR) 3/4 plants with Reactor Core Isolation Cooling systems, and Westinghouse 4-loop plants. Wide variability was observed for the plant CDFs and for the CDFs of the contributing accident classes. On average, transients-with loss of injection, station blackout sequences, and transients with loss of decay heat removal are important contributors for the BWR 3/4 plants, while transients, station blackout sequences, and loss-of-coolant accidents are important for the Westinghouse 4-loop plants. The key factors that contribute to the variability in the results are discussed. The results are often driven by plant-specific design and operational characteristics, but differences in modeling approaches are also important for some accident classes

Millstone 3 risk evaluation report. An overall review and evaluation of the Millstone Unit 3 probabilistic safety study

International Nuclear Information System (INIS)

Kelly, G.; Barrett, R.; Buslik, A.

1986-06-01

In 1981, the US Nuclear Regulatory Commission (NRC) requested Northeast Utilities to perform a design-specific probabilistic safety study (PSS) for Millstone Nuclear Power Station, Unit No. 3 (Millstone 3). In 1983, Northeast Utilities submitted the Millstone 3 Probabilistic Safety Study for review by the NRC staff. The NRC staff prepared the Millstone 3 Risk Evaluation Report, which discusses the findings regarding the PSS. The PSS estimates that the mean annual core damage frequency due to internal and external events is 5 x 10 -5 and 2 x 10 -5 , respectively. The NRC staff's Risk Evaluation Report estimates that the mean annual core damage frequency is about 2 x 10 -4 for internal events and lies between 1 x 10 -5 and 2 x 10 -4 for external events. The NRC staff estimates that station blackout dominates internal and external event core damage frequencies. The staff recommends that Northeast Utilities perform an engineering analysis on upgrading the diesel generator lube oil cooler anchorage system and on adding a manually operated, AC-independent containment spray system. The staff also recommends that Northeast Utilities prepare two emergency procedures (loss of room cooling and relay chatter due to an earthquake) to help reduce uncertainties. (Subsequent to the completion of this document, Northeast Utilities and the NRC staff have continued a dialogue regarding station blackout from events other than earthquakes. Both Northeast Utilities and the staff have performed additional evaluations, which have drawn their results closer together. Final requirements, if any, for the prevention or mitigation of station blackout from events other than earthquakes have not yet been determined.) 26 refs., 16 tabs

Quantification of uncertainties in source term estimates for a BWR with Mark I containment

International Nuclear Information System (INIS)

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

1988-01-01

A methodology for quantification and uncertainty analysis of source terms for severe accident in light water reactors (QUASAR) has been developed. The objectives of the QUASAR program are (1) to develop a framework for performing an uncertainty evaluation of the input parameters of the phenomenological models used in the Source Term Code Package (STCP), and (2) to quantify the uncertainties in certain phenomenological aspects of source terms (that are not modeled by STCP) using state-of-the-art methods. The QUASAR methodology consists of (1) screening sensitivity analysis, where the most sensitive input variables are selected for detailed uncertainty analysis, (2) uncertainty analysis, where probability density functions (PDFs) are established for the parameters identified by the screening stage and propagated through the codes to obtain PDFs for the outputs (i.e., release fractions to the environment), and (3) distribution sensitivity analysis, which is performed to determine the sensitivity of the output PDFs to the input PDFs. In this paper attention is limited to a single accident progression sequence, namely; a station blackout accident in a BWR with a Mark I containment buildings. Identified as an important accident in the draft NUREG-1150 a station blackout involves loss of both off-site power and DC power resulting in failure of the diesels to start and in the unavailability of the high pressure injection and core isolation coding systems

Study on depressurization measurements and effect in PWR

International Nuclear Information System (INIS)

Ji Duan; Cao Xuewu

2006-01-01

Implementation of new regulations on nuclear powered plant design and operation raise new design and management requirement for plants, and the operational plants also need accident management to enhance the reactor operation safety. Thus, for sake of reducing risk of high-pressure and mitigating the consequence, depressurization is a measure carried out to reduce primary pressure. With SCDAP/RELAP5 this paper studies the depressurization measurements and effect factors in pressurized water reactor under the important severe accident sequences induced by very small break lost of coolant accident (VSBLOCA), anticipated transient without scram (ATWS) and station blackout (SBO) plus auxiliary feedwater failure. (author)

Emergent transformation games: exploring social innovation agency and activation through the case of the Belgian electricity blackout threat

Directory of Open Access Journals (Sweden)

Bonno Pel

2016-06-01

Full Text Available The persistence of current societal problems has given rise to a quest for transformative social innovations. As social innovation actors seek to become change makers, it has been suggested that they need to play into impactful macrodevelopments or "game-changers". Here, we aim to deepen the understanding of the social innovation agency in these transformation games. We analyze assumptions about the game metaphor, invoking insights from actor-network theory. The very emergence of transformation games is identified as a crucial but easily overlooked issue. As explored through the recent electricity blackout threat in Belgium, some current transformation games are populated with largely passive players. This illustrative case demonstrates that socially innovative agency cannot be presupposed. In some transformation games, the crucial game-changing effect is to start the game by activating the players.

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Source Term Analysis for the Nuclear Power Station Goesgen-Daeniken; Quelltermanalysen fuer das Kernkraftwerk Goesgen-Daeniken

Energy Technology Data Exchange (ETDEWEB)

Hosemann, J.P.; Megaritis, G.; Guentay, S.; Hirschmann, H.; Luebbesmeyer, D.; Lieber, K.; Jaeckel, B.; Birchley, J.; Duijvestijn, G

2001-08-01

Analyses are performed for three accident scenarios postulated to occur in the Goesgen Nuclear Power Plant, a 900 MWe Pressurised Water Reactor of Siemens design. The scenarios investigated comprise a Station Blackout and two separate cases of small break loss-of-coolant accident which lead, respectively, to high, intermediate and low pressure conditions in the reactor system. In each case the accident assumptions are highly pessimistic, so that the sequences span a large range of plant states and a damage phenomena. Thus the plant is evaluated for a diversity of potential safety challenges. A suite of analysis tools are used to examine the reactor coolant system response, the core heat-up, melting, fission product release from the reactor system, the transport and chemical behaviour of those fission products in the containment building, and the release of radioactivity (source term) to the environment. Comparison with reference values used by the licensing authority shows that the use of modern analysis tools and current knowledge can provide substantial reduction in the estimated source term. Of particular interest are insights gained from the analyses which indicate opportunities for operators to reduce or forestall the release. (author)

From city’s station to station city. An integrative spatial approach to the (redevelopment of station areas

Directory of Open Access Journals (Sweden)

Ana Luísa Martins da Conceição

2014-12-01

Full Text Available Since its origin, the railway station has had a complicated relationship with the city, demanding periodical updates, particularly regarding spatial issues. With the aim of improving the liveability of station areas, current redevelopment projects are reconceptualising them as balanced transport ‘nodes’ and ‘places’ in the city. However, the proposed spatial solutions do not fully support the sought after economic, social and environmental performances. These intentions continue to be predominantly bounded with the (abstract planological level, not finding appropriate translation at the (concrete spatial design level. Further, the interdisciplinary nature of the highly complex planning and design processes of station areas, which should contribute to enhance the performance of their spaces, reinforces constraints and relegates architecture to a marginal role in this quest. It is thus necessary to understand how architecture can contribute to the improvement of the spatial performance of contemporary stations areas, supporting their current reconceptualization. To gain this understanding, the research explored the factors which influence the spatial definition and performance of European High Speed Train station areas, using “design research” and “research by design”. Via a theoretical integrative framework, synthesized from knowledge developed by architecture and other sciences, case studies of ‘through’ stations were analysed and compared. Six cases, encapsulating the most recurrent relative positions of the railway (infrastructure and the station building towards the(ir direct built environment, were chosen out of a large sample. For each category (cases with railway tracks at (a ground level, (b elevated level and (c underground level, two cases, featuring an adapted station building and a newly built one, were studied. Their physical and functional characteristics were mapped at several scales and moments (in history, as

UMTS Network Stations

Science.gov (United States)

Hernandez, C.

2010-09-01

The weakness of small island electrical grids implies a handicap for the electrical generation with renewable energy sources. With the intention of maximizing the installation of photovoltaic generators in the Canary Islands, arises the need to develop a solar forecasting system that allows knowing in advance the amount of PV generated electricity that will be going into the grid, from the installed PV power plants installed in the island. The forecasting tools need to get feedback from real weather data in "real time" from remote weather stations. Nevertheless, the transference of this data to the calculation computer servers is very complicated with the old point to point telecommunication systems that, neither allow the transfer of data from several remote weather stations simultaneously nor high frequency of sampling of weather parameters due to slowness of the connection. This one project has developed a telecommunications infrastructure that allows sensorizadas remote stations, to send data of its sensors, once every minute and simultaneously, to the calculation server running the solar forecasting numerical models. For it, the Canary Islands Institute of Technology has added a sophisticated communications network to its 30 weather stations measuring irradiation at strategic sites, areas with high penetration of photovoltaic generation or that have potential to host in the future photovoltaic power plants connected to the grid. In each one of the stations, irradiance and temperature measurement instruments have been installed, over inclined silicon cell, global radiation on horizontal surface and room temperature. Mobile telephone devices have been installed and programmed in each one of the weather stations, which allow the transfer of their data taking advantage of the UMTS service offered by the local telephone operator. Every minute the computer server running the numerical weather forecasting models receives data inputs from 120 instruments distributed

Assessment of impact of a severe accident at nuclear power plant of Angra dos Reis with release of radionuclides to the atmosphere

International Nuclear Information System (INIS)

Aguiar, Andre Silva de

2015-01-01

This study had as purpose the assess the impact of a severe accident, and also analyze the dispersion of 131 I in the atmosphere, so that, through concentrating and inhaling dose of the plume, were possible to verify if the results are in accordance with the indicated data by the Plan of Emergency of the CNAAA regarding the Impact Zone and Control. This exercise was performed with the aid of an atmospheric model and a dispersion where to atmospheric modeling we used the data coupling WRF / CALMET and of dispersion, CALPUFF. The suggested accident consists of a Station Blackout at Nuclear Power of Angra (Unit 1), where through the total core involvement, will release 100% of the 131 I to the atmosphere. The value of the total activity in the nucleus to this radionuclide is 7.44 x 1017 Bq, that is relative on the sixth day of burning. This activity will be released through the chimney at a rate in Bq/s in the scenario of 12, 24, 48 and 72 hours of release. Applying the model in the proposed scenario, it is verified that the plume has concentrations of the order of 1020 Bq/m³ and dose of about 108 Sv whose value is beyond of the presented by Eletronuclear in your current emergency plan. (author)

Avoiding Service Station Fraud.

Science.gov (United States)

Burton, Grace M.; Burton, John R.

1982-01-01

High school students are warned against service station fraud. A problem-solving section is designed to help students calculate consumer costs for various fraudulent transactions. Several ways of reducing fraud or of lessening the chances of problems are noted. (MP)

Establishment of Karadeniz Technical University Permanent GNSS Station as Reactivated of TRAB IGS Station

Directory of Open Access Journals (Sweden)

Kazancı Selma Zengin

2017-12-01

Full Text Available In recent years, Global Navigation Satellite Systems (GNSS have gained great importance in terms of the benefi ts it provides such as precise geodetic point positioning, determining crustal deformations, navigation, vehicle monitoring systems and meteorological applications etc. As in Turkey, for this purpose, each country has set up its own GNSS station networks like Turkish National Permanent RTK Network analyzed precise station coordinates and velocities together with the International GNSS Service, Turkish National Fundamental GPS Network and Turkish National Permanent GNSS Network (TNPGN stations not only are utilized as precise positioning but also GNSS meteorology studies so total number of stations are increased. This work is related to the reactivated of the TRAB IGS station which was established in Karadeniz Technical University, Department of Geomatics Engineering. Within the COST ES1206 Action (GNSS4SWEC KTU analysis center was established and Trop-NET system developed by Geodetic Observatory Pecny (GOP, RIGTC in order to troposphere monitoring. The project titled “Using Regional GNSS Networks to Strengthen Severe Weather Prediction” was accepted to the scientifi c and technological research council of Turkey (TUBITAK. With this project, we will design 2 new constructed GNSS reference station network. Using observation data of network, we will compare water vapor distribution derived by GNSS Meteorology and GNSS Tomography. At this time, KTU AC was accepted as E-GVAP Analysis Centre in December 2016. KTU reference station is aimed to be a member of the EUREF network with these studies.

Establishment of Karadeniz Technical University Permanent GNSS Station as Reactivated of TRAB IGS Station

Science.gov (United States)

Kazancı, Selma Zengin; Kayıkçı, Emine Tanır

2017-12-01

In recent years, Global Navigation Satellite Systems (GNSS) have gained great importance in terms of the benefi ts it provides such as precise geodetic point positioning, determining crustal deformations, navigation, vehicle monitoring systems and meteorological applications etc. As in Turkey, for this purpose, each country has set up its own GNSS station networks like Turkish National Permanent RTK Network analyzed precise station coordinates and velocities together with the International GNSS Service, Turkish National Fundamental GPS Network and Turkish National Permanent GNSS Network (TNPGN) stations not only are utilized as precise positioning but also GNSS meteorology studies so total number of stations are increased. This work is related to the reactivated of the TRAB IGS station which was established in Karadeniz Technical University, Department of Geomatics Engineering. Within the COST ES1206 Action (GNSS4SWEC) KTU analysis center was established and Trop-NET system developed by Geodetic Observatory Pecny (GOP, RIGTC) in order to troposphere monitoring. The project titled "Using Regional GNSS Networks to Strengthen Severe Weather Prediction" was accepted to the scientifi c and technological research council of Turkey (TUBITAK). With this project, we will design 2 new constructed GNSS reference station network. Using observation data of network, we will compare water vapor distribution derived by GNSS Meteorology and GNSS Tomography. At this time, KTU AC was accepted as E-GVAP Analysis Centre in December 2016. KTU reference station is aimed to be a member of the EUREF network with these studies.

A Sounding-based Severe Weather Tool to Support Daily Operations at Kennedy Space Center and Cape Canaveral Air Force Station

Science.gov (United States)

Bauman, William H.; Roeder, William P.

2014-01-01

People and property at Kennedy Space Center (KSC) and Cape Canaveral Air Force Station (CCAFS) are at risk when severe weather occurs. Strong winds, hail and tornadoes can injure individuals and cause costly damage to structures if not properly protected. NASA's Launch Services Program and Ground Systems Development and Operations Program and other KSC programs use the daily and weekly severe weather forecasts issued by the 45th Weather Squadron (45 WS) to determine if they need to limit an activity such as working on gantries, or protect property such as a vehicle on a pad. The 45 WS requested the Applied Meteorology Unit (AMU) develop a warm season (May-September) severe weather tool for use in the Meteorological Interactive Data Display System (MIDDS) based on the late morning, 1500 UTC (1100 local time), CCAFS (XMR) sounding. The 45 WS frequently makes decisions to issue a severe weather watch and other severe weather warning support products to NASA and the 45th Space Wing in the late morning, after the 1500 UTC sounding. The results of this work indicate that certain stability indices based on the late morning XMR soundings can depict differences between days with reported severe weather and days with no reported severe weather. The AMU determined a frequency of reported severe weather for the stability indices and implemented an operational tool in MIDDS.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

failure. Firstly, the life time of emergency batteries after the station blackout (SBO) was insufficient to operate targeted instruments. Secondly, the absence of proper protection in the containment building for extremely harsh environment after severe accident caused malfunctions. Lastly, since the power or communication cable was cut off, the instruments in the containment building could not transmit information to the outside, or their power source could be lost because most of the equipment in NPPs is wired-system based.

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

International Nuclear Information System (INIS)

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

2013-01-01

failure. Firstly, the life time of emergency batteries after the station blackout (SBO) was insufficient to operate targeted instruments. Secondly, the absence of proper protection in the containment building for extremely harsh environment after severe accident caused malfunctions. Lastly, since the power or communication cable was cut off, the instruments in the containment building could not transmit information to the outside, or their power source could be lost because most of the equipment in NPPs is wired-system based

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Passive cooling applications for nuclear power plants using pulsating steam-water heat pipes

International Nuclear Information System (INIS)

Aparna, J.; Chandraker, D.K.

2015-01-01

Gen IV reactors incorporate passive principles in their system design as an important safety philosophy. Passive safety systems use inherent physical phenomena for delivering the desired safe action without any external inputs or intrusion. The accidents in Fukushima have renewed the focus on passive self-manageable systems capable of unattended operation, for long hours even in extended station blackout (SBO) and severe accident conditions. Generally, advanced reactors use water or atmospheric air as their ultimate heat sink and employ passive principles in design for enhanced safety. This paper would be discussing the experimental results on pulsating steam water heat-pipe devices and their applications in passive cooling. (author)

Program plan for environmental qualification of mechanical and dynamic (including seismic) qualification of mechanical and electrical equipment program (EDQP)

International Nuclear Information System (INIS)

Weidenhamer, G.H.

1986-06-01

The equipment qualification program described in this plan is intended to provide the technical basis for resolving uncertainties in existing equipment qualification standards. In addition, research results are contributing to the resolution of safety issues GI-23, GI-87, USI-A44, titled, ''Reactor Coolant Pump Seal Failure,'' ''Failure of HPCI Steam Line Without Isolation,'' and ''Station Blackout,'' respectively. Also, research effort is being directed at providing information on the behavior of containment isolation valves under severe accident environments. Although the results of the latter research will not contribute to resolving uncertainties in qualification standards, it has proven cost effective to obtain this information under this program

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

International Nuclear Information System (INIS)

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

2017-02-01

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

Principles of nuclear power station control

International Nuclear Information System (INIS)

Knowles, J.B.

1975-12-01

This memorandum represents lecture notes first distributed as part of a UKAEA introductory course on Reactor Technology held during November 1975. A nuclear power station is only one element of a dispersed interconnected arrangement of other nuclear and fossil-fired units which together constitute the national 'grid'. Thus the control of any one station must relate to the objectives of the grid network as a whole. A precise control of the supply frequency of the grid is achieved by regulating the output power of individual stations, and it is necessary for each station to be stable when operating in isolation with a variable load. As regards individual stations, several special control problems concerned with individual plant items are discussed, such as: controlled reactivity insertions, temperature reactivity time constants and flow instability. A simplified analysis establishes a fundamental relationship between the stored thermal energy of a boiler unit (a function of mechanical construction) and the flexibility of the heat source (nuclear or fossil-fired) if the station is to cope satisfactorily with demands arising from unscheduled losses of other generating sets or transmission capacity. Two basic control schemes for power station operation are described known as 'coupled' and 'decoupled control'. Each of the control modes has its own merits, which depend on the proposed station operating strategy (base load or load following) and the nature of the heat source. (U.K.)

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

International Nuclear Information System (INIS)

Uspuras, E.; Kaliatka, A.

2001-01-01

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

Operation feedback of hydrogen filling station

International Nuclear Information System (INIS)

Pregassame, S.; Barral, K.; Allidieres, L.; Charbonneau, T.; Lacombe, Y.

2004-01-01

One of the technical challenges of hydrogen technology is the development of hydrogen infrastructures which satisfy either safety requirements and reliability of filling processes. AIR LIQUIDE realized an hydrogen filling station in Sassenage (France) operational since September 2003. This station is able to fill 3 buses a day up to 350bar by equilibrium with high pressure buffers. In parallel with commercial stations, the group wanted to create a testing ground in real conditions running with several objectives: validate on a full scale bench a simulation tool able to predict the temperature of both gas and cylinder's materials during filling processes; define the best filling procedures in order to reach mass, temperature and filling time targets; analyse the temperature distribution and evolution inside the cylinder; get a general knowledge about hydrogen stations from safety and reliability point of view; operate the first full scale refuelling station in France. The station is also up-graded for 700bar filling from either a liquid hydrogen source or a gas booster, with cold filling possibility. This paper presents the results concerning 350bar filling : thermal effects, optimal filling procedures and influence of parameters such as climatic conditions are discussed. (author)

Polycapillary based μXRF station for 3D colour tomography

Science.gov (United States)

Hampai, D.; Cherepennikov, Yu. M.; Liedl, A.; Cappuccio, G.; Capitolo, E.; Iannarelli, M.; Azzutti, C.; Gladkikh, Yu. P.; Marcelli, A.; Dabagov, S. B.

2018-04-01

The "Rainbow X-Ray" (RXR) experimental station at XLab Frascati of the Frascati's National Laboratories (LNF) INFN is a dedicated station for X-ray fluorescence studies based on the use of polycapillary lenses in a confocal geometry. The flexible RXR layout allows investigating specimens of the dimensions ranging from several millimeters up to half meter and weighting up to several tens of kilograms. Compared to similar existing XRF stations, apart of the possibility for investigating large samples, the main advantage of this equipment is the detection system with two spectrometers optimized to work separately at high and at low X-ray energies. The confocal geometry combined with a 3-axes fine motion system makes possible 3D μXRF elemental tomographic acquisitions (colour tomography). At present this station in operation at high XRF energies is used for cultural heritage and geological applications. We present and discuss here the analytical performances of this experimental station pointing out the advantages in different application areas.

H2FIRST Reference Station Design Task: Project Deliverable 2-2

Energy Technology Data Exchange (ETDEWEB)

Pratt, Joseph [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Terlip, Danny [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Ainscough, Chris [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Kurtz, Jennifer [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Elgowainy, Amgad [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-04-20

This report presents near-term station cost results and discusses cost trends of different station types. It compares various vehicle rollout scenarios and projects realistic near-term station utilization values using the station infrastructure rollout in California as an example. It describes near-term market demands and matches those to cost-effective station concepts. Finally, the report contains detailed designs for five selected stations, which include piping and instrumentation diagrams, bills of materials, and several site-specific layout studies that incorporate the setbacks required by NFPA 2, the National Fire Protection Association Hydrogen Technologies Code. This work identified those setbacks as a significant factor affecting the ability to site a hydrogen station, particularly liquid stations at existing gasoline stations. For all station types, utilization has a large influence on the financial viability of the station.

Evaluation of severe accident risks and the potential for risk reduction: Peach Bottom, Unit 2. Main report. Draft for comment, February 1987

Energy Technology Data Exchange (ETDEWEB)

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

1987-04-01

The Severe Accident Risk Reduction Program (SARRP) has completed a rebaselining of the risks to the public from a boiling water reactor with a Mark I containment (Peach Bottom, Unit 2). Emphasis was placed on determining the magnitude and character of the uncertainties, rather than focusing on a point estimate. The risk-reduction potential of a set of proposed safety option backfits was also studied, and their costs and benefits were also evaluated. It was found that the risks from internal events are generally low relative to previous studies; for example, most of the uncertainty range is lower than the point estimate of risk for the Peach Bottom plant in the Reactor Safety Study (RSS). However, certain unresolved issues cause the top of the uncertainty band to appear at a level that is comparable with the RSS point estimate. These issues include the modeling of the common-mode failures for the dc power system, the likelihood of offsite power recovery versus time during a station blackout, the probability of drywell failure resulting from meltthrough of the drywell shell, the magnitude of the fission product releases during core-concrete interactions, and the decontamination effectiveness of the reactor enclosure building. Most of the postulated safety options do not appear to be cost effective, although some based on changes to procedures or inexpensive hardware additions may be marginally cost effective. This draft for comment of the SARRP report for Peach Bottom does not include detailed technical appendices, which are still in preparation. The appendices will be issued under separate cover when completed. This work supports the Nuclear Regulatory Commission's assessment of severe accidents in NUREG-1150. (author)

Summary of the Current Status of Lessons Learned From Fukushima Accident

International Nuclear Information System (INIS)

Pasamehmetoglu, Kemal

2013-01-01

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

Analysis of a Natural Circulation in the Reactor Coolant System Following a High Pressure Severe Accident at APR1400

International Nuclear Information System (INIS)

Kim, Han Chul; Cho, Yong Jin; Park, Jae Hong; Cho, Song Won

2011-01-01

Under a high temperature and pressure condition during a severe accident, hot leg pipes or steam generator tubes could fail due to creep rupture following natural circulation in the Reactor Coolant System (RCS) unless depressurization of the system is performed at a proper time. Natural circulation in the RCS can be a multi-dimensional circulation in the reactor vessel, a partial loop circulation of two-phase flow from the core up to steam generators (SGs), or circulation in the total loop. It can delay the reactor vessel failure time by removing heat from the reactor core. This natural phenomenon can be hardly simulated with a single flow path model for the hot spots of the RCS, since it cannot deal with the counter-current flow. Thus it may estimate accident progression faster than reality, which may cause troubles for optimized implementation of severe accident management strategies. An earlier damage in the RCS other than the reactor pressure vessel may make subsequent behaviors of hydrogen or fission products in the containment quite different from the single reactor vessel failure. Therefore, a RCS model which treats natural circulation is needed to evaluate the RCS response and the safety depressurization strategy in a best-estimate way. The aim of this study is to develop a detailed model which allows natural circulation between the reactor vessel and steam generators through hot legs, based on the existing APR1400 RCS model. The station blackout sequence was selected to be the representative high-pressure scenario. Sensitivity study on the effect of node configuration of the upper plenum and addition of cross flow paths from the upper plenum to the hot legs were carried out. This model is described herein and representative calculation results are presented

Tests of Shaft Seal Systems of Circulation Pumps during Station Blackout

Energy Technology Data Exchange (ETDEWEB)

Beisiegel, A.; Foppe, F.; Wich, M.

2014-07-01

AREVA GmbH operates a unique Thermal-hydraulic plat form in Germany, France and USA. It is recognised as a test body according to ISO 17025. The Deutsche Akkreditierungsstelle GmbH (DAkkS - German Society for Accreditation) has also certified the Thermal-hydraulic platform as an independent inspection body Type C according to ISO 17020. A part of this platform is the Component Laboratory located in Karlstein, Germany which is in operation since more than 50 years. The testing activities cover a wide range as: Critical Heat Flux Tests, Valve Testing and Environmental Qualification for safety related components. Since 2011 the Component Qualification Karlstein extended their testing scope for different types of Shaft Seal Systems. (Author)

Risk analysis of NPP in multi-unit site for configuration of AAC power source

International Nuclear Information System (INIS)

Kim, Myung Ki

2000-01-01

Because of the difficulties in finding new sites for nuclear power plants, more units are being added to the existing sites. In these multi-unit sites, appropriate countermeasures should be established to cope with the potential station blackout (SBO) accident. Currently, installation of additional diesel generator (DG) is considered to ensure an alternative AC power source, but it has not been decided yet how many DGs should be installed in a multi-unit site. In this paper, risk informed decision making method, which evaluates reliability of electrical system, core damage frequency, and site average core damage frequency, is introduced to draw up the suitable number of DG in multi-unit site. The analysis results show that installing two DGs lowered the site average core damage frequency by 1.4% compared to one DG in six unit site. In the light of risk-informed decisions in regulatory guide 1.174, there is no difference of safety between two alternatives. It is concluded that one emergency diesel generator sufficiently guarantees safety against station blackout of nuclear power plants in multi-unit site. (author)

Heat transfer characteristics of horizontal steam generators under natural circulation conditions

International Nuclear Information System (INIS)

Hyvaerinen, J.

1996-01-01

This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent limitation of the decay energy removal in VVER steam generators. The limitation arises as a consequence of the steam generator tube bundle geometry. As an example, it is shown that the grace period associated with pressurizer safety valve opening during a station black-out is 2 1/2-3 hours instead of the 5-6 hours reported in several earlier studies. (However, the change in core heat-up timing is much less-about 1 h at most.) The heat transfer limitation explains the fact that, in the Greifswald VVER 440 station black-out accident in 1975, the steam generators never boiled dry. In addition, the stability of single-phase natural circulation is discussed and insights on the modelling of horizontal steam generators with general-purpose thermal-hydraulic system codes are also presented. (orig.)

Imaging Total Stations - Modular and Integrated Concepts

Science.gov (United States)

Hauth, Stefan; Schlüter, Martin

2010-05-01

Keywords: 3D-Metrology, Engineering Geodesy, Digital Image Processing Initialized in 2009, the Institute for Spatial Information and Surveying Technology i3mainz, Mainz University of Applied Sciences, forces research towards modular concepts for imaging total stations. On the one hand, this research is driven by the successful setup of high precision imaging motor theodolites in the near past, on the other hand it is pushed by the actual introduction of integrated imaging total stations to the positioning market by the manufacturers Topcon and Trimble. Modular concepts for imaging total stations are manufacturer independent to a large extent and consist of a particular combination of accessory hardware, software and algorithmic procedures. The hardware part consists mainly of an interchangeable eyepiece adapter offering opportunities for digital imaging and motorized focus control. An easy assembly and disassembly in the field is possible allowing the user to switch between the classical and the imaging use of a robotic total station. The software part primarily has to ensure hardware control, but several level of algorithmic support might be added and have to be distinguished. Algorithmic procedures allow to reach several levels of calibration concerning the geometry of the external digital camera and the total station. We deliver insight in our recent developments and quality characteristics. Both the modular and the integrated approach seem to have its individual strengths and weaknesses. Therefore we expect that both approaches might point at different target applications. Our aim is a better understanding of appropriate applications for robotic imaging total stations. First results are presented. Stefan Hauth, Martin Schlüter i3mainz - Institut für Raumbezogene Informations- und Messtechnik FH Mainz University of Applied Sciences Lucy-Hillebrand-Straße 2, 55128 Mainz, Germany

Development of Methodology for Spent Fuel Pool Severe Accident Analysis Using MELCOR Program

Energy Technology Data Exchange (ETDEWEB)

Kim, Won-Tae; Shin, Jae-Uk [RETech. Co. LTD., Yongin (Korea, Republic of); Ahn, Kwang-Il [KAERI, Daejeon (Korea, Republic of)

2015-05-15

The general reason why SFP severe accident analysis has to be considered is that there is a potential great risk due to the huge number of fuel assemblies and no containment in a SFP building. In most cases, the SFP building is vulnerable to external damage or attack. In contrary, low decay heat of fuel assemblies may make the accident processes slow compared to the accident in reactor core because of a great deal of water. In short, its severity of consequence cannot exclude the consideration of SFP risk management. The U.S. Nuclear Regulatory Commission has performed the consequence studies of postulated spent fuel pool accident. The Fukushima-Daiichi accident has accelerated the needs for the consequence studies of postulated spent fuel pool accidents, causing the nuclear industry and regulatory bodies to reexamine several assumptions concerning beyond-design basis events such as a station blackout. The tsunami brought about the loss of coolant accident, leading to the explosion of hydrogen in the SFP building. Analyses of SFP accident processes in the case of a loss of coolant with no heat removal have studied. Few studies however have focused on a long term process of SFP severe accident under no mitigation action such as a water makeup to SFP. USNRC and OECD have co-worked to examine the behavior of PWR fuel assemblies under severe accident conditions in a spent fuel rack. In support of the investigation, several new features of MELCOR model have been added to simulate both BWR fuel assembly and PWR 17 x 17 assembly in a spent fuel pool rack undergoing severe accident conditions. The purpose of the study in this paper is to develop a methodology of the long-term analysis for the plant level SFP severe accident by using the new-featured MELCOR program in the OPR-1000 Nuclear Power Plant. The study is to investigate the ability of MELCOR in predicting an entire process of SFP severe accident phenomena including the molten corium and concrete reaction. The

Preliminary thermal design of a pressurized water reactor containment for handling severe accident consequences

International Nuclear Information System (INIS)

Abdullah, A.M.; Karameldin, A.

1998-01-01

A one-dimensional mathematical model has been developed for a 4250 MW(th) Advanced Pressurized Water Reactor containment analysis following a severe accident. The cooling process of the composite containment-steel shell and concrete shield- is achievable by natural circulation of atmospheric air. However, for purpose of gettering higher degrees of safety margin, the present study undertakes two objectives: (1) Installment of a diesel engine-driven air blower to force air through the annular space between the steel shell and concrete shield. The engine can be remotely operated to be effective in case of station blackout. (ii) Fixing longitudinally plate fins on the circumference of the inside and outside containment steel shell. These fins increase the heat transfer areas and hence the rate of heat removal from the containment atmosphere. In view of its importance - from the safety viewpoint - the long term behaviour of the containment which is a quasi-steady state problem, is formulated through a system of coupled nonlinear algebraic equations which describe the thermal-hydraulic and thermodynamic behaviour of the double shell containment. The calculated results revealed the following: (i) the passively air cooled containment can remove maximum heat load of 11.5 MW without failure, (ii) the effect of finned surface in the air passage tends to decrease the containment pressure by 20 to 30%, depending on the heat load, (iii) the effect of condensing fins is negligible for the proposed fin dimensions and material. However, by reducing the fin width, increasing their thickness, doubling their number, and using a higher conductive metal than the steel, it is expected that the containment pressure can be further reduced by 10% or more, (iv) the fins' dimensions and their number must be optimized via maximizing the difference or the ratio between the heat removed and pressure drop to get maximum heat flow rate

Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

Energy Technology Data Exchange (ETDEWEB)

Fynan, Douglas A.; Ahn, Kwang-Il, E-mail: kiahn@kaeri.re.kr

2016-12-15

Highlights: • Pressure drop-flow rate curves for superheated steam in U-tubes were generated. • Forward flow of hot steam is favored in the longer and taller U-tubes. • Reverse flow of cold steam is favored in short U-tubes. • Steam generator U-tube bundle geometry and tube diameter are important. • Need for correlation development for natural convention heat transfer coefficient. - Abstract: Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.

IAEA Completes Expert Mission to Kori 1 Nuclear Power Plant in the Republic of Korea

International Nuclear Information System (INIS)

2012-01-01

further corrective actions in safety culture, process and design enhancements. The expert mission team performed its own analysis of the station blackout event, including the non-reporting of the event, the violation of plant Technical Specifications by not immediately taking steps restore emergency diesel generators to operable conditions, and by starting a fuel movement the next day without operable emergency diesel generators. As a result of this analysis, the team identified additional contributing causes and corrective actions for Kori 1 NPP. Some of the important contributing causes identified by the expert mission include: Shortcomings in safety culture at Kori 1 NPP led to an inability to counter the errors in handling the station blackout event and the subsequent leadership failures in communication and reporting; The potential consequences of the plant's maintenance schedule changes were inadequately considered; and Kori 1 NPP internal oversight systems failed to reveal the station blackout event. The expert mission made recommendations to address these contributing factors, including to improve safety culture at the plant, to support leadership in decision-making, and to develop stronger defences in dealing with unexpected or difficult situations. In addition, the team made recommendations and suggestions concerning the plant's practice of reporting events, event analysis, independent oversight, main control room activities and the workmanship of maintenance staff and contractors. Implementation of the IAEA recommendations and suggestions will considerably improve safety at Kori 1 NPP. Plant management stated its commitment to address all the areas identified for improvement. The expert mission also identified good plant conditions and programmes, including: Detailed visits of the plant confirmed the good condition of equipment, including that which caused the station blackout, which has since been improved; Kori 1 NPP has implemented several equipment

Parameters of passenger facilities according to railway station characteristics

Directory of Open Access Journals (Sweden)

Ondřej HAVLENA

2014-12-01

Full Text Available The article presents ways and goals of categorization of railway stations and stops in the Czech railway network. The aim of this categorization is to classify railway stations (or stops in the railway network according to a suite of entrance parameters (e.g. municipality population, transfer links, job opportunities, tourist attractiveness. On the basis of these parameters, railway stations and stops will be classified into several categories, which will be used to specify the conclusions for station equipment concerning ticket offices, commercial services, waiting rooms and other accessories. Research results can be used as a support for infrastructure managers and railway operators to optimise the scale of their services.

Attempt to evaluate the cumulative deposit of strontium 90 in several French centres; Essai d'evaluation du depot cumulatif de strontium 90 dans plusieurs stations francaises

Energy Technology Data Exchange (ETDEWEB)

Garnier, A [Commissariat a l' Energie Atomique, Fontenay aux Roses (France). Centre d' Etudes Nucleaires

1966-07-01

In order to be able to interpret results of radioactivity controls of the ambient atmosphere and of the supply chain carried out in several French centres, it is necessary to know the value of the cumulative strontium-90 deposit resulting from fall-out. The estimation of the fall-out which occurred locally in the years before systematic controls were established can be made by assuming that the {sup 90}Sr concentration in rain varies little from one place to another (if they are fairly close) and that the deposit can be deduced from that measured in a single station by making corrections for differences in the rainfall. This assumption is confirmed by applying the method to the calculation of {sup 90}Sr fall-out at Ispra, Mol and at several French centres: the values calculated are compared to those obtained by measurements carried out since 1958 in some cases, and since 1960 in others. The results show that the method is valid for calculating fall-out prior to 1960 (in the absence of information) and to deduce with sufficient approximation the subsequent cumulative deposits in the French centres under consideration. (author) [French] L'evaluation du depot cumulatif de Sr 90 consecutif aux retombees est necessaire a l'interpretation des resultats du controle de la radioactivite du milieu ambiant et de la cha e alimentaire effectue en plusieurs stations francaises. L'estimation des retombees locale des annees anterieures a la mise en oeuvre systematique des moyens de surveillance peut etre faite en admettant que la concentration de Sr 90 dans la pluie varie peu d'un endroit a l'autre (dans des regions assez rapprochees) et que le depot peut etre deduit de celui mesure en une station gr e a des corrections tenant compte de la variation de pluviosite. Cette hypothese est verifiee en appliquant la methode au calcul des retombees de Sr 90 a Ispra, a Mol, et en plusieurs stations francaises: les valeurs calculees sont comparees a celles obtenues par les mesures

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

Hydrogen Fueling Station Using Thermal Compression: a techno-economic analysis

Energy Technology Data Exchange (ETDEWEB)

Kriha, Kenneth [Gas Technology Inst., Des Plaines, IL (United States); Petitpas, Guillaume [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Melchionda, Michael [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Soto, Herie [Shell, Houston TX (United States); Feng, Zhili [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Yanli [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-11

The goal of this project was to demonstrate the technical and economic feasibility of using thermal compression to create the hydrogen pressure necessary to operate vehicle hydrogen fueling stations. The concept of utilizing the exergy within liquid hydrogen to build pressure rather than mechanical components such as compressors or cryogenic liquid pumps has several advantages. In theory, the compressor-less hydrogen station will have lower operating and maintenance costs because the compressors found in conventional stations require large amounts of electricity to run and are prone to mechanical breakdowns. The thermal compression station also utilizes some of the energy used to liquefy the hydrogen as work to build pressure, this is energy that in conventional stations is lost as heat to the environment.

Impacts and societal benefits of research activities at Summit Station, Greenland

Science.gov (United States)

Hawley, R. L.; Burkhart, J. F.; Courville, Z.; Dibb, J. E.; Koenig, L.; Vaughn, B. H.

2017-12-01

Summit Station began as the site for the Greenland Ice Sheet Project 2 ice core in 1989. Since then, it has hosted both summer campaign science, and since 1997, year-round observations of atmospheric and cryospheric processes. The station has been continuously occupied since 2003. While most of the science activities at the station are supported by the US NSF Office of Polar Programs, the station also hosts many interagency and international investigations in physical glaciology, atmospheric chemistry, satellite validation, astrophysics and other disciplines. Summit is the only high elevation observatory north of the Arctic circle that can provide clean air or snow sites. The station is part of the INTER-ACT consortium of Arctic research stations with the main objective to identify, understand, predict and respond to diverse environmental changes, and part of the International Arctic Systems for Observing the Atmosphere (IASOA) that coordinates Arctic research activities and provides a networked, observations-based view of the Arctic. The Summit Station Science Summit, sponsored by NSF, assembled a multidisciplinary group of scientists to review Summit Station science, define the leading research questions for Summit, and make community-based recommendations for future science goals and governance for Summit. The impact of several on-going observation records was summarized in the report "Sustaining the Science Impact of Summit Station, Greenland," including the use of station data in weather forecasts and climate models. Observations made at the station as part of long-term, year-round research or during shorter summer-only campaign seasons contribute to several of the identified Social Benefit Areas (SBAs) outlined in the International Arctic Observations Assessment Framework published by the IDA Science and Technology Policy Institute and Sustaining Arctic Observing Networks as an outcome of the 2016 Arctic Science Ministerial. The SBAs supported by research

Major power system failures (Europe, USA) and deregulation of electricity markets

International Nuclear Information System (INIS)

Merlin, A.

2004-01-01

The year 2003 was marked by several electric power blackouts which affected major power systems in the world, and gave rise to many opinions about possible links with developments brought by the opening of the electricity markets. At first the paper presents the outages experienced by North America or August 14 and by Switzerland and Italy on September 28, 2003. Feedback on previous electric power blackouts makes possible to find out already known lessons, then to bring out how the latest blackouts may reveal similarities or differences towards developments brought by the opening of the electricity markets occurred ire the last ten years. Finally, seen from the transmission system operator of a european country within the context of continued institutional developments, the paper points out significant items which must be considered to prevent possible further blackouts. (author)

Prehospital severity scoring at major rock concert events.

Science.gov (United States)

Erickson, T B; Koenigsberg, M; Bunney, E B; Schurgin, B; Levy, P; Willens, J; Tanner, L

1997-01-01

Rock and contemporary music concerts are popular, recurrent events requiring on-site medical staffing. To describe a novel severity score used to stratify the level of acuity of patients presenting to first-aid stations at these events. Retrospective review of charts generated at the first-aid stations of five major rock concerts within a 60,000 spectator capacity, outdoor, professional sports stadium. Participants included all concert patrons presenting to the stadium's first-aid stations as patients. Data were collected on patient demographics, history of drug or ethanol usage while at the concert event, first-aid station time, treatment rendered, diagnosis, and disposition. All patients evaluated were retrospectively assigned a "DRUG-ROCK" Injury Severity Score (DRISS) to stratify their level of acuity. Individual concert events and patient dispositions were compared statistically using chi-square, Fisher's exact, and the ANOVA Mean tests. Approximately 250,000 spectators attended the five concert events. First-aid stations evaluated 308 patients (utilization rate of 1.2 per 1,000 patrons). The most common diagnosis was minor trauma (130; 42%), followed in frequency by ethanol/illicit drug intoxication (98; 32%). The average time in the first-aid station was 23.5 +/- 22.5 minutes (+/- standard deviation; range: 5-150 minutes). Disposition of patients included 100 (32.5%) who were treated and released; 98 (32%) were transported by paramedics to emergency departments (EDs); and 110 (35.5%) signed-out against medical advise (AMA), refusing transport. The mean DRISS was 4.1 (+/- 2.65). Two-thirds (67%) of the study population were ranked as mild by DRISS criteria (score = 1-4), with 27% rated as moderate (score = 5-9), and 6% severe (score > 10). The average of severity scores was highest (6.5) for patients transported to hospitals, and statistically different from the scores of the average of the treated and released and AMA groups (p rock concerts.

Atmosphere composition monitor for space station and advanced missions application

International Nuclear Information System (INIS)

Wynveen, R.A.; Powell, F.T.

1987-01-01

Long-term human occupation of extraterrestrial locations may soon become a reality. The National Aeronautics and Space Administration (NASA) has recently completed the definition and preliminary design of the low earth orbit (LEO) space station. They are now currently moving into the detailed design and fabrication phase of this space station and are also beginning to analyze the requirements of several future missions that have been identified. These missions include, for example, Lunar and Mars sorties, outposts, bases, and settlements. A requirement of both the LEO space station and future missions are environmental control and life support systems (ECLSS), which provide a comfortable environment for humans to live and work. The ECLSS consists of several major systems, including atmosphere revitalization system (ARS), atmosphere pressure and composition control system, temperature and humidity control system, water reclamation system, and waste management system. Each of these major systems is broken down into subsystems, assemblies, units, and instruments. Many requirements and design drivers are different for the ECLSS of the LEO space station and the identified advanced missions (e.g., longer mission duration). This paper discusses one of the ARS assemblies, the atmosphere composition monitor assembly (ACMA), being developed for the LEO space station and addresses differences that will exist for the ACMA of future missions

Space station interior design: Results of the NASA/AIA space station interior national design competition

Science.gov (United States)

Haines, R. F.

1975-01-01

The results of the NASA/AIA space station interior national design competition held during 1971 are presented in order to make available to those who work in the architectural, engineering, and interior design fields the results of this design activity in which the interiors of several space shuttle size modules were designed for optimal habitability. Each design entry also includes a final configuration of all modules into a complete space station. A brief history of the competition is presented with the competition guidelines and constraints. The first place award entry is presented in detail, and specific features from other selected designs are discussed. This is followed by a discussion of how some of these design features might be applied to terrestrial as well as space situations.

Using MAAP 4.0 to determine risks from steam generator tube leaks or ruptures

International Nuclear Information System (INIS)

Fuller, E.L.; Kenton, M.A.

1996-01-01

As part of the Electric Power Research Institute (EPRI) program on steam generator degradation specific management (SGDSM), the nuclear industry is investigating the effects on plant risk of severe accidents involving steam generator tube leaks or ruptures. Such accidents fall into three classes: those caused by spontaneous, steam generator tube ruptures (SGTRs) that subsequently result in core damage; those caused by design-basis accidents that lead to induced tube ruptures and subsequent core damage; and those that progress to core damage, such as a station blackout (SBO), with subsequent induced tube leakage or rupture. In each case, the potential exists for a significant fraction of the fission products released from a damaged core to reach the environment through the leaking or ruptured tubes

Radiation Hardened Electronics Destined For Severe Nuclear Reactor Environments

Energy Technology Data Exchange (ETDEWEB)

Holbert, Keith E. [Arizona State Univ., Tempe, AZ (United States); Clark, Lawrence T. [Arizona State Univ., Tempe, AZ (United States)

2016-02-19

Post nuclear accident conditions represent a harsh environment for electronics. The full station blackout experience at Fukushima shows the necessity for emergency sensing capabilities in a radiation-enhanced environment. This NEET (Nuclear Energy Enabling Technologies) research project developed radiation hardened by design (RHBD) electronics using commercially available technology that employs commercial off-the-shelf (COTS) devices and present generation circuit fabrication techniques to improve the total ionizing dose (TID) hardness of electronics. Such technology not only has applicability to severe accident conditions but also to facilities throughout the nuclear fuel cycle in which radiation tolerance is required. For example, with TID tolerance to megarads of dose, electronics could be deployed for long-term monitoring, inspection and decontamination missions. The present work has taken a two-pronged approach, specifically, development of both board and application-specific integrated circuit (ASIC) level RHBD techniques. The former path has focused on TID testing of representative microcontroller ICs with embedded flash (eFlash) memory, as well as standalone flash devices that utilize the same fabrication technologies. The standalone flash devices are less complicated, allowing better understanding of the TID response of the crucial circuits. Our TID experiments utilize biased components that are in-situ tested, and in full operation during irradiation. A potential pitfall in the qualification of memory circuits is the lack of rigorous testing of the possible memory states. For this reason, we employ test patterns that include all ones, all zeros, a checkerboard of zeros and ones, an inverse checkerboard, and random data. With experimental evidence of improved radiation response for unbiased versus biased conditions, a demonstration-level board using the COTS devices was constructed. Through a combination of redundancy and power gating, the demonstration

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Safety Distances for hydrogen filling stations

Energy Technology Data Exchange (ETDEWEB)

Matthijsen, A. J. C. M.; Kooi, E. S.

2005-07-01

In the Netherlands there is a growing interest in using natural gas as a transport fuel. The most important drivers behind this development are formed by poor inner city air quality and the decision to close several LPG filling stations. Dwellings are not allowed within the safety distances of 45 or 110 meters from the tanker filling point of these LPG stations, depending on the capacity of the station. Another driver is global warming. We are carrying out a study on station supply, compression, storage and filling for natural gas stations, and a similar, simultaneous study on hydrogen as a followup to our risk analysis for the hydrogen filling station in Amsterdam. Here, three buses drive on hydrogen as part of the European CUTE project. Driving on natural gas is an important step in the transition to cars on hydrogen. This study was commissioned by the Dutch Ministry of Spatial Planning, Housing and the Environment to advise on external safety aspects of future hydrogen filling stations. According to Dutch law homes may not be built within an individual risk contour of 10-6 per year of a dangerous object, such as a plant with hazardous materials or a filling station. An individual risk contour of 10-6 is represented by a line around a dangerous object that connects locations with an individual risk level of 10-6 per year. An individual 'located' within this contour line has a chance of one per million per year or more to be killed as a result of an accident caused by this object. The longest distance between the object and such a contour is called a 'safety distance'. A study on safety distances is now in progress for different kinds of hydrogen filling stations (e. g. gaseous and liquid hydrogen) and for different capacities, such as big, medium and small stations. The focus is on different kinds of hydrogen production and the hydrogen supply of the filling station. To decide on the design and supply of the hydrogen station, we examined the

A molecular epidemiological study of methicillin-resistant Staphylococci environmental contamination in railway stations and coach stations in Guangzhou of China.

Science.gov (United States)

Lin, J L; Peng, Y; Ou, Q T; Lin, D X; Li, Y; Ye, X H; Zhou, J L; Yao, Z J

2017-02-01

Methicillin-resistant Staphylococcus aureus (MRSA) has caused a series of public health problems since it was first found in 1961. However, there are few research studies on the MRSA environmental contamination in railway stations and coach stations. Therefore, the aim of this study was to determine MRSA environmental contamination in public transport stations. Between December 2013 and January 2014, 380 surface samples from three railway stations (180) and four coach stations (200) in Guangzhou were collected to isolate and determine the prevalence and characteristics of Staphylococci strains. 39·21% of all samples were Staphylococci isolates, 1·58% of Staphylococci isolates were MRSA isolates, and 6·05% were methicillin-susceptible S. aureus. The proportion of multidrug resistant among 149 Staphylococci isolates was 75·84%. None of MRSA isolates was identified with the Panton-Valentine Leukocidin (PVL) genes, and one of them was identified with the qac gene. Four MRSA isolates were Staphylococcal Cassette Chromosome mec IVa, and the other two were nontypeable. Staphylococcus aureus isolates were classified into several sequence types (STs), and STs showed possible cross-transmissions of isolates from various sources. Methicillin-resistant Staphylococci contamination prevalence was high, and the environment of stations may be the vectors transmitting the Staphylococci to passengers. This is the first study to comprehensively report the prevalence, antibiotic resistance, and molecular characteristics of contamination of Staphylococci isolates in railway stations and coach stations of China. It will have great public health implications on infection control in community settings because of the serious hazard of Staphylococci, especially methicillin-resistant Staphylococci. Our findings have provided evidence for relevant departments to reduce the contamination of Staphylococci in environment of public transport stations. © 2016 The Society for Applied

Measurements of MIMO Indoor Channels at 1800 MHz with Multiple Indoor and Outdoor Base Stations

Directory of Open Access Journals (Sweden)

Jaldén Niklas

2007-01-01

Full Text Available This paper proposes several configurations for multiple base stations in indoor MIMO systems and compares their performance. The results are based on channel measurements realized with a MIMO testbed. The receiver was moved along several routes and floors on an office building. Both outdoor and indoor locations are considered for the transmitters or base stations, which allow the analysis of not only indoor but also outdoor-to-indoor environment. The use of 2 base stations with different system level combinations of the two is analyzed. We show that the configuration with base station selection provides almost as good performance as a full water-filling scheme when the 2 base stations are placed at different locations. Also the spatial correlation properties for the different configurations are analyzed and the importance of considering path loss when evaluating capacity is highlighted.

Study of Containment Vent Strategies During Severe Accident Progression for the CANDU6 Plant

Energy Technology Data Exchange (ETDEWEB)

Jin, Youngho; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

2015-05-15

In March, 2011, Fukushima daichi nuclear power plants experienced a long term station blackout. Severe core damage occurred and a large amount of radioactive materials are released outside of the plants. After this terrible accident Nuclear Safety and Security Commission (NSSC) enforced to increase nuclear safety for all operating plants in Korea. To increase plant safety, both hardware reinforcement and software improvement are encouraged. Hardware reinforcement includes the preparation of the external water injection paths to the RCS and the spent fuel pool, a filtered containment venting system (CFVS), and AC power generating truck. Software improvement includes the increase of the effectiveness of the severe accident management guidance (SAMG) and plant staff training. To comply with NSSC's request, Wolsong Unit 1 has fulfilled the hardware reinforcement including the installation of a CFVS and started the extension of a SAMG to the low power and shutdown operation mode. Current SAMG deals accident occurred during full power operation only. The CFVS is designed to open and to close isolation valves manually. It does not require AC power. The operation of the CFVS prevents the reactor containment building failure due to the over-pressurization but it may release radioactive materials out of the reactor containment building. This paper discusses the radiological source terms for the containment vent strategy during severe accident progression which occurred during shutdown operation mode. This work is a part of the development of shutdown SAMG.. The CFVS is an effective means to control the containment pressure when the local air coolers are unavailable. Radioactive materials may release through the CFVS, but their amounts are reduced significantly. The alternative means, i.e., containment vent through the ventilation system which does not have an effective filter, is not a good choice to control the containment condition. It can maintain the containment

Verification of SAMG entry condition for APR1400

Energy Technology Data Exchange (ETDEWEB)

Yun, Junghyun; Kim, Taewan; Kim, Jonghyun [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2013-10-15

In the wake of the Fukushima accident, severe accident management has become important more than ever. While Emergency Operating Procedures (EOPs) focuses on cooling down the core during a design basis accident, the prime objective of severe accident management guideline (SAMG) is to prevent the release of radioactive material into the environment during a severe accident. Only one fixed value of core exit temperature (CET) has been applied for several decades to different types of nuclear power plants (NPPs) in Korea. Although the different types of NPPs have different cladding materials and system designs, the identical CET, i. e., 650 .deg. C, is applied as the entry condition to the SAMG of all the NPPs (except CANDU plants), i. e., Westinghouse type, OPR1000, and APR1400. In this paper, the transition point is re-evaluated for dominant severe accident sequences; station black-out (SBO), small break loss of coolant accident (SBLOCA), and me dium break loss of coolant accident (MBLOCA). In cases of SBO and SBLOCA, the current SAMG entry condition seems proper, while it needs to be reconsidered for MBLOCA case.

Performance assessment of Point Lepreau Generating Station

International Nuclear Information System (INIS)

Alikhan, S.

1991-01-01

The Point Lepreau Generating Station, a 680 MWe CANDU unit, is located about 40 km southwest of the city of Saint John, New Brunswick, Canada. It was declared in-service on 1 February, 1983 and, since then, has demonstrated an average cross capacity factor of over 93% up to the end of 1990. This paper compared the performance of the station with other sister CANDU units and the Light Water Reactors world-wide using the following ten performance indicators, as applicable: - gross capacity factor; - fuel burn-up; - heavy water upkeep; - unplanned reactor trips while critical; - forced outage rate; - fuel handling performance; - derived emission of radioactive effluents to environment; - personnel radiation dose; - industrial safety; - low-level solid radioactive wastes. The paper examines various areas of station activities including management and organization, operations and maintenance, technical support, fuel handling and health physics in order to highlight some of the 'good practices' which are believed to have made a significant contribution towards achieving the demonstrated performance of Point Lepreau G.S. In addition, several areas of potential improvement are discussed in order to maintain and enhance, where practicable, the safety, reliability and economic performance of the station. In this context, a careful review of the operating experiences, both in-house and at other stations, and a judicious application of lessons learned plays a significant role. (author)

Performance assessment of Point Lepreau Generating Station

Energy Technology Data Exchange (ETDEWEB)

Alikhan, S [Point Lepreau Generating Station, Lepreau, NB (Canada)

1991-04-01

The Point Lepreau Generating Station, a 680 MWe CANDU unit, is located about 40 km southwest of the city of Saint John, New Brunswick, Canada. It was declared in-service on 1 February, 1983 and, since then, has demonstrated an average cross capacity factor of over 93% up to the end of 1990. This paper compared the performance of the station with other sister CANDU units and the Light Water Reactors world-wide using the following ten performance indicators, as applicable: - gross capacity factor; - fuel burn-up; - heavy water upkeep; - unplanned reactor trips while critical; - forced outage rate; - fuel handling performance; - derived emission of radioactive effluents to environment; - personnel radiation dose; - industrial safety; - low-level solid radioactive wastes. The paper examines various areas of station activities including management and organization, operations and maintenance, technical support, fuel handling and health physics in order to highlight some of the 'good practices' which are believed to have made a significant contribution towards achieving the demonstrated performance of Point Lepreau G.S. In addition, several areas of potential improvement are discussed in order to maintain and enhance, where practicable, the safety, reliability and economic performance of the station. In this context, a careful review of the operating experiences, both in-house and at other stations, and a judicious application of lessons learned plays a significant role. (author)

Technical Analysis of the Hydrogen Energy Station Concept, Phase I and Phase II

Energy Technology Data Exchange (ETDEWEB)

TIAX, LLC

2005-05-04

Phase I Due to the growing interest in establishing a domestic hydrogen infrastructure, several hydrogen fueling stations already have been established around the country as demonstration units. While these stations help build familiarity with hydrogen fuel in their respective communities, hydrogen vehicles are still several years from mass production. This limited number of hydrogen vehicles translates to a limited demand for hydrogen fuel, a significant hurdle for the near-term establishment of commercially viable hydrogen fueling stations. By incorporating a fuel cell and cogeneration system with a hydrogen fueling station, the resulting energy station can compensate for low hydrogen demand by providing both hydrogen dispensing and combined heat and power (CHP) generation. The electrical power generated by the energy station can be fed back into the power grid or a nearby facility, which in turn helps offset station costs. Hydrogen production capacity not used by vehicles can be used to support building heat and power loads. In this way, an energy station can experience greater station utility while more rapidly recovering capital costs, providing an increased market potential relative to a hydrogen fueling station. At an energy station, hydrogen is generated on-site. Part of the hydrogen is used for vehicle refueling and part of the hydrogen is consumed by a fuel cell. As the fuel cell generates electricity and sends it to the power grid, excess heat is reclaimed through a cogeneration system for use in a nearby facility. Both the electrical generation and heat reclamation serve to offset the cost of purchasing the equivalent amount of energy for nearby facilities and the energy station itself. This two-phase project assessed the costs and feasibility of developing a hydrogen vehicle fueling station in conjunction with electricity and cogenerative heat generation for nearby Federal buildings. In order to determine which system configurations and operational

Performance diagnostic system for emergency diesel generators

International Nuclear Information System (INIS)

Logan, K.P.

1991-01-01

Diesel generators are commonly used for emergency backup power at nuclear stations. Emergency diesel generators (EDGs) are subject to both start-up and operating failures, due to infrequent and fast-start use. EDG reliability can be critical to plant safety, particularly when station blackout occurs. This paper describes an expert diagnostic system designed to consistently evaluate the operating performance of diesel generators. The prototype system is comprised of a suite of sensor monitoring, cylinder combustion analyzing, and diagnostic workstation computers. On-demand assessments of generator and auxiliary equipment performance are provided along with color trend displays comparing measured performance to reference-normal conditions

Environmental management at the Grand Rapids Generating Station following the Unit No.1 headcover failure

International Nuclear Information System (INIS)

Windsor, D.C.

1993-01-01

Failure of the headcover of Unit 1 in the Grand Rapids generating station in March, 1992 caused the station to flood, releasing several thousand gallons of oil and removing the station from service for several weeks. Environmental considerations were a considerable part of the station restoration activities, reservoir and flow management programs and responses to public concerns arising from the accident. A major oil spill containment and cleanup program was undertaken, with station cleanup and debris disposal carried out in a manner acceptable to environmental authorities. Reservoir spillage was necessitated by the station shutdown. The spill recreated fish habitat in the spillway and walleye spawning were documented. A compensation program was developed to respond to problems caused by debris flushed from the spillway channel. On spill termination, a fish salvage program removed fish from a scour hole in the spillway channel. A proactive program of public information provided local residents with the facts about the incident and response program, and allayed concerns about public safety. 4 refs., 2 figs

Non-Coop Station History

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Station history documentation for stations outside the US Cooperative Observer network. Primarily National Weather Service stations assigned WBAN station IDs. Other...

Service Station Attendant. Performance Objectives. Basic Course.

Science.gov (United States)

Davis, John

Several intermediate performance objectives and corresponding criterion measures are listed for each of 24 terminal objectives for a basic secondary level service station attendant course. The materials were developed for a two-semester course (2 and 3 hours daily). The specialized classroom and shop experiences are designed to enable the student…

47 CFR 80.107 - Service of private coast stations and marine-utility stations.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 5 2010-10-01 2010-10-01 false Service of private coast stations and marine...) SAFETY AND SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Operating Requirements and Procedures Operating Procedures-Land Stations § 80.107 Service of private coast stations and marine-utility stations. A...

Comparison of ASTEC 1.3 and ASTEC 1.3 R2 calculations in case of SBO for VVER-1000 reactor

International Nuclear Information System (INIS)

Atanasova, B.; Stefanova, A.; Grudev, P.

2009-01-01

The report presents the results from severe accident analyses performed with the both versions of ASTEC v1.3 and ASTEC v1.3R2 computer code for a VVER 1000 type of reactor. The purpose of this analysis is to assess the progress of ASTEC code modeling of main phenomena arising during hypothetical severe accidents. The final target of these analyses is to estimate the behaviour of the ASTEC code, its capability for simulation of severe accidents, including safety systems and Severe Accident Management (SAM) procedures. The analyses have been performed assuming a station blackout with simultaneous loss of HPIS, LPIS (ECCSs), EFWS and spray system due to failure of DGs. Hydro accumulators are not available. In the calculation it is assumed opening and stuck-open of PRZ relief valves. It has been organized the Fission Products path through the SEMPELL valve. It should be said that this investigation was limited to the 'in-vessel' phase of the sequence; therefore the effect of sprays on containment atmosphere has not been studied. (authors)

Streamflow characteristics at hydrologic bench-mark stations

Science.gov (United States)

Lawrence, C.L.

1987-01-01

The Hydrologic Bench-Mark Network was established in the 1960's. Its objectives were to document the hydrologic characteristics of representative undeveloped watersheds nationwide and to provide a comparative base for studying the effects of man on the hydrologic environment. The network, which consists of 57 streamflow gaging stations and one lake-stage station in 39 States, is planned for permanent operation. This interim report describes streamflow characteristics at each bench-mark site and identifies time trends in annual streamflow that have occurred during the data-collection period. The streamflow characteristics presented for each streamflow station are (1) flood and low-flow frequencies, (2) flow duration, (3) annual mean flow, and (4) the serial correlation coefficient for annual mean discharge. In addition, Kendall's tau is computed as an indicator of time trend in annual discharges. The period of record for most stations was 13 to 17 years, although several stations had longer periods of record. The longest period was 65 years for Merced River near Yosemite, Calif. Records of flow at 6 of 57 streamflow sites in the network showed a statistically significant change in annual mean discharge over the period of record, based on computations of Kendall's tau. The values of Kendall's tau ranged from -0.533 to 0.648. An examination of climatological records showed that changes in precipitation were most likely the cause for the change in annual mean discharge.

Accident sequences simulated at the Juragua nuclear power plant

International Nuclear Information System (INIS)

Carbajo, J.J.

1998-01-01

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

Analysis of induced steam generator tube rupture using MAAP 4.0

International Nuclear Information System (INIS)

Kenton, M.; Epstein, M.; Henry, R.E.; Paik, C.; Fuller, E.

1996-01-01

The nuclear industry has initiated a program of Steam Generator Degradation Specific Management (SGDSM) to cope with the various types of corrosion that have been observed in pressurized water reactor (PWR) steam generators. In parallel, the U.S. Nuclear Regulatory Commission is promulgating revised rules on steam generator tube integrity. To support these efforts, the Electric Power Research Institute has sponsored calculations with the MAAP 4 code. The principal objective of these calculations is to estimate the peak temperatures experienced by the steam generator tubes during high-pressure severe accidents. These results are used to evaluate the potential for degraded tubes to leak or rupture. Attention was focused on station blackout (SBO) accidents with loss of turbine-driven auxiliary feedwater because these generally result in the greatest threat to the tubes

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

Zhang Yadong; Guo Yue; Wu Yuanyuan; Zou Yao

2015-01-01

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

47 CFR 73.6016 - Digital Class A TV station protection of TV broadcast stations.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 4 2010-10-01 2010-10-01 false Digital Class A TV station protection of TV... Class A TV station protection of TV broadcast stations. Digital Class A TV stations must protect authorized TV broadcast stations, applications for minor changes in authorized TV broadcast stations filed on...

Concurrent communication among multi-transceiver stations over shared media. Ph.D. Thesis

Science.gov (United States)

Birk, Yitzhak

1987-01-01

In order to increase a local-area network's throughput beyond a single bus data rate without using dedicated switching nodes, multiple buses and multitransceiver stations are required. The design space of single-hop interconnections is explored among such stations. Interconnections are presented whose throughput can grow quadratically with the number of transmitters and receivers per station. These are referred to as selective broadcast interconnections (SBIs). The performance of various SBIs are studied. A spread-spectrum channel can accommodate several current successful transmission, and a single-transceiver node can thus utilize only a small fraction of the channel capacity. In order to allocate the appropriate fraction of capacity to a busy node, it is proposed to equip it with several transmitters and receivers, thereby turning it into a supernode. Several architectures and operation policies for supernodes are suggested and compared. It is shown that a supernode can significantly outperform a collection of independent conventional nodes with the same total numbers of transmitters and receivers. Packet-radio networks with half-duplex nodes, as well as networks with full-duplex nodes, are considered.

Fire Stations

Data.gov (United States)

Department of Homeland Security — Fire Stations in the United States Any location where fire fighters are stationed or based out of, or where equipment that such personnel use in carrying out their...

Do cities deserve more railway stations? The choice of a departure railway station in a multiple-station region

NARCIS (Netherlands)

Givoni, M.; Rietveld, P.

2014-01-01

Promoting the use of rail is an important element in sustainable transport policy. One of the most important decisions to make in planning the railway network is on the number of stations to provide. Stations are the access points to rail services and while each additional station increases rail's

Study of risk reduction by improving operation of reactor core isolation cooling system

International Nuclear Information System (INIS)

Watanabe, Yamato; Tazai, Ayuko; Yamagishi, Shohei; Muramatsu, Ken; Muta, Hitoshi

2014-01-01

The Fukushima Daiichi nuclear power plant fell into a station blackout (SBO) due to the earthquake and tsunami in which most of the core cooling systems were disabled. In the units 2 and 3, water injection to the core was performed only by water injection system with turbine driven pumps. In particular, it is inferred from observed plant parameters that the reactor core isolation cooling system (RCIC) continued its operation much longer than it was originally expected (8 hours). Since the preparation of safety measures did not work, the reactor core damaged. With a view to reduce risk of station blackout events in a BWR by accident management, this study investigated the efficacy of operation procedures that takes advantage of RCIC which can be operated with only equipment inside reactor building and does not require an AC power source. The efficacy was assessed in this study by two steps. The first step is a thermal hydraulic analysis with the RETRAN3D code to estimate the potential extension of duration of core cooling by RCIC and the second step is the estimation of time required for recovery of off-site power from experiences at nuclear power stations under the 3.11 earthquake. This study showed that it is possible to implement more reliable measures for accident termination and to greatly reduce the risk of SBO by the installation of accident management measures with use of RCIC for extension of core cooling under SBO conditions. (author)

47 CFR 73.6018 - Digital Class A TV station protection of DTV stations.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 4 2010-10-01 2010-10-01 false Digital Class A TV station protection of DTV... TV station protection of DTV stations. Digital Class A TV stations must protect the DTV service that... application for digital operation of an existing Class A TV station or to change the facilities of a digital...

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Acquisition of wood fuel at the Joseph C. McNeil Generating Station

Energy Technology Data Exchange (ETDEWEB)

Kropelin, W. [Burlington Electric Dept., VT (United States)

1993-12-31

The Joseph C. McNeil Generating Station is the world`s largest single boiler, municipally-owned, wood-fired electrical generating plant. The 50 megawatt McNeil Station is located in Burlington, Vermont and is owned by several Vermont public and private electric utilities. The operator and majority owner is the City of Burlington Electric Department (BED). Wood fuel procurement for the McNeil Station has been conducted in an environmentally sensitive way. Harvesting is carried out in conformance with a comprehensive wood chip harvesting policy and monitored by professional foresters. Unpredictable levels of Station operation require rigid adherence to a wood storage plan that minimizes the risk of over heating and spontaneous combustion of stockpiled fuel.

Space station orbit maintenance

Science.gov (United States)

Kaplan, D. I.; Jones, R. M.

1983-01-01

The orbit maintenance problem is examined for two low-earth-orbiting space station concepts - the large, manned Space Operations Center (SOC) and the smaller, unmanned Science and Applications Space Platform (SASP). Atmospheric drag forces are calculated, and circular orbit altitudes are selected to assure a 90 day decay period in the event of catastrophic propulsion system failure. Several thrusting strategies for orbit maintenance are discussed. Various chemical and electric propulsion systems for orbit maintenance are compared on the basis of propellant resupply requirements, power requirements, Shuttle launch costs, and technology readiness.

Some novel on-power refuelling features of CANDU stations

International Nuclear Information System (INIS)

Erwin, D.; Pendlebury, B.; Watson, J.F.; Welch, A.C.

1976-01-01

Part A of the paper describes the reasons for, and advantages resulting from, the use of flow assisted refuelling in the CANDU type nuclear reactors at the Pickering Generating Station. A separate fuel handling system is used for each reactor unit, as distinct from the system employed at the Bruce Generating station, where the fuel handling system is shared among several units. Part B of the paper describes some of the advantages of the shared concept with particular emphasis on the availability of the fuel handling system. (author)

Sizewell 'dirty tricks'

International Nuclear Information System (INIS)

Martin, Steve.

1987-01-01

The pro-nuclear lobby is accused of dubious tactics to promote their case for the Sizewell-B reactor. These include reassessing future electricity demand which, it is claimed, could only be met by nuclear power, claiming that new power stations were needed to avoid blackouts during cold spells, and the reporting of a major design fault in the control rods of the Torness and Heysham AGR stations. The latter is felt to be related to the promotion by the South of Scotland Electricity Board of the AGR case, as opposed to the Central Electricity Generating Board's advocation of a PWR reactor design. The author argues for a series of coal-fired power stations instead, and a major energy conservation programme. (UK)

Cumulative effects of cascade hydropower stations on total dissolved gas supersaturation.

Science.gov (United States)

Ma, Qian; Li, Ran; Feng, Jingjie; Lu, Jingying; Zhou, Qin

2018-03-01

Elevated levels of total dissolved gas (TDG) may occur downstream of dams during the spill process. These high levels would increase the incidence of gas bubble disease in fish and cause severe environmental impacts. With increasing numbers of cascade hydropower stations being built or planned, the cumulative effects of TDG supersaturation are becoming increasingly prominent. The TDG saturation distribution in the downstream reaches of the Jinsha River was studied to investigate the cumulative effects of TDG supersaturation resulting from the cascade hydropower stations. A comparison of the effects of the joint operation and the single operation of two hydropower stations (XLD and XJB) was performed to analyze the risk degree to fish posed by TDG supersaturation. The results showed that water with supersaturated TDG generated at the upstream cascade can be transported to the downstream power station, leading to cumulative TDG supersaturation effects. Compared with the single operation of XJB, the joint operation of both stations produced a much higher TDG saturation downstream of XJB, especially during the non-flood discharge period. Moreover, the duration of high TDG saturation and the lengths of the lethal and sub-lethal areas were much higher in the joint operation scenario, posing a greater threat to fish and severely damaging the environment. This work provides a scientific basis for strategies to reduce TDG supersaturation to the permissible level and minimize the potential risk of supersaturated TDG.

A Demonstration Training Program for Potential School Dropouts. A Service Station Training School for Dropout-Prone Students.

Science.gov (United States)

Rull, Marvin H.; Moore, Richard O.

One phase of the curriculum demonstration program sponsored jointly by the Quincy Public Schools and Southern Illinois University is the Service Station Training School described within this report. The Service Station Training School was one of several sheltered work stations which were developed to provide preemployment experiences and training…

Securing Localization With Hidden and Mobile Base Stations

DEFF Research Database (Denmark)

Capkun, Srdjan; Srivastava, Mani; Cagalj, Mario

2006-01-01

localization based on hidden and mobile base stations. Our approach enables secure localization with a broad spectrum of localization techniques: ultrasonic or radio, based on received signal strength or signal time of flight. Through several examples we show how this approach can be used to secure nodecentric...

Assessment of impact of a severe accident at nuclear power plant of Angra dos Reis with release of radionuclides to the atmosphere; Avaliacao do impacto de um acidente severo na usina de Angra dos Reis com liberacao de radionuclideos para a atmosfera

Energy Technology Data Exchange (ETDEWEB)

Aguiar, Andre Silva de

2015-07-01

This study had as purpose the assess the impact of a severe accident, and also analyze the dispersion of {sup 131}I in the atmosphere, so that, through concentrating and inhaling dose of the plume, were possible to verify if the results are in accordance with the indicated data by the Plan of Emergency of the CNAAA regarding the Impact Zone and Control. This exercise was performed with the aid of an atmospheric model and a dispersion where to atmospheric modeling we used the data coupling WRF / CALMET and of dispersion, CALPUFF. The suggested accident consists of a Station Blackout at Nuclear Power of Angra (Unit 1), where through the total core involvement, will release 100% of the {sup 131}I to the atmosphere. The value of the total activity in the nucleus to this radionuclide is 7.44 x 1017 Bq, that is relative on the sixth day of burning. This activity will be released through the chimney at a rate in Bq/s in the scenario of 12, 24, 48 and 72 hours of release. Applying the model in the proposed scenario, it is verified that the plume has concentrations of the order of 1020 Bq/m³ and dose of about 108 Sv whose value is beyond of the presented by Eletronuclear in your current emergency plan. (author)

Enhancement of safety for reprocessing facilities

International Nuclear Information System (INIS)

2012-06-01

The adequacy of the safety measures for utility loss accidents in nuclear fuel reprocessing facilities which have been formulated by the nuclear enterprises is investigated in JNES which organizes an advanced committee to specifically study this problem. The results are reviewed in the present report including the case of such severe accidents as in Fukushima Daiichi Nuclear Power Plant. The report also represents a tentative proposal for examination standards of such unimaginable severe accidents as 'station blackout,' urgent safety measures necessary for reoperation of nuclear power plants and requested by nuclear and industrial safety agency, and pointing out and clarification of the potential weakness from the safety point of view, and collective and composite evaluation of safety of the relevant facilities. Furthermore, the definition of accident management is given as of controlled condition and the authorized way of thinking for the cases of plural events happening at the same time and the cases when risks exist radioactivity emits with explosion. (S. Ohno)

Analysis of regulatory requirement for beyond design basis events of SMART

International Nuclear Information System (INIS)

Kim, W. S.; Seol, K. W.

2000-01-01

To enhance the safety of SMART reactor, safety and regulatory requirements associated with beyond design basis events (beyond BDE), which were developed and applied to advanced light water reactor designs, were analyzed along with a design status of passive reactor. And, based on these requirements, their applicability on the SMART design was evaluated. In the design aspect, severe accident prevention and mitigation features, containment performance, and accident management were analyzed. The evaluation results show that the requirement related to beyond DBE such as ATWS, loss of residual heat removal during shutdown operation, station blackout, fire, inter-system LOCA, and well-known events from severe accident phenomena is applicable to the SMART design. However, comprehensive approach against beyond DBE is not yet provided in the SMART design, and then it is required to designate and analyze the beyond DBE-related features. This study is expected to contribute to efforts to improve plant safety and to establish regulatory requirements for safety review

Alcohol-related blackouts among college students: impact of low level of response to alcohol, ethnicity, sex, and environmental characteristics

Directory of Open Access Journals (Sweden)

Priscila D. Gonçalves

2017-08-01

Full Text Available Objective: To explore how a genetically-influenced characteristic (the level of response to alcohol [LR], ethnicity, and sex relate to environmental and attitudinal characteristics (peer drinking [PEER], drinking to cope [COPE], and alcohol expectancies [EXPECT] regarding future alcohol-related blackouts (ARBs. Methods: Structural equation models (SEMs were used to evaluate how baseline variables related to ARB patterns in 462 college students over 55 weeks. Data were extracted from a longitudinal study of heavy drinking and its consequences at a U.S. university. Results: In the SEM analysis, female sex and Asian ethnicity directly predicted future ARBs (beta weights 0.10 and -0.11, respectively, while all other variables had indirect impacts on ARBs through alcohol quantities (beta weights ~ 0.23 for European American ethnicity and low LR, 0.21 for cannabis use and COPE, and 0.44 for PEER. Alcohol quantities then related to ARBs with beta = 0.44. The SEM explained 23% of the variance. Conclusion: These data may be useful in identifying college students who are more likely to experience future ARBs over a 1-year period. They enhance our understanding of whether the relationships of predictors to ARBs are direct or mediated through baseline drinking patterns, information that may be useful in prevention strategies for ARBs.

Estimating Pedestrian flows at train stations using the Station Transfer Model

NARCIS (Netherlands)

Van den Heuvel, J.P.A.; Dekkers, K.; De Vos, S.

2012-01-01

Train stations play a vital role in the door to door travel experience of train passengers. From the passengers’ value of time perspective, the station is the weakest link in total time value of the journey. Within the station the transfer function – moving between the various transport modes and

Broadcasting Stations of the World; Part III. Frequency Modulation Broadcasting Stations.

Science.gov (United States)

Foreign Broadcast Information Service, Washington, DC.

This third part of "Broadcasting Stations of the World", which lists all reported radio broadcasting and television stations, with the exception of those in the United States which broadcast on domestic channels, covers frequency modulation broadcasting stations. It contains two sections: one indexed alphabetically by country and city, and the…

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

OpenAIRE

TANAKA, Shun-ichi

2012-01-01

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

Simulation of Martian EVA at the Mars Society Arctic Research Station

Science.gov (United States)

Pletser, V.; Zubrin, R.; Quinn, K.

The Mars Society has established a Mars Arctic Research Station (M.A.R.S.) on Devon Island, North of Canada, in the middle of the Haughton crater formed by the impact of a large meteorite several million years ago. The site was selected for its similarities with the surface of the Mars planet. During the Summer 2001, the MARS Flashline Research Station supported an extended international simulation campaign of human Mars exploration operations. Six rotations of six person crews spent up to ten days each at the MARS Flashline Research Station. International crews, of mixed gender and professional qualifications, conducted various tasks as a Martian crew would do and performed scientific experiments in several fields (Geophysics, Biology, Psychology). One of the goals of this simulation campaign was to assess the operational and technical feasibility of sustaining a crew in an autonomous habitat, conducting a field scientific research program. Operations were conducted as they would be during a Martian mission, including Extra-Vehicular Activities (EVA) with specially designed unpressurized suits. The second rotation crew conducted seven simulated EVAs for a total of 17 hours, including motorized EVAs with All Terrain Vehicles, to perform field scientific experiments in Biology and Geophysics. Some EVAs were highly successful. For some others, several problems were encountered related to hardware technical failures and to bad weather conditions. The paper will present the experiment programme conducted at the Mars Flashline Research Station, the problems encountered and the lessons learned from an EVA operational point of view. Suggestions to improve foreseen Martian EVA operations will be discussed.

New Seismic Monitoring Station at Mohawk Ridge, Valles Caldera

Energy Technology Data Exchange (ETDEWEB)

Roberts, Peter Morse [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-09-20

Two new broadband digital seismic stations were installed in the Valles Caldera in 2011 and 2012. The first is located on the summit of Cerros del Abrigo (station code CDAB) and the second is located on the flanks of San Antonio Mountain (station code SAMT). Seismic monitoring stations in the caldera serve multiple purposes. These stations augment and expand the current coverage of the Los Alamos Seismic Network (LASN), which is operated to support seismic and volcanic hazards studies for LANL and northern New Mexico (Figure 1). They also provide unique continuous seismic data within the caldera that can be used for scientific studies of the caldera’s substructure and detection of very small seismic signals that may indicate changes in the current and evolving state of remnant magma that is known to exist beneath the caldera. Since the installation of CDAB and SAMT, several very small earthquakes have already been detected near San Antonio Mountain just west of SAMT (Figure 2). These are the first events to be seen in that area. Caldera stations also improve the detection and epicenter determination quality for larger local earthquakes on the Pajarito Fault System east of the Preserve and the Nacimiento Uplift to the west. These larger earthquakes are a concern to LANL Seismic Hazards assessments and seismic monitoring of the Los Alamos region, including the VCNP, is a DOE requirement. Currently the next closest seismic stations to the caldera are on Pipeline Road (PPR) just west of Los Alamos, and Peralta Ridge (PER) south of the caldera. There is no station coverage near the resurgent dome, Redondo Peak, in the center of the caldera. Filling this “hole” is the highest priority for the next new LASN station. We propose to install this station in 2018 on Mohawk Ridge just east of Redondito, in the same area already occupied by other scientific installations, such as the MCON flux tower operated by UNM.

Guidelines for Learning Stations.

Science.gov (United States)

Fehrle, Carl C.; Schulz, Jolene

Guidelines for designing and planning learning stations for pupils at the elementary grade level include suggestions on how to develop a station that will be successful in meeting the learners' needs. Instructions for the use of tapes at a station and matching pupils with stations are given, as are guidelines on classroom arrangement and record…

Temperature Discontinuity Caused by Relocation of Meteorological Stations in Taiwan

Directory of Open Access Journals (Sweden)

Chih-wen Hung

2009-01-01

Full Text Available With global warming upon us, it has be come increasingly important to identify the extent of this warming trend and in doing so be able to rank mean temperature changes in particular seasons and years. This requires a need for homogeneous climate data, which do not reflect individual anomalies in instruments, station locations or local environments (urbanization. Ac curate homogeneous long-term meteorological data helps show how temperature variations have truly occurred in the climate. Many possible factors contribute to artificial abrupt changes or sharp discontinuities in long time series data, such as the impact of station relocation, changes in observational schedules and instrumentation. Homogeneity adjustments of in situ climate data are very important processes for preparing observational data to be used in further analysis and research. Users require a well-documented history of stations to make appropriate homogeneity adjustments because precise historical back ground records of stations can provide researchers with knowledge of when artificial discontinuity has occurred and its causes. With out such de tailed historical data for each meteorological station, abrupt changes are difficult to interpret. Unfortunately, no homogeneity adjustments for temperature records have been con ducted previously in Tai wan, and present available sources of the history of Taiwan's meteorological stations exhibit in consistencies. In this study, information pertaining to station history, especially relocation records, is pro vided. This information is essential for anal y sis of continuous time series data for temperature and climate warming studies. Temperature data from several stations is given in this study to show how artificial discontinuity occurs due to station relocation. Al though there is no homogeneous adjusted climate data provided in this preliminary work, the summarizing of information regarding station relocations should be of assistance

Review of DDH 280 citadel cleansing station development. Memorandum report

Energy Technology Data Exchange (ETDEWEB)

Dyck, W.R.; McAndless, J.M.; Ho, J.

1994-02-01

A mock-up of a DDH 280 class destroyer Aft Cleansing Station has been modified following several technical evaluations. Chemical and biological simulant trials have been carried out and in the conclusion of these trials it was stated that the integrity of the citadel was maintained within a chemical or biological environment. Procedures have been developed which allowed a single man, or a group of men, or a casualty to proceed through an Aft Cleansing Station with minimum effort, and in a coordinated fashion. Evaluations have now been completed where the cleansing station and the personnel passing through it were challenged with the biological simulant Bacillus globigii (BG), and the chemical intake simulant dimethyl sulfoxide (DMSO). It was concluded, that when the cleansing station was operated properly, and the new procedures were being followed, it was possible to process contaminated personnel through the citadel entry system, and maintain the integrity of the citadel.

Model of nuclear reactor type VVER-1000/V-320 built by computer code ATHLET-CD

International Nuclear Information System (INIS)

Georgiev, Yoto; Filipov, Kalin; Velev, Vladimir

2014-01-01

A model of nuclear reactor type VVER-1000 V-320 developed for computer code ATHLET-CD2.1A is presented. Validation of the has been made, in the analysis of the station blackout scenario with LOCA on fourth cold leg is shown. As the calculation has been completed, the results are checked through comparison with the results from the computer codes ATHLET-2.1A, ASTEC-2.1 and RELAP5mod3.2

Computer simulation of black out followed by multiple failures in PWR type nuclear power plants

International Nuclear Information System (INIS)

Silva Filho, E.

1989-01-01

The computer code RELAP 5/MOD 1 has been utilized to investigate the thermal-hydraulic behaviour of a standard 1300 MWe pressurized water reactor plant of the KWU design during a station blackout following a inadequate performance of the pressurizer and steam generator safety valves. During the simulation the reactor scram system the emergency coolant system of the primary loop and the emergency Feedwater system of the secondary loop are considered inactive. (author) [pt

Developmental state and perspectives of USSR power stations, espec. nuclear power stations

International Nuclear Information System (INIS)

1983-01-01

According to the resolutions of the 25th and 26th party congresses of the CPSU, the Soviet electric and thermal energy economy envisages as the mainstreams in development: Energy projects based on nuclear fuel, i.e. nuclear power stations (NPS), nuclear heat- and -power stations (NHPS) and nuclear heat stations (NHS); fuel-energy complexes: Ekibastuz, Kansk-Achinsk, West-Siberian complex (Tyumen); power stations utilizing non-conventional regenerative energy sources, i.e. solar, geothermal, MHD power stations. Further down, an overview is given on the developmental perspectives of nuclear-heat and nuclear-power economy and on the development of energy management based on fossil fuels. (orig./UA) [de

47 CFR 95.139 - Adding a small base station or a small control station.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 5 2010-10-01 2010-10-01 false Adding a small base station or a small control... base station or a small control station. (a) Except for a GMRS system licensed to a non-individual, one or more small base stations or a small control station may be added to a GMRS system at any point...

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

Directory of Open Access Journals (Sweden)

C. M. Allison

2012-01-01

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

Stage measurement at gaging stations

Science.gov (United States)

Sauer, Vernon B.; Turnipseed, D. Phil

2010-01-01

Stream and reservoir stage are critical parameters in the computation of stream discharge and reservoir volume, respectively. In addition, a record of stream stage is useful in the design of structures that may be affected by stream elevation, as well as for the planning for various uses of flood plains. This report describes equipment and methodology for the observation, sensing, and recording of stage in streams and reservoirs. Although the U.S. Geological Survey (USGS) still uses the traditional, basic stilling-well float system as a predominant gaging station, modern electronic stage sensors and water-level recorders are now commonly used. Bubble gages coupled with nonsubmersible pressure transducers eliminate the need for stilling wells. Submersible pressure transducers have become common in use for the measurement of stage in both rivers and lakes. Furthermore, noncontact methods, such as radar, acoustic, and laser methods of sensing water levels, are being developed and tested, and in the case of radar, are commonly used for the measurement of stage. This report describes commonly used gaging-station structures, as well as the design and operation of gaging stations. Almost all of the equipment and instruments described in this report will meet the accuracy standard set by the USGS Office of Surface Water (OSW) for the measurement of stage for most applications, which is ±0.01 foot (ft) or 0.2 percent of the effective stage. Several telemetry systems are used to transmit stage data from the gaging station to the office, although satellite telemetry has become the standard. These telemetry systems provide near real-time stage data, as well as other information that alerts the hydrographer to extreme or abnormal events, and instrument malfunctions.

[Exploring nurse, usage effectiveness of mobile nursing station].

Science.gov (United States)

Chang, Fang-Mei; Lee, Ting-Ting

2013-04-01

A mobile nursing station is an innovative cart that integrates a wireless network, information technology devices, and online charts. In addition to improving clinical work and workflow efficiencies, data is integrated among different information systems and hardware devices to promote patient safety. This study investigated the effectiveness of mobile nursing cart use. We compared different distributions of nursing activity working samples to evaluate the nursing information systems in terms of interface usability and usage outcomes. There were two parts of this study. Part one used work sampling to collect nursing activity data necessary to compare a unit that used a mobile nursing cart (mobile group, n = 18) with another that did not (traditional group, n = 14). Part two applied a nursing information system interface usability questionnaire to survey the mobile unit with nurses who had used a mobile nursing station (including those who had worked in this unit as floating nurses) (n = 30) in order to explore interface usability and effectiveness. We found that using the mobile nursing station information system increased time spent on direct patient care and decreased time spent on indirect patient care and documentation. Results further indicated that participants rated interface usability as high and evaluated usage effectiveness positively. Comments made in the open-ended question section raised several points of concern, including problems / inadequacies related to hardware devices, Internet speed, and printing. This study indicates that using mobile nursing station can improve nursing activity distributions and that nurses hold generally positive attitudes toward mobile nursing station interface usability and usage effectiveness. The authors thus encourage the continued implementation of mobile nursing stations and related studies to further enhance clinical nursing care.

Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

International Nuclear Information System (INIS)

Carvalho, F. de A.T. de.

1985-01-01

Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt

Developments of space station; Uchu station no kaihatsu

Energy Technology Data Exchange (ETDEWEB)

Hashimoto, H. [National Space Development Agency of Japan, Tokyo (Japan)

1996-03-05

This paper introduces the Japanese experiment module (JEM) in developing a space station. The JEM consists of systems of a pressurizing section, an exposure section, a pressurizing portion of a supply section, a manipulator and an exposure portion of the supply section. The pressurizing section circulates and controls air so that crews can perform experiments under pressurized environment. The exposure section is a part in which experiments are carried out under exposure environment. The supply section runs between a station and the ground, with required devices loaded on it. The manipulator performs attaching a payload for the exposure section and replaces experimental samples. The JEM undergoes a schedule of fabricating an engineering model, testing for a certification a prototype flight model, and putting the model on a flight. The pressurizing section, exposure section and manipulator are at the stage of system tests. Surveillance of the JEM and control of the experiments are carried out at the Tsukuba Space Center. The Center is composed of a space experiment building, a zero-gravity environment testing building, an astronaut training building, a space station operating building, and a space station testing building. 7 figs., 2 tabs.

AECL's participation in the commissioning of Point Lepreau generating station unit 1

International Nuclear Information System (INIS)

Chawla, S.; Singh, K.; Yerramilli, S.

1983-05-01

Support from Atomic Energy of Canada Ltd. (AECL) to Point Lepreau during the commissioning program has been in the form of: seconded staff for commissioning program management, preparation of commissioning procedures, and hands-on commissioning of several systems; analysis of test results; engineering service for problem solving and modifications; design engineering for changes and additions; procurement of urgently-needed parts and materials; technological advice; review of operational limits; interpretation of design manuals and assistance with and preparation of submissions to regulatory authorities; and development of equipment and procedures for inspection and repairs. This, together with AECL's experience in the commissioning of other 600 MWe stations, Douglas Point and Ontario Hydro stations, provides AECL with a wide range of expertise for providing operating station support services for CANDU stations

Improving Station Performance by Building Isolation Walls in the Tunnel

Science.gov (United States)

Jia, Yan; Horn, Nikolaus; Leohardt, Roman

2014-05-01

Conrad Observatory is situated far away from roads and industrial areas on the Trafelberg in Lower Austria. At the end of the seismic tunnel, the main seismic instrument of the Observatory with a station code CONA is located. This station is one of the most important seismic stations in the Austrian Seismic Network (network code OE). The seismic observatory consists of a 145m long gallery and an underground laboratory building with several working areas. About 25 meters away from the station CONA, six temporary seismic stations were implemented for research purposes. Two of them were installed with the same equipment as CONA, while the remaining four stations were set up with digitizers having lower noise and higher resolution (Q330HR) and sensors with the same type (STS-2). In order to prevent possible disturbances by air pressure and temperature fluctuation, three walls were built inside of the tunnel. The first wall is located ca 63 meters from the tunnel entrance, while a set of double walls with a distance of 1.5 meters is placed about 53 meters from the first isolation wall but between the station CONA and the six temporary stations. To assess impact of the isolation walls on noise reduction and detection performance, investigations are conducted in two steps. The first study is carried out by comparing the noise level and detection performance between the station CONA behind the double walls and the stations in front of the double walls for verifying the noise isolation by the double walls. To evaluate the effect of the single wall, station noise level and detection performance were studied by comparing the results before and after the installation of the wall. Results and discussions will be presented. Additional experiment is conducted by filling insulation material inside of the aluminium boxes of the sensors (above and around the sensors). This should help us to determine an optimal insulation of the sensors with respect to pressure and temperature

The APS optics topography station

International Nuclear Information System (INIS)

Krasnicki, S.

1996-01-01

An in-house station for topographic testing of x-ray optical elements for the Advanced Photon Source experimental beamlines was set up by the Experimental Facilities Division of Argonne National Laboratory. A new double-crystal x-ray diffractometer was designed and built keeping in mind the need for testing large crystals possibly attached to cooling manifolds and lines. A short description of the new facility is given. The instrument performance fully satisfies imposed requirements, and the machine was successfully used for testing several silicon and diamond crystals

Newport Research Station

Data.gov (United States)

Federal Laboratory Consortium — The Newport Research Station is the Center's only ocean-port research facility. This station is located at Oregon State University's Hatfield Marine Science Center,...

Station Set Residual: Event Classification Using Historical Distribution of Observing Stations

Science.gov (United States)

Procopio, Mike; Lewis, Jennifer; Young, Chris

2010-05-01

Analysts working at the International Data Centre in support of treaty monitoring through the Comprehensive Nuclear-Test-Ban Treaty Organization spend a significant amount of time reviewing hypothesized seismic events produced by an automatic processing system. When reviewing these events to determine their legitimacy, analysts take a variety of approaches that rely heavily on training and past experience. One method used by analysts to gauge the validity of an event involves examining the set of stations involved in the detection of an event. In particular, leveraging past experience, an analyst can say that an event located in a certain part of the world is expected to be detected by Stations A, B, and C. Implicit in this statement is that such an event would usually not be detected by Stations X, Y, or Z. For some well understood parts of the world, the absence of one or more "expected" stations—or the presence of one or more "unexpected" stations—is correlated with a hypothesized event's legitimacy and to its survival to the event bulletin. The primary objective of this research is to formalize and quantify the difference between the observed set of stations detecting some hypothesized event, versus the expected set of stations historically associated with detecting similar nearby events close in magnitude. This Station Set Residual can be quantified in many ways, some of which are correlated with the analysts' determination of whether or not the event is valid. We propose that this Station Set Residual score can be used to screen out certain classes of "false" events produced by automatic processing with a high degree of confidence, reducing the analyst burden. Moreover, we propose that the visualization of the historically expected distribution of detecting stations can be immediately useful as an analyst aid during their review process.

Weather Radar Stations

Data.gov (United States)

Department of Homeland Security — These data represent Next-Generation Radar (NEXRAD) and Terminal Doppler Weather Radar (TDWR) weather radar stations within the US. The NEXRAD radar stations are...

Cosmic ray runs acquired with ATLAS muon stations

CERN Multimedia

Cerutti, F.

Starting in the fall 2005 several cosmic ray runs have been acquired in the ATLAS pit with six muon stations. These were three large outer and three large middle chambers of the feet sector (sector 13) that have been readout in the ATLAS cavern. In the first data taking period the trigger was based on two large scintillators (~300x30 cm2) positioned in sector 13 just below the large chambers. In this first run the precision chambers (the Monitored Drift Tubes) were operated in a close to final configuration. Typical trigger rates with this setup were of the order of 1 Hz. Several data sets of 10k events were acquired with final electronics up to the muon ROD and analysed with ATHENA-based software. These data allowed the first checks of the functionality and efficiency of the MDT stations in the ATLAS pit and the first measurement of the FE electronics noise in the ATLAS environment. A few event were also collected in a combined run with the TILE barrel calorimeter. An event display of a cosmic ray a...

Mission Analysis for LEO Microwave Power-Beaming Station in Orbital Launch of Microwave Lightcraft

Science.gov (United States)

Myrabo, L. N.; Dickenson, T.

2005-01-01

A detailed mission analysis study has been performed for a 1 km diameter, rechargeable satellite solar power station (SPS) designed to boost 20m diameter, 2400 kg Micr,oWave Lightcraft (MWLC) into low earth orbit (LEO) Positioned in a 476 km daily-repeating oi.bit, the 35 GHz microwave power station is configured like a spinning, thin-film bicycle wheel covered by 30% efficient sola cells on one side and billions of solid state microwave transmitter elements on the other, At the rim of this wheel are two superconducting magnets that can stor,e 2000 G.J of energy from the 320 MW, solar array over a period of several orbits. In preparation for launch, the entire station rotates to coarsely point at the Lightcraft, and then phases up using fine-pointing information sent from a beacon on-board the Lightcraft. Upon demand, the station transmits a 10 gigawatt microwave beam to lift the MWLC from the earth surface into LEO in a flight of several minutes duration. The mission analysis study was comprised of two parts: a) Power station assessment; and b) Analysis of MWLC dynamics during the ascent to orbit including the power-beaming relationships. The power station portion addressed eight critical issues: 1) Drag force vs. station orbital altitude; 2) Solar pressure force on the station; 3) Station orbital lifetime; 4) Feasibility of geo-magnetic re-boost; 5) Beta angle (i..e., sola1 alignment) and power station effective area relationship; 6) Power station percent time in sun vs, mission elapsed time; 7) Station beta angle vs.. charge time; 8) Stresses in station structures.. The launch dynamics portion examined four issues: 1) Ascent mission/trajecto1y profile; 2) MWLC/power-station mission geometry; 3) MWLC thrust angle vs. time; 4) Power station pitch rate during power beaming. Results indicate that approximately 0 58 N of drag force acts upon the station when rotated edge-on to project the minimum frontal area of 5000 sq m. An ion engine or perhaps an electrodynamic

Optimal Charging Schedule Planning and Economic Analysis for Electric Bus Charging Stations

Directory of Open Access Journals (Sweden)

Rong-Ceng Leou

2017-04-01

Full Text Available The battery capacity of electric buses (EB used for public transportation is greater than that of electric cars, and the charging power is also several times greater than that used in electric cars; this can result in high energy consumption and negatively impact power distribution networks. This paper proposes a framework to determine the optimal contracted power capacity and charging schedule of an EB charging station in such a way that energy costs can be reduced. A mathematical model of controlled charging, which includes the capacity and energy charges of the station, was developed to minimize costs. The constraints of the model include the charging characteristics of an EB and the operational guidelines of the bus company. A practical EB charging station was used to verify the proposed model. The financial viability of this EB charging station is also studied in this paper. The economic analysis model for this charging station considers investment and operational costs, and the operational revenue. Sensitivity analyses with respect to some key parameters are also performed in this paper. Based on actual operational routes and EB charging schemes, test results indicate that the EB charging station investment is feasible, and the planning model proposed can be used to determine optimal station power capacity and minimize energy costs.

Streamflow Gaging Stations

Data.gov (United States)

Department of Homeland Security — This map layer shows selected streamflow gaging stations of the United States, Puerto Rico, and the U.S. Virgin Islands, in 2013. Gaging stations, or gages, measure...

Ocean Station Vessel

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Ocean Station Vessels (OSV) or Weather Ships captured atmospheric conditions while being stationed continuously in a single location. While While most of the...

Essence and characteristics of the Westinghouse technology AP1000

International Nuclear Information System (INIS)

Llovet, Ricardo

2014-01-01

The AP1000 nuclear power plant can place the reactor in a Safe Shutdown Condition within the first 72 hours of a Station Blackout, without the use of AC power or operator action •With some operator action after 3 days, the AP1000 nuclear power plant continues to maintain reactor core cooling and Spent Fuel Pool cooling indefinitely •The AP1000 nuclear power plant has superior coping capabilities as well as significantly reduced risk for core damage

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Reference Climatological Stations

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — The Reference Climatological Stations (RCS) network represents the first effort by NOAA to create and maintain a nationwide network of stations located only in areas...

Fire Stations - 2007

Data.gov (United States)

Kansas Data Access and Support Center — Fire Station Locations in Kansas Any location where fire fighters are stationed at or based out of, or where equipment that such personnel use in carrying out their...

Fire Stations - 2009

Data.gov (United States)

Kansas Data Access and Support Center — Fire Stations in Kansas Any location where fire fighters are stationed or based out of, or where equipment that such personnel use in carrying out their jobs is...

Contaminated concrete scabbling at the Shippingport station decommissioning project

International Nuclear Information System (INIS)

Bauer, R.G.

1989-01-01

The Shippingport atomic power station was the first commercial nuclear power plant in the United States, joining the Duquesne Light Company (DLC) grid in December 1957. The Shippingport station was shut down in October 1982 and defueled in preparation for dismantling. On September 6, 1984, the Shippingport Station Decommissioning Project (SSDP) office of the US Department of Energy (DOE) assumed responsibility for the site. At turnover, there were several areas in the plant where radioactive contamination was entrained in concrete surfaces. The removal of contaminated concrete at SSDP was an important part of the decontamination to meet site release criteria, which is a major consideration in the decommissioning of nuclear power reactors. The highlights of this activity include: (1) development and application of remote scabbling tools, which effectively removed the contaminated concrete surfaces, and (2) use of scabblers minimized the removal of noncontaminated concrete by removing shallow layers of the surface and contributed to waste control, since the waste form enabled good packaging efficiency

[STEM on Station Education

Science.gov (United States)

Lundebjerg, Kristen

2016-01-01

The STEM on Station team is part of Education which is part of the External Relations organization (ERO). ERO has traditional goals based around BHAG (Big Hairy Audacious Goal). The BHAG model is simplified to a saying: Everything we do stimulates actions by others to advance human space exploration. The STEM on Station education initiate is a project focused on bringing off the earth research and learning into classrooms. Educational resources such as lesson plans, activities to connect with the space station and STEM related contests are available and hosted by the STEM on Station team along with their partners such as Texas Instruments. These educational activities engage teachers and students in the current happenings aboard the international space station, inspiring the next generation of space explorers.

2. All-Polish Seminar: Monitoring of Radioactive Environment Contamination Using ASS-500 and PMS Stations

International Nuclear Information System (INIS)

1998-01-01

During last 10 years the modern system of radiological pollution monitoring has been developed in Poland. At the beginning it was composed of ASS-500 stations. At the end of 1995 five automatic PMS stations have been installed as the result of the collaboration between Polish and Danish governments. Several stations of both types were include later. The system is coordinated by Central Laboratory for Radiological Protection. The report reviews the meeting dedicated to problems and future developments of the system

Resonance sensitivity of hydropower and pumping stations

Energy Technology Data Exchange (ETDEWEB)

Popescu, M.; Halanay, A.

1984-09-01

Comparative analysis of resonance diagrams for several hydropower and pumping stations with surge tanks and air chambers shows large differences in the maximum resonance pressures. A strategy is advocated which consists of hydraulic resonance computations coupled with practical surveillance measures during the operation of resonance sensitive hydraulic systems. A fundamental hydraulic scheme is considered consisting of a reservoir, a pressure tunnel, a surge tank, a penstock and a turbine combined into a hydropower station. It is suggested that for each hydraulic surge system it is necessary to carry out special resonance analyses following the normal procedure to obtain the resonance sensitivity. For hydraulic systems which are resonance sensitive, mechanical electronic equipment should be used to measure non-stationary pressures of the water in the conduit as a way of continuous surveillance during functioning. 6 references, 6 figures.

Initiating Training Stations As Clusters of Learning in Fashion Merchandising

Science.gov (United States)

Garber, Jayne

1974-01-01

A Chicago business school offers fashion merchandising as one of several business curriculums that combines on-the-job training and classroom instruction. Instruction is organized around the occupational cluster concept which requires training stations that provide a wide variety of learning experiences. (EA)

Station History Of The Seismic Station In Ahmadu Bello University ...

African Journals Online (AJOL)

Dominants in the selected events are events from Meditterranian, East Kazakhstan, India/Burma/China, South and Central America and North Ascension island regions. The limited number of events reporting at the station was due to low operational gain at the station which permitted only events whose magnitudes are ...

The Trencin water power station

International Nuclear Information System (INIS)

2005-01-01

This leaflet describes the Trencin water power station. The Trencin water power station was built seven years after the Dubnica nad Vahom water power station started its operation and was the last stage of the first and the oldest derived cascade of water power stations on the Vah River. After completing water power stations at Ladce (1936), Ilava (1946) and Dubnica nad Vahom (1949) and before constructing the Trencin water power station, the whole second derived cascade of water power stations including water power stations at Kostolna, Nove Mesto nad Vahom and Horna Streda was built as soon as possible mainly because the need to get compensation for discontinued electricity supplies as well as energetic coal from the Czech Republic. Hereby, experiences from the construction of previous grades were used, mainly as far as the dimensioning was concerned, as the fi rst installed power stations had, in comparison with the growing requirements on the electricity supplies, very low absorption capacity - only 150 m 3 .s -1 . Thus the Trencin power station (original name was the Skalka power station) was already dimensioned for the same absorption capacity as the cascade located downstream the river, that is 180 m 3 .s -1 . That was related also to growing demands on electricity supplies during the peaks in the daily electric system load diagram, and thus to the transfer from continuous operation of the water power station to semi-peak or even peak performance. According to the standards of power station classification, the Trencin water power station is a medium size, low pressure, channel power station with two units equipped by Kaplan turbines and synchronous hydro-alternators. The water power station installed capacity is 16.1 MW in total and its designed annual production of electrical energy for medium water year is 85,000 MWh, while the average annual production during the last 30 years is 86,252 MWh. Installed unit has a four-blade Kaplan turbine with the diameter

Tether applications for space station

Science.gov (United States)

Nobles, W.

1986-01-01

A wide variety of space station applications for tethers were reviewed. Many will affect the operation of the station itself while others are in the category of research or scientific platforms. One of the most expensive aspects of operating the space station will be the continuing shuttle traffic to transport logistic supplies and payloads to the space station. If a means can be found to use tethers to improve the efficiency of that transportation operation, it will increase the operating efficiency of the system and reduce the overall cost of the space station. The concept studied consists of using a tether to lower the shuttle from the space station. This results in a transfer of angular momentum and energy from the orbiter to the space station. The consequences of this transfer is studied and how beneficial use can be made of it.

On demand processing of climate station sensor data

Science.gov (United States)

Wöllauer, Stephan; Forteva, Spaska; Nauss, Thomas

2015-04-01

Large sets of climate stations with several sensors produce big amounts of finegrained time series data. To gain value of this data, further processing and aggregation is needed. We present a flexible system to process the raw data on demand. Several aspects need to be considered to process the raw data in a way that scientists can use the processed data conveniently for their specific research interests. First of all, it is not feasible to pre-process the data in advance because of the great variety of ways it can be processed. Therefore, in this approach only the raw measurement data is archived in a database. When a scientist requires some time series, the system processes the required raw data according to the user-defined request. Based on the type of measurement sensor, some data validation is needed, because the climate station sensors may produce erroneous data. Currently, three validation methods are integrated in the on demand processing system and are optionally selectable. The most basic validation method checks if measurement values are within a predefined range of possible values. For example, it may be assumed that an air temperature sensor measures values within a range of -40 °C to +60 °C. Values outside of this range are considered as a measurement error by this validation method and consequently rejected. An other validation method checks for outliers in the stream of measurement values by defining a maximum change rate between subsequent measurement values. The third validation method compares measurement data to the average values of neighboring stations and rejects measurement values with a high variance. These quality checks are optional, because especially extreme climatic values may be valid but rejected by some quality check method. An other important task is the preparation of measurement data in terms of time. The observed stations measure values in intervals of minutes to hours. Often scientists need a coarser temporal resolution

Mobile environmental radiation monitoring station

International Nuclear Information System (INIS)

Assido, H.; Shemesh, Y.; Mazor, T.; Tal, N.; Barak, D.

1997-01-01

A mobile environmental radiation monitoring station has been developed and established for the Israeli Ministry of Environment. The radiation monitoring station is ready for immediate placing in any required location, or can be operated from a vehicle. The station collects data Tom the detector and transfers it via cellular communication network to a Computerized Control Center for data storage, processing, and display . The mobile station is fully controlled from the. Routinely, the mobile station responses to the data request accumulated since the last communication session. In case of fault or alarm condition in the mobile station, a local claim is activated and immediately initiates communication with the via cellular communication network. (authors)

Real-time trend monitoring of gas compressor stations

Energy Technology Data Exchange (ETDEWEB)

Van Hardeveld, T. (Nova, an Alberta Corp., AB (Canada))

1991-02-01

The authors' company has developed a machinery health monitoring system (MHealth) for short-term and long-term historical trending and analysis of data from its 40 gas compressor stations. The author discusses the benefits of real-time trending in troubleshooting operations, in preventative maintenance scheduling and cites specific applications in the startup operations of several new gas compressor/centrifugal compressor units.

CDIP Station Data Collection - All Stations

Data.gov (United States)

Scripps Institution of Oceanography, UC San Diego — The Coastal Data Information Program's station data collection consists of all publicly-released coastal environment measurements taken over the program's history, a...

Greening radio access networks using distributed base station architectures

DEFF Research Database (Denmark)

Kardaras, Georgios; Soler, José; Dittmann, Lars

2010-01-01

Several actions for developing environmentally friendly technologies have been taken in most industrial fields. Significant resources have also been devoted in mobile communications industry. Moving towards eco-friendly alternatives is primarily a social responsibility for network operators....... However besides this, increasing energy efficiency represents a key factor for reducing operating expenses and deploying cost effective mobile networks. This paper presents how distributed base station architectures can contribute in greening radio access networks. More specifically, the advantages...... energy saving. Different subsystems have to be coordinated real-time and intelligent network nodes supporting complicated functionalities are necessary. Distributed base station architectures are ideal for this purpose mainly because of their high degree of configurability and self...

Commercial opportunities utilizing the International Space Station

Science.gov (United States)

Kearney, Michael E.; Mongan, Phil; Overmyer, Carolyn M.; Jackson, Kenneth

1998-01-01

The International Space Station (ISS) has the unique capability of providing a low-g environment for both short- and long-duration experimentation. This environment can provide a unique and competitive research capability to industry; but until recently, utilization of this environment by the private sector has been limited if not totally unavailable. NASA has recently expressed an interest in the commercial development of space and this is now an integral part of the Agency's enabling legislation through the Space Act. NASA's objective is to foster the use of the space environment for the development of commercial products and processes. Through alliances and agreements with several commercial companies and universities, SPACEHAB, Inc., has built a comprehensive package of services designed to provide low-cost reliable access to space for experimenters. These services provide opportunities to support engineering test beds for materials exposure analysis, to mitigate structural failures as observed on the Hubble Space Telescope; materials processing, remote sensing; space environment definition; and electronic experiments. The intent of this paper is to identify commercial opportunities for utilizing the International Space Station and provide examples of several facilities currently being designed and manufactured by commercial companies with the purpose of providing access to the space environment for commercial users.

A customer-friendly Space Station

Science.gov (United States)

Pivirotto, D. S.

1984-01-01

This paper discusses the relationship of customers to the Space Station Program currently being defined by NASA. Emphasis is on definition of the Program such that the Space Station will be conducive to use by customers, that is by people who utilize the services provided by the Space Station and its associated platforms and vehicles. Potential types of customers are identified. Scenarios are developed for ways in which different types of customers can utilize the Space Station. Both management and technical issues involved in making the Station 'customer friendly' are discussed.

BWR/5 Pressure-Suppression Pool Response during an SBO

Directory of Open Access Journals (Sweden)

Javier Ortiz-Villafuerte

2013-01-01

Full Text Available RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liquid water inventory in the PSP is vaporized. During the SBO, without RCIC, this inventory is reduced about 5% more within six hours. Further, a significant pressure rise occurs in containment at about the time when a sharp increase of heat generation occurs in RPV due to cladding oxidation. Failure temperature of fuel clad is also reached at this point. As the accident progresses, conditions for containment venting can be reached in about nine hours, although there still exists considerable margin before reaching containment design pressure. Detailed information of accident progress in reactor vessel and containment is presented and discussed.

The Miksova water power station

International Nuclear Information System (INIS)

2005-01-01

This leaflet describes the Miksova water power station. The Miksova water power station is part of the second derived cascade of hydro power stations on the river Vah. It was built at the end of a huge development in Slovak hydro-energy in the late 1950's and the beginning of the 1960's. It is the second water power station on this derived cascade, which is situated downstream the Hricov reservoir and water power station. At the power station, three turbine sets with vertical Kaplan turbines are installed with a total power output of 3 x 31.2 = 93.6 MW. With this power output the Miksova water power station (Miksova I) was the biggest water power station in the Slovak Republic until the construction of Pumping water power station Liptovska Mara. And it is still the biggest channel water power station on the Vah so far. It was put into operation during the period 1963 to 1965. There are three turbine sets with Kaplan turbines from CKD Blansko, with a synchronous hydro-alternator installed in the power station. Their installed capacity is 93.6 MW in total and the projected annual production of electrical energy is 207 GWh. The turbines are fi ve-bladed (on the Hricov and Povazska Bystrica water power stations they are four-bladed) and the impeller wheel has a diameter of 4800 mm. They are designed for extension of the head from 24.1 to 22.21 m and each of them has an absorption capacity of 134 m 3 .s -1 nd a nominal operating speed of 2.08 m 3 .s -1 , runaway speed 4.9 m 3 .s -1 . Each synchronous hydro-alternator has a maximum power output of 31.2 MW, a nominal voltage of 10.5 kV and power factor cos φ of 0.8. Power from the power station is led out through 110 kV switchgear. The water power station operates under automatic turbine mode of operation with remote indication and control from the Dispatch Centre at Vodne elektrarne, in Trencin. From start of operation until the end of 2003 all three turbine sets operated for a total of 450,500 running hours and the

Space station propulsion requirements study

Science.gov (United States)

Wilkinson, C. L.; Brennan, S. M.

1985-01-01

Propulsion system requirements to support Low Earth Orbit (LEO) manned space station development and evolution over a wide range of potential capabilities and for a variety of STS servicing and space station operating strategies are described. The term space station and the overall space station configuration refers, for the purpose of this report, to a group of potential LEO spacecraft that support the overall space station mission. The group consisted of the central space station at 28.5 deg or 90 deg inclinations, unmanned free-flying spacecraft that are both tethered and untethered, a short-range servicing vehicle, and a longer range servicing vehicle capable of GEO payload transfer. The time phasing for preferred propulsion technology approaches is also investigated, as well as the high-leverage, state-of-the-art advancements needed, and the qualitative and quantitative benefits of these advancements on STS/space station operations. The time frame of propulsion technologies applicable to this study is the early 1990's to approximately the year 2000.

Hydrogen Station Cost Estimates: Comparing Hydrogen Station Cost Calculator Results with other Recent Estimates

Energy Technology Data Exchange (ETDEWEB)

Melaina, M. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Penev, M. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2013-09-01

This report compares hydrogen station cost estimates conveyed by expert stakeholders through the Hydrogen Station Cost Calculation (HSCC) to a select number of other cost estimates. These other cost estimates include projections based upon cost models and costs associated with recently funded stations.

Principal facts for about 16,000 gravity stations in the Nevada Test Site and vicinity

International Nuclear Information System (INIS)

Harris, R.N.; Ponce, D.A.; Oliver, H.W.; Healey, D.L.

1989-01-01

The Nevada Test Site (NTS) and vicinity includes portions of the Goldfield, Caliente, Death Valley, and Las Vegas. This report documents and consolidates previously published and recently compiled gravity data to establish a gravity data base of about 16,000 stations for the NTS and vicinity. While compiling data sets, redundant stations and stations having doubtful locations or gravity values were excluded. Details of compiling the gravity data sets are discussed in later sections. Where feasible, an accuracy code has been assigned to each station so that the accuracy or reliability of each station can be evaluated. This data base was used in preparing complete Bouguer and isostatic gravity maps of the NTS and vicinity. Since publication of the complete Bouguer gravity map, additional data were incorporated into the isostatic gravity map. Gravity data were compiled from five sources: 14,183 stations from the US Geological Survey (USGS), 326 stations from Exploration Data Consultants (EDCON) of Denver, Colorado, 906 stations from the Los Alamos National Laboratory (LANL), 212 stations from the University of Texas at Dallas (UTD), and 48 stations from the Defense Mapping Agency (DMA). This investigation is an effort to study several areas for potential storage of high-level radioactive waste. Gravity stations established under YMP are shown. The objective of this gravity survey was to explore for the presence of plutons. This volume contains only compiled data

Principal facts for about 16,000 gravity stations in the Nevada Test Site and vicinity

International Nuclear Information System (INIS)

Harris, R.N.; Ponce, D.A.; Oliver, H.W.; Healey, D.L.

1989-01-01

The Nevada Test Site (NTS) and vicinity includes portions of the Goldfield, Caliente, Death Valley, and Las Vegas. This report documents and consolidates previously published and recently compiled gravity data to establish a gravity data base of about 16,000 stations for the NTS and vicinity. While compiling data sets, redundant stations and stations having doubtful locations or gravity values were excluded. Details of compiling the gravity data sets are discussed in later sections. Where feasible, an accuracy code has been assigned to each station so that the accuracy or reliability of each station can be evaluated. This data base was used in preparing complete Bouguer and isostatic gravity maps of the NTS and vicinity. Since publication of the complete Bouguer gravity map, additional data were incorporated into the isostatic gravity map. Gravity data were compiled from five sources: 14,183 stations from the US Geological Survey (USGS), 326 stations from Exploration Data Consultants (EDCON) of Denver, Colorado, 906 stations from the Los Alamos National Laboratory (LANL), 212 stations from the University of Texas at Dallas (UTD), and 48 stations from the Defense Mapping Agency (DMA). This investigation is an effort to study several areas for potential storage of high-level radioactive waste. Gravity stations established under YMP are shown. The objective of this gravity survey was to explore for the presence of plutons. 33 refs., 24 figs., 9 tabs

Gas Stations, US, 2010, NAVTEQ

Data.gov (United States)

U.S. Environmental Protection Agency — The Gas_Stations dataset is derived from the Navteq 'AUTOSVC' SDC layer (FAC_TYPE=5540) and contains gas stations and petrol stations. This NAVTEQ dataset is...

Non-Coop Station History (Unindexed)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Station history documentation for stations outside the US Cooperative Observer network. Documents should be compared with those in the Non-Coop Station History...

Single-Station Sigma for the Iranian Strong Motion Stations

Science.gov (United States)

Zafarani, H.; Soghrat, M. R.

2017-11-01

In development of ground motion prediction equations (GMPEs), the residuals are assumed to have a log-normal distribution with a zero mean and a standard deviation, designated as sigma. Sigma has significant effect on evaluation of seismic hazard for designing important infrastructures such as nuclear power plants and dams. Both aleatory and epistemic uncertainties are involved in the sigma parameter. However, ground-motion observations over long time periods are not available at specific sites and the GMPEs have been derived using observed data from multiple sites for a small number of well-recorded earthquakes. Therefore, sigma is dominantly related to the statistics of the spatial variability of ground motion instead of temporal variability at a single point (ergodic assumption). The main purpose of this study is to reduce the variability of the residuals so as to handle it as epistemic uncertainty. In this regard, it is tried to partially apply the non-ergodic assumption by removing repeatable site effects from total variability of six GMPEs driven from the local, Europe-Middle East and worldwide data. For this purpose, we used 1837 acceleration time histories from 374 shallow earthquakes with moment magnitudes ranging from M w 4.0 to 7.3 recorded at 370 stations with at least two recordings per station. According to estimated single-station sigma for the Iranian strong motion stations, the ratio of event-corrected single-station standard deviation ( Φ ss) to within-event standard deviation ( Φ) is about 0.75. In other words, removing the ergodic assumption on site response resulted in 25% reduction of the within-event standard deviation that reduced the total standard deviation by about 15%.

Central Station Design Options

DEFF Research Database (Denmark)

2011-01-01

. The work identifies the architecture, sizing and siting of prospective Central Stations in Denmark, which can be located at shopping centers, large car parking lots or gas stations. Central Stations are planned to be integrated in the Danish distribution grid. The Danish island of Bornholm, where a high...... overloading, more reference points might be necessary to represent various transformer loading levels. The subject of safety in Central Station is also addressed. A number of safety rules based on European standards apply to AC charging equipment up to 44 kW. The connection interlock and the automatic de......-energization are identified as fundamental requirements for safety in such a charging station. The connection interlock is a solution which ensures that no power is applied to the DC cable when the EV connector is not connected. The automatic de-energization device ensures that whenever a strain on the cable is detected, e...

Enhanced Master Station History Report

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — The Enhanced Master Station History Report (EMSHR) is a compiled list of basic, historical information for every station in the station history database, beginning...

SPS rectifier stations

CERN Multimedia

CERN PhotoLab

1974-01-01

The first of the twelves SPS rectifier stations for the bending magnets arrived at CERN at the end of the year. The photograph shows a station with the rectifiers on the left and in the other three cubicles the chokes, capacitors and resistor of the passive filter.

47 CFR 97.109 - Station control.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 5 2010-10-01 2010-10-01 false Station control. 97.109 Section 97.109... SERVICE Station Operation Standards § 97.109 Station control. (a) Each amateur station must have at least one control point. (b) When a station is being locally controlled, the control operator must be at the...

Independent measuring station for the GNEJS neutron spectrometer information and measurement system

International Nuclear Information System (INIS)

Gorokhov, I.S.; Laptev, A.B.; Marchenkov, V.V.; Tubol'tsev, Yu.V.; Fokin, E.Yu.; Shcherbakov, O.A.

1984-01-01

An independent measuring station (IMS) being a part of the information and measuring system of the neutron time-of-flight spectrometer is described. IMS represents the time (time-to-amplitude or time-to-angular) module analyzer assembled on the base of one or several CAMAC crates. The station permits to perform time measurements as well as the time measurements with amplitude weighing in an independent mode with subseqUent data transmission the central PDP-11/05 computer. The station processor unit is an independent controller. The software consists of the programming langUage of the controller from local terminal and programs ensuring measuring and shaping the spectra with their output on TV display devic. communication with the computer and dialogue with the operator

Environmental Control and Life Support Systems technology options for Space Station application

Science.gov (United States)

Hall, J. B., Jr.; Ferebee, M. J., Jr.; Sage, K. H.

1985-01-01

Continuous assessments regarding the suitability of candidate technologies for manned Space Stations will be needed over the next several years to obtain a basis for recommending the optimum system for an Initial Operating Capability (IOC) Space Station which is to be launched in the early 1990's. This paper has the objective to present analysis programs, the candidate recommendations, and the recommended approach for integration these candidates into the NASA Space Station reference configuration. Attention is given to ECLSS (Environmental Control and Life Support System) technology assessment program, an analysis approach for candidate technology recommendations, mission model variables, a candidate integration program, metabolic oxygen recovery, urine/flush water and all waste water recovery, wash water and condensate water recovery, and an integration analysis.

Power station instrumentation

International Nuclear Information System (INIS)

Jervis, M.W.

1993-01-01

Power stations are characterized by a wide variety of mechanical and electrical plant operating with structures, liquids and gases working at high pressures and temperatures and with large mass flows. The voltages and currents are also the highest that occur in most industries. In order to achieve maximum economy, the plant is operated with relatively small margins from conditions that can cause rapid plant damage, safety implications, and very high financial penalties. In common with other process industries, power stations depend heavily on control and instrumentation. These systems have become particularly significant, in the cost-conscious privatized environment, for providing the means to implement the automation implicit in maintaining safety standards, improving generation efficiency and reducing operating manpower costs. This book is for professional instrumentation engineers who need to known about their use in power stations and power station engineers requiring information about the principles and choice of instrumentation available. There are 8 chapters; chapter 4 on instrumentation for nuclear steam supply systems is indexed separately. (Author)

Station Capacity

DEFF Research Database (Denmark)

Landex, Alex

2011-01-01

the probability of conflicts and the minimum headway times into account. The last method analyzes how optimal platform tracks are used by examining the arrival and departure pattern of the trains. The developed methods can either be used separately to analyze specific characteristics of the capacity of a station......Stations are often limiting the capacity of railway networks. This is due to extra need of tracks when trains stand still, trains turning around, and conflicting train routes. Although stations are often the capacity bottlenecks, most capacity analysis methods focus on open line capacity. Therefore...... for platform tracks and the probability that arriving trains will not get a platform track immediately at arrival. The third method is a scalable method that analyzes the conflicts in the switch zone(s). In its simplest stage, the method just analyzes the track layout while the more advanced stages also take...

Secure base stations

NARCIS (Netherlands)

Bosch, Peter; Brusilovsky, Alec; McLellan, Rae; Mullender, Sape J.; Polakos, Paul

2009-01-01

With the introduction of the third generation (3G) Universal Mobile Telecommunications System (UMTS) base station router (BSR) and fourth generation (4G) base stations, such as the 3rd Generation Partnership Project (3GPP) Long Term Evolution (LTE) Evolved Node B (eNB), it has become important to

Data communications method for mobile network in fourth generation communications system, involves delivering decoded data to mobile station from relay station, where mobile station receives data from both relay and base stations

DEFF Research Database (Denmark)

2008-01-01

The method involves utilizing a base station (BS) (100) to transmit data to a relay station (RS) (110) and a mobile station (MS) (120), where the data includes two messages. The BS is utilized to transmit the two messages by utilizing a linear combination method, and the data is received in the RS...

Improved social force model based on exit selection for microscopic pedestrian simulation in subway station

Institute of Scientific and Technical Information of China (English)

郑勋; 李海鹰; 孟令云; 许心越; 陈旭

2015-01-01

An improved social force model based on exit selection is proposed to simulate pedestrians’ microscopic behaviors in subway station. The modification lies in considering three factors of spatial distance, occupant density and exit width. In addition, the problem of pedestrians selecting exit frequently is solved as follows: not changing to other exits in the affected area of one exit, using the probability of remaining preceding exit and invoking function of exit selection after several simulation steps. Pedestrians in subway station have some special characteristics, such as explicit destinations, different familiarities with subway station. Finally, Beijing Zoo Subway Station is taken as an example and the feasibility of the model results is verified through the comparison of the actual data and simulation data. The simulation results show that the improved model can depict the microscopic behaviors of pedestrians in subway station.

Waste Transfer Stations

DEFF Research Database (Denmark)

Christensen, Thomas Højlund

2011-01-01

tion and transport is usually the most costly part of any waste management system; and when waste is transported over a considerable distance or for a long time, transferring the waste from the collection vehicles to more efficient transportation may be economically beneficial. This involves...... a transfer station where the transfer takes place. These stations may also be accessible by private people, offering flexibility to the waste system, including facilities for bulky waste, household hazardous waste and recyclables. Waste transfer may also take place on the collection route from small...... describes the main features of waste transfer stations, including some considerations about the economical aspects on when transfer is advisable....

Slope Monitoring using Total Station: What are the Challenges and ...

African Journals Online (AJOL)

Afeni

implications of incorrect use or negligence during slope monitoring surveys ... Data collection, processing and the presentation of results in a concise format ..... There are several software packages on the market for total station error propagation, ..... Thomas, H.G., 2011, Slope stability prism monitoring: A guide for practising ...

Meyrin Petrol Station

CERN Multimedia

2006-01-01

Please note that the Meyrin petrol station will be closed for maintenance work on Tuesday 19 and Wednesday 20 December 2006. If you require petrol during this period we invite you to use the Prévessin petrol station, which will remain open. TS-IC-LO Section Tel.: 77039 - 73793

Intrinsically Safe and Economical Reactor (ISER)

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki; Asahi, Yoshiro

1991-01-01

The Intrinsically Safe and Economical Reactor (ISER) is designed based on the principle of a process inherent ultimate safe reactor, PIUS, a so-called inherently safe reactor (ISR). ISER has been developed joingly by the members of the Kanagawa Institute of Technology, the University of Tokyo, the Japan Atomic Energy Research Institute (JAERI) and several industrial firms in Japan. This paper describes the requirements for the next generation of power reactor, the safety design philosphy of ISR and ISER, the controllability of ISER and the results of analyses of some of the design-based accidents (DBA) of ISER, namely station blackout, accidents in which the pressurizer relief valve becomes jammed and stuck in open position and tube breaks in the steam generator. It is concluded that the ISER can ensure a wide range of contraollabitily and fuel integrity for all the analysed DBAs. (orig.)

Total organizational team building at the V.C. Summer Nuclear Station

International Nuclear Information System (INIS)

Carswell, R.W.; Parks, J.W.

1988-01-01

The V.C. Summer station of South Carolina Electric and Gas Company began a total organizational team building program in 1985 as one of several methods by which to make overall operational improvements. The team model selected was developed by Carswell ampersand Associated, Inc. The model has been used successfully in a number of nonnuclear applications. This was its first usage in a nuclear power plant. Total organizational team building has had positive results at the V.C. Summer station. The data suggest that including all employees in decision making can assist with increasing efficiency, morale, and safety in a nuclear power plant

Experience from the construction and operation of Tarapur and Rajasthan Nuclear Power Stations

International Nuclear Information System (INIS)

Shah, J.C.; Pardiwala, T.F.; Kothare, V.V.; Rao, M.H.P.; Nanjundeswaran, K.

1977-01-01

India's experience in construction and operation of nuclear power stations so far covers two BWR and four PHWR Units in three power stations. Two more PHWR units are at an early stage of construction. The twin unit Tarapur Station (2x210 MWe BWR) was built as a turnkey project which restricted participation of Indian engineers in design and construction considerably. The contrasting approach adopted for Rajasthan Station (2x220 MWe PHWR) involved Indian personnel and contractors fully in construction and commissioning, with Canadians providing supervisory assistance in Rajasthan I and essentially consultative help for Rajasthan II. Subsequent stations are wholly Indian efforts. Tarapur went into commercial operation in 1969, 60 months after breaking of ground. Construction was essentially uneventful, major problems faced being stress corrosion induced cracks in the reactor lining and complete change of steam generator tubes. In its seven years of operations, Tarapur has faced several problems mainly arising from rather early designs, indifferent fuel performance, constraints of twin-unit approach and operations in an inadequately developed grid system apart from those generally stemming from assimilation of an advanced technology in a developing country. The Station has undergone six refuellings during this period. Most of the problems have been overcome by design changes, system augmentations and experience and the Station operation since mid 1974 have generally been steady at around 90% of the rated capacity. Construction of Rajasthan I at a remote and isolated site proceeded relatively slowly. Local availability of skilled and semi-skilled manpower was poor, affecting construction. Inadequate roads impeded movements of overdimensioned components. Observing strict Quality Assurance standards required several major rectifications at site. Rajasthan I went on line in 1973 after overcoming major turbine bearing problems during commissioning. Since then, while

Evaluation of River Bend Station Unit 1 Technical Specifications

International Nuclear Information System (INIS)

Baxter, D.E.; Bruske, S.J.

1985-08-01

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the River Bend Station Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the requirements of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the River Bend T/S. Several discrepancies were identified and subsequently resolved through discussions with the cognizant NRC reviewer, NRC staff reviewers and/or utility representatives. The River Bend Station Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

Enhancing base station connectivity with directional information of movement

Directory of Open Access Journals (Sweden)

Eugene Kim

2016-09-01

Full Text Available High-performance smartphones with various sensor capabilities, including the inertial measurement unit, allow us to have higher freedom to localize our movements and utilize them for other applications. In this paper, we present a simple method to estimate the heading of a user using only a digital compass, accelerometer, and gyroscope. Several experiments were conducted using an Apple iPhone. In addition, we also present an algorithm to reduce the number of handover attempts in a dense base station environment. A series of simulations were conducted on OPNET in various emulated environments to present the effectiveness of the proposed method. The simulation results show that our proposed algorithm effectively reduces the ping-pong effect in a dense base station environment.

Assessment of radiofrequency/microwave radiation emitted by the antennas of rooftop-mounted mobile phone base stations

International Nuclear Information System (INIS)

Keow, M. A.; Radiman, S.

2006-01-01

Radiofrequency (RF) and microwave (MW) radiation exposures from the antennas of rooftop-mounted mobile telephone base stations have become a serious issue in recent years due to the rapidly evolving technologies in wireless telecommunication systems. In Malaysia, thousands of mobile telephone base stations have been erected all over the country, most of which are mounted on the rooftops. In view of public concerns, measurements of the RF/MW levels emitted by the base stations were carried out in this study. The values were compared with the exposure limits set by several organisations and countries. Measurements were performed at 200 sites around 47 mobile phone base stations. It was found that the RF/MW radiation from these base stations were well below the maximum exposure limits set by various agencies. (authors)

Base Station Performance Model

OpenAIRE

Walsh, Barbara; Farrell, Ronan

2005-01-01

At present the testing of power amplifiers within base station transmitters is limited to testing at component level as opposed to testing at the system level. While the detection of catastrophic failure is possible, that of performance degradation is not. This paper proposes a base station model with respect to transmitter output power with the aim of introducing system level monitoring of the power amplifier behaviour within the base station. Our model reflects the expe...

Review of freeboard: Grand Rapids Generating Station

Energy Technology Data Exchange (ETDEWEB)

Groeneveld, J.L.; Harding, W.B.; Bonin, D.V.; Fuchs, D.M. [Acres Manitoba Ltd., Winnipeg, MB (Canada); Warner, B.J. [Manitoba Hydro, Winnipeg, MB (Canada)

2001-10-01

Constructed during the period 1960-1965, the Grand Rapids Generating Station is a 472 MW hydroelectric station located approximately 400 kilometres northwest of Winnipeg, Manitoba, on the Saskatchewan River. An intake structure, four penstocks, a four-unit plus house unit powerhouse, wing walls, extensive dyke structures and a four-bay spillway are the components of the generating station. A little over ten years ago, the Manitoba Hydro Dam Safety Program was initiated. The program included a detailed dam safety review of the Grand rapids Generating Station. A potential deficiency in the freeboard allowance for several of the earthen dykes was revealed by the review process. The dam safety guidelines sponsored by the Canadian Dam Association were not met. The occurrence of a 1:1000 year wind event from the critical direction when the reservoir was at or near its full supply level was compounded by the analysis. The establishment of a wind and wave monitoring program was included in the deficiency studies commissioned. The objective was to confirm the empirical estimates concerning wave height, the development and usage of a two dimensional numerical wave model, and additional freeboard analyses to refine estimates of the recurrence interval of the current level of protection. A statistical Monte Carlo analysis was performed in the case of the estimates of the recurrence interval to determine the joint probabilities of seasonal variations in wind direction, wind speed, and reservoir level. The estimate of the actual risk of overtopping was therefore refined, and the details pertaining to the methodology and the conclusions of the analysis are all included in the paper. 15 refs., 4 tabs., 9 figs.

Space Station fluid management logistics

Science.gov (United States)

Dominick, Sam M.

1990-01-01

Viewgraphs and discussion on space station fluid management logistics are presented. Topics covered include: fluid management logistics - issues for Space Station Freedom evolution; current fluid logistics approach; evolution of Space Station Freedom fluid resupply; launch vehicle evolution; ELV logistics system approach; logistics carrier configuration; expendable fluid/propellant carrier description; fluid carrier design concept; logistics carrier orbital operations; carrier operations at space station; summary/status of orbital fluid transfer techniques; Soviet progress tanker system; and Soviet propellant resupply system observations.

Development and Performance of the Alaska Transportable Array Posthole Broadband Seismic Station

Science.gov (United States)

Aderhold, K.; Enders, M.; Miner, J.; Bierma, R. M.; Bloomquist, D.; Theis, J.; Busby, R. W.

2017-12-01

The final stations of the Alaska Transportable Array (ATA) will be constructed in 2017, completing the full footprint of 280 new and existing broadband seismic stations stretching across 19 degrees of latitude from western Alaska to western Canada. Through significant effort in planning, site reconnaissance, permitting and the considerable and concerted effort of field crews, the IRIS Alaska TA team is on schedule to successfully complete the construction of 194 new stations and upgrades at 28 existing stations over four field seasons. The station design and installation method was developed over the course of several years, leveraging the experience of the L48 TA deployments and existing network operators in Alaska as well as incorporating newly engineered components and procedures. A purpose-built lightweight drill was designed and fabricated to facilitate the construction of shallow boreholes to incorporate newly available posthole seismometers. This allowed for the development of a streamlined system of procedures to manufacture uniform seismic stations with minimal crew and minimal time required at each station location. A new station can typically be constructed in a single day with a four-person field crew. The ATA utilizes a hammer-drilled, cased posthole emplacement method adapted to the remote and harsh working environment of Alaska. The same emplacement design is implemented in all ground conditions to preserve uniformity across the array and eliminate the need for specialized mechanical equipment. All components for station construction are ideally suited for transport via helicopter, and can be adapted to utilize more traditional methods of transportation when available. This emplacement design delivers high quality data when embedded in bedrock or permafrost, reaching the low noise levels of benchmark permanent global broadband stations especially at long periods over 70 seconds. The TA will operate the network of real-time stations through at least

Hammond Bay Biological Station

Data.gov (United States)

Federal Laboratory Consortium — Hammond Bay Biological Station (HBBS), located near Millersburg, Michigan, is a field station of the USGS Great Lakes Science Center (GLSC). HBBS was established by...

47 CFR 74.793 - Digital low power TV and TV translator station protection of broadcast stations.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 4 2010-10-01 2010-10-01 false Digital low power TV and TV translator station... DISTRIBUTIONAL SERVICES Low Power TV, TV Translator, and TV Booster Stations § 74.793 Digital low power TV and TV translator station protection of broadcast stations. (a) An application to construct a new digital low power...

An Integration of the Restructured Melcor for the Midas Computer Code

International Nuclear Information System (INIS)

Sunhee Park; Dong Ha Kim; Ko-Ryu Kim; Song-Won Cho

2006-01-01

The developmental need for a localized severe accident analysis code is on the rise. KAERI is developing a severe accident code called MIDAS, which is based on MELCOR. In order to develop the localized code (MIDAS) which simulates a severe accident in a nuclear power plant, the existing data structure is reconstructed for all the packages in MELCOR, which uses pointer variables for data transfer between the packages. During this process, new features in FORTRAN90 such as a dynamic allocation are used for an improved data saving and transferring method. Hence the readability, maintainability and portability of the MIDAS code have been enhanced. After the package-wise restructuring, the newly converted packages are integrated together. Depending on the data usage in the package, two types of packages can be defined: some use their own data within the package (let's call them independent packages) and the others share their data with other packages (dependent packages). For the independent packages, the integration process is simple to link the already converted packages together. That is, the package-wise structuring does not require further conversion of variables for the integration process. For the dependent packages, extra conversion is necessary to link them together. As the package-wise restructuring converts only the corresponding package's variables, other variables defined from other packages are not touched and remain as it is. These variables are to be converted into the new types of variables simultaneously as well as the main variables in the corresponding package. Then these dependent packages are ready for integration. In order to check whether the integration process is working well, the results from the integrated version are verified against the package-wise restructured results. Steady state runs and station blackout sequences are tested and the major variables are found to be the same each other. In order to verify the results, the integrated

Implementation green and low cost on landscape design of Manggarai Integrated Station, Jakarta

Science.gov (United States)

Suryanti, T.; Meilianti, H.

2018-01-01

The Manggarai Integrated Station is the transit of various transportation modes. The Integrated Station located in Manggarai Jakarta and managed by PT. KAI. The Manggarai station is integrated and have terminal nature of transit areas (switching mode of transportation). There are several problems in the site, such as the problem of the site conditions in the urban area, topography, soil, vegetation, space, visual, users on the site can provide ideas for the concepts. The data was analyzed using the quantitative descriptive methode. The purpose of this research is to design the integrated station atmosphere, not only can support of the activities station users, but can also accommodate the needs of the community. It will “Green, Low cost” at the Manggarai integrated transport transit station in Jakarta. The potential that exists in this area is the lowliest integrated from various areas of the mode of transportation that make the users to facilitate transit transportation to the other. The basic concept of this design refers to the “Green, Low Cost” which unite with theme “user friendly” land use on a more efficient and effective site. The result of this research is landscape design development of Manggarai integrated station. Its consists of landscape design in west and east area, transition area, parking area, solar panel area, and social interaction area.

Port construction works in the Sendai Nuclear Power Station

International Nuclear Information System (INIS)

Narahara, Akio; Minamata, Hisashi; Harada, Kensaku

1982-01-01

Sendai Nuclear Power Station is the second nuclear power station of Kyushu Electric Power Co., Inc., with two PWR plants of 890 MW each, and the operation of No.1 plant will be started in July, 1984, and that of No.2 plant in March, 1986. The civil engineering works for both plants were started in June, 1978, and March, 1981, respectively, and the rate of progress as of the end of September, 1982, was 97 % and 66 %, respectively. In the construction of this power station, the port facility was provided for the transport of construction materials and spent fuel, and for the intake of condenser cooling water. In order to make the construction by dry work, the double cofferdam structures with steel sheet piles were made offshore. The use of the wharf was started in March, 1980, though typhoons hit the area several times, and the dredging in the port was completed in May, 1982. The outline of the plan of this power station, the state of affairs before the start of construction, the outline of the port construction works, the topography, geological features and sea conditions, the design of the port such as breakwaters, unloading wharf and water intake, the manufacture and installation of caissons, dredging, and the temporary cofferdam works for water intake are described. (author)

Big Game Reporting Stations

Data.gov (United States)

Vermont Center for Geographic Information — Point locations of big game reporting stations. Big game reporting stations are places where hunters can legally report harvested deer, bear, or turkey. These are...

Weigh-in-Motion Stations

Data.gov (United States)

Department of Homeland Security — The data included in the GIS Traffic Stations Version database have been assimilated from station description files provided by FHWA for Weigh-in-Motion (WIM), and...

Air and radiation monitoring stations

CERN Multimedia

AUTHOR|(SzGeCERN)582709

2015-01-01

CERN has around 100 monitoring stations on and around its sites. New radiation measuring stations, capable of detecting even lower levels of radiation, were installed in 2014. Two members of HE-SEE group (Safety Engineering and Environment group) in front of one of the new monitoring stations.

47 CFR 73.1120 - Station location.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 4 2010-10-01 2010-10-01 false Station location. 73.1120 Section 73.1120... Rules Applicable to All Broadcast Stations § 73.1120 Station location. Each AM, FM, TV and Class A TV... be the geographical station location. [65 FR 30003, May 10, 2000] ...

Water Level Station History

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Images contain station history information for 175 stations in the National Water Level Observation Network (NWLON). The NWLON is a network of long-term,...

Cost comparison of 4x500 MW coal-fuelled and 4x850 MW CANDU nuclear generating stations

International Nuclear Information System (INIS)

Costa, M.

1981-01-01

The lifetime costs for a 4x850 MW CANDU generating station are compared to those for 4x500 MW bituminous coal-fuelled generating stations. Two types of coal-fuelled stations are considered; one burning U.S. coal which includes flue gas desulfurization and one burning Western Canadian coal. Current estimates for the capital costs, operation and maintenance costs, fuel costs, decommissioning costs and irradiated fuel management costs are shown. The results show: (1) The accumulated discounted costs of nuclear generation, although initially higher, are lower than coal-fuelled generation after two or three years. (2) Fuel costs provide the major contribution to the total lifetime costs for coal-fuelled stations whereas capital costs are the major item for the nuclear station. (3) The break even lifetime capacity factor between nuclear and U.S. coal-fuelled generation is projected to be 5%; that for nuclear and Canadian coal-fuelled generation is projected to be 9%. (4) Large variations in the costs are required before the cost advantage of nuclear generation is lost. (5) Comparison with previous results shows that the nuclear alternative has a greater cost advantage in the current assessment. (6) The total unit energy cost remains approximately constant throughout the station life for nuclear generation while that for coal-fuelled generation increases significantly due to escalating fuel costs. The 1978 and 1979 actual total unit energy cost to the consumer for several Ontario Hydro stations are detailed, and projected total unit energy costs for several Ontario Hydro stations are shown in terms of escalated dollars and in 1980 constant dollars

Reduction of carbon dioxide emissions by modernisation or substitution of existing coal power stations in the EC and the resulting cost effect

International Nuclear Information System (INIS)

Folke, C.

2000-01-01

Trends in carbon dioxide emissions and cost of coal power stations are analyzed EC-wide. The data base derives from several generations of power stations for the period between 2000 and 2050 assuming several different substitution strategies. The reference scenario is one in which decisions on new power station types are made purely on the basis of short-term economic aspects (annual minimisation of the nominal cost). The method and the parameters used for calculating and analyzing carbon dioxide strategies in the context of this study are presented and explained [de

NASA Glenn Research Center Solar Cell Experiment Onboard the International Space Station

Science.gov (United States)

Myers, Matthew G.; Wolford, David S.; Prokop, Norman F.; Krasowski, Michael J.; Parker, David S.; Cassidy, Justin C.; Davies , William E.; Vorreiter, Janelle O.; Piszczor, Michael F.; Mcnatt, Jeremiah S.;

The modeling and analysis of in-vessel corium/structure interaction in boiling water reactors

International Nuclear Information System (INIS)

Podowski, M.Z.; Kurul, N.; Kim, S.-W.; Baltyn, W.; Frid, W.

1997-01-01

A complete stand-alone state-of-the-art model has been developed of the interaction between corium debris in the lower plenum and the RPV walls and internal structures, including the vessel failure mechanisms. This new model has been formulated as a set of consistent computer modules which could be linked with other existing models and/or computer codes. The combined lower head and lower plenum modules were parametrically tested and applied to predict the consequences of a hypothetical station blackout in a Swedish BWR. (author)

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

International Nuclear Information System (INIS)

Silva Filho, E.

1986-01-01

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

Dose management programmes at Kaiga Generating Station

International Nuclear Information System (INIS)

Vijayan, P.; Prabhakaran, V.; Managavi, Sadashiv B.; Danannavar, Veerendra; Biju, P.; Manoj Kumar, M.; Shrikrishna, U.V.

2001-01-01

Kaiga Generating Station (KGS) has two units of pressurized heavy water reactors of 220 MWe each capacity. KGS-2 started power generation since 1999 and KGS-1 since 2000. Several programmes such as assessment of radioactive condition, training on radiological safety aspects, job planning in radioactive areas, etc. are conducted periodically to implement an effective dose control programmes in KGS. These efforts are briefly discussed in this report. Facilities and techniques to implement ALARA programs are also highlighted in this report. (author)

Feasibility study of cocos, condensation of containment atmosphere on structures

International Nuclear Information System (INIS)

Rij, H.M. van; Vonka, V.

1989-12-01

The aim of this report is to assess the state of the art of the knowledge of the thermo-hydraulic conditions within a LWR containment in order to determine both the radioactive and the non-radioactive aerosol deposition rates during a severe reactor accident. The radioactive aerosol in the containment atmosphere is, together with the noble gases, responsible for the radioactive source term into the biosphere when a containment failure occurs. The dominant aerosol removal mechanisms depend strongly upon the thermal-hydraulic state of the containment atmosphere. It is demonstrated that the thermohydraulic state, determined by heat sources and the sensible heat transport, is predominantly super-heated when fission products are released into the containment. Hence the thermohydraulic conditions are not favorable for an intensive bulk condensation onto aerosol particles during an essential period of time. A station black-out scenario, in which the primary system of the considered 500 MWe PWR with a dry cavity is depressurized prior to vessel failure, is used as an example to demonstrate this effect. The results, obtained with the CONTAIN code, show the relevance of the sensible heat transport in the period of time (c.a. 30 minutes) between the end of the injection of the steam and fission products into the containment, and the molten core concrete interaction. All considered variation of the station black-out scenario, in which the decay heat dissipated in the containment atmosphere has been 10% of the total decay heat, indicate that the atmosphere becomes super-heated within the 30 minutes. Reducing the fraction of the decay heat in the containment from the 10% to 5% increases the time period with saturated conditions. The amount of the decay heat released into the containment atmosphere forms the major factor determining the thermohydraulic state. It influences the duration of the bulk condensation period, which in turn has an influence on the aerosol deposition

Local society and nuclear power stations

International Nuclear Information System (INIS)

1984-02-01

This report was made by the expert committee on region investigation, Japan Atomic Industrial Forum Inc., in fiscal years 1981 and 1982 in order to grasp the social economic influence exerted on regions by the location of nuclear power stations and the actual state of the change due to it, and to search for the way the promotion of local community should be. The influence and the effect were measured in the regions around the Fukushima No. 1 Nuclear Power Station of Tokyo Electric Power Co., Inc., the Mihama Power Station of Kansai Electric Power Co., Inc., and the Genkai Nuclear Power Station of Kyushu Electric Power Co., Inc. The fundamental recognition in this discussion, the policy of locating nuclear power stations and the management of regions, the viewpoint and way of thinking in the investigation of the regions where nuclear power stations are located, the actual state of social economic impact due to the location of nuclear power stations, the connected mechanism accompanying the location of nuclear power stations, and the location of nuclear power stations and the acceleration of planning for regional promotion are reported. In order to economically generate electric power, the rationalization in the location of nuclear power stations is necessary, and the concrete concept of building up local community must be decided. (Kako, I.)

CANDU severe accident management guidance update

International Nuclear Information System (INIS)

Jones, L.; Popov, N.; Gilbert, L.; Weed, J.

2014-01-01

The CANDU Owners Group (COG) developed a set of generic and initial station-specific Severe Accident Management Guidance (SAMG) documents to mitigate the consequences to the public in the event of a severe accident. The generic portion of the COG SAMG was completed in 2006; the overall project including the station-specific phase was completed in April 2007. Over the years, the CANDU industry and utilities have continuously increased the knowledge base for SAMG and have incorporated various engineered features based on the knowledge obtained. As a result of the event that occurred at the Fukushima Daiiachi nuclear power plant (NPP) in Japan, the Canadian Nuclear Safety Commission (CNSC) established the CNSC Fukushima Task Force. The results of the task force were documented in INFO-0828, CNSC Staff Action Plan on the CNSC Fukushima Task Force Recommendations. Among the recommendation documented in INFO-828 were Fukushima Action Items (FAIs) directed towards the CANDU utilities in Canada; a portion of which are related to SAMG documentation updates and directed at enhancing SAM response. A COG joint project was established to support the closure of the CNSC FAIs and to revise the current CANDU documentation accordingly. This paper provides a high level summary of the COG project scope and results. It also demonstrates that the CANDU SAMG programs in Canada provide robust protection and mitigation of severe accidents. (author)

CANDU severe accident management guidance update

Energy Technology Data Exchange (ETDEWEB)

Jones, L., E-mail: lisa.m.jones@opg.com [Ontario Power Generation, Pickering, ON (Canada); Popov, N., E-mail: nik.popov@rogers.com [Candu Owners Group, Toronto, ON (Canada); Gilbert, L., E-mail: lovell.gilbert@brucepower.com [Bruce Power, Tiverton, ON (Canada); Weed, J., E-mail: jeff.weed@candu.gov [Candu Owners Group, Toronto, ON (Canada)

2014-07-01

The CANDU Owners Group (COG) developed a set of generic and initial station-specific Severe Accident Management Guidance (SAMG) documents to mitigate the consequences to the public in the event of a severe accident. The generic portion of the COG SAMG was completed in 2006; the overall project including the station-specific phase was completed in April 2007. Over the years, the CANDU industry and utilities have continuously increased the knowledge base for SAMG and have incorporated various engineered features based on the knowledge obtained. As a result of the event that occurred at the Fukushima Daiiachi nuclear power plant (NPP) in Japan, the Canadian Nuclear Safety Commission (CNSC) established the CNSC Fukushima Task Force. The results of the task force were documented in INFO-0828, CNSC Staff Action Plan on the CNSC Fukushima Task Force Recommendations. Among the recommendation documented in INFO-828 were Fukushima Action Items (FAIs) directed towards the CANDU utilities in Canada; a portion of which are related to SAMG documentation updates and directed at enhancing SAM response. A COG joint project was established to support the closure of the CNSC FAIs and to revise the current CANDU documentation accordingly. This paper provides a high level summary of the COG project scope and results. It also demonstrates that the CANDU SAMG programs in Canada provide robust protection and mitigation of severe accidents. (author)

47 CFR 25.137 - Application requirements for earth stations operating with non-U.S. licensed space stations.

Science.gov (United States)

2010-10-01

... space stations. (a) Earth station applicants or entities filing a “letter of intent” or “Petition for... Union. (d) Earth station applicants requesting authority to operate with a non-U.S.-licensed space... 47 Telecommunication 2 2010-10-01 2010-10-01 false Application requirements for earth stations...

Burning issue of energy problem after Fukushima disaster of TEPCO's atomic power stations

International Nuclear Information System (INIS)

Harada, Shoji

2012-01-01

Strikes of unanticipated enormous earthquake and subsequent tsunami brought unbelievable disaster in eastern Japan on March 11, 2012. In particular, collapse of cooling system of TEPCO's Fukushima atomic power stations resulted in IAEA-defined level7 accident including heavy radiation, hydrogen explosion -induced collapse of the building of power station No.2 and No.4 and melt through of nuclear pressure vessel No1.3.4 At an initial stage of the disaster, nobody knew precisely what happened at the power stations. According to the recent report of the national investigation committee, precise reason of the collapse of the cooling system whether it was induced by the strike of huge earthquake or tsunami is still unclear. Due to poor risk management of the government and TEPCO and closure of the precise disaster information, people became suspicious and nervous about the atomic power station. Fifty four atomic power stations have been constructed for these forty years in Japan. On last May 04, all the atomic power stations were shut down due to periodic inspection. However, restart of them became hot discussion. Although atomic power station was regarded as a powerful tool to reduce carbon dioxide several years ago, this situation after March 11 completely changed. In many countries which possess atomic power station, making a road map to develop recyclable energy is a burning issue. It should be noted that German spent about thirty years to declare atomic energy free society. Finally necessity of succession of technology of utilizing atomic power is emphasized. Politics on depending atomic power differs in each country. Therefore, study from Fukushima disaster should be widely used to prevent from unexpected accident of atomic power station.

Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems

Directory of Open Access Journals (Sweden)

Siniša Šadek

2017-01-01

Full Text Available The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France and Gesellschaft für Anlagen und Reaktorsicherheit (GRS, Germany. Codes’ overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes’ modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper.

Characterization of airborne bacteria at an underground subway station.

Science.gov (United States)

Dybwad, Marius; Granum, Per Einar; Bruheim, Per; Blatny, Janet Martha

2012-03-01

The reliable detection of airborne biological threat agents depends on several factors, including the performance criteria of the detector and its operational environment. One step in improving the detector's performance is to increase our knowledge of the biological aerosol background in potential operational environments. Subway stations are enclosed public environments, which may be regarded as potential targets for incidents involving biological threat agents. In this study, the airborne bacterial community at a subway station in Norway was characterized (concentration level, diversity, and virulence- and survival-associated properties). In addition, a SASS 3100 high-volume air sampler and a matrix-assisted laser desorption ionization-time of flight mass spectrometry-based isolate screening procedure was used for these studies. The daytime level of airborne bacteria at the station was higher than the nighttime and outdoor levels, and the relative bacterial spore number was higher in outdoor air than at the station. The bacterial content, particle concentration, and size distribution were stable within each environment throughout the study (May to September 2010). The majority of the airborne bacteria belonged to the genera Bacillus, Micrococcus, and Staphylococcus, but a total of 37 different genera were identified in the air. These results suggest that anthropogenic sources are major contributors to airborne bacteria at subway stations and that such airborne communities could harbor virulence- and survival-associated properties of potential relevance for biological detection and surveillance, as well as for public health. Our findings also contribute to the development of realistic testing and evaluation schemes for biological detection/surveillance systems by providing information that can be used to mimic real-life operational airborne environments in controlled aerosol test chambers.

Characterization of Airborne Bacteria at an Underground Subway Station

Science.gov (United States)

Dybwad, Marius; Granum, Per Einar; Bruheim, Per

2012-01-01

The reliable detection of airborne biological threat agents depends on several factors, including the performance criteria of the detector and its operational environment. One step in improving the detector's performance is to increase our knowledge of the biological aerosol background in potential operational environments. Subway stations are enclosed public environments, which may be regarded as potential targets for incidents involving biological threat agents. In this study, the airborne bacterial community at a subway station in Norway was characterized (concentration level, diversity, and virulence- and survival-associated properties). In addition, a SASS 3100 high-volume air sampler and a matrix-assisted laser desorption ionization–time of flight mass spectrometry-based isolate screening procedure was used for these studies. The daytime level of airborne bacteria at the station was higher than the nighttime and outdoor levels, and the relative bacterial spore number was higher in outdoor air than at the station. The bacterial content, particle concentration, and size distribution were stable within each environment throughout the study (May to September 2010). The majority of the airborne bacteria belonged to the genera Bacillus, Micrococcus, and Staphylococcus, but a total of 37 different genera were identified in the air. These results suggest that anthropogenic sources are major contributors to airborne bacteria at subway stations and that such airborne communities could harbor virulence- and survival-associated properties of potential relevance for biological detection and surveillance, as well as for public health. Our findings also contribute to the development of realistic testing and evaluation schemes for biological detection/surveillance systems by providing information that can be used to mimic real-life operational airborne environments in controlled aerosol test chambers. PMID:22247150

Harmonic tidal analysis at a few stations using the least squares method

Digital Repository Service at National Institute of Oceanography (India)

Fernandes, A.A.; Das, V.K.; Bahulayan, N.

Using the least squares method, harmonic analysis has been performed on hourly water level records of 29 days at several stations depicting different types of non-tidal noise. For a tidal record at Mormugao, which was free from storm surges (low...

Advances in power station construction

International Nuclear Information System (INIS)

Anon.

1986-01-01

This book is about power stations - specifically about the construction of modern power stations by the Central Electricity Generating Board in England and Wales over the past decade. It describes the work of the CEGB's Generation Development and Construction Division, perhaps better known throughout the world as simply 'Barnwood' where it has its Headquarters in Gloucester, UK. Barnwood was formed in the early 1970s to concentrate the CEGB's then dispersed engineering construction resources to cope with the smaller number but greatly increased size and complexity of modern power station projects. Perhaps uniquely over the ten years since its formation Barnwood has managed the construction of all types of station; coal-fired, oil-fired, nuclear, pumped storage and hydro. This book tells the story of these various projects and gives detailed descriptions of the respective stations. However, it is not intended as a comprehensive description of power station technology. Rather it is intended to convey the scale of such projects and the many decisions and compromises which have to be made in the course of managing their construction

The Princess Elisabeth Station

Science.gov (United States)

Berte, Johan

2012-01-01

Aware of the increasing impact of human activities on the Earth system, Belgian Science Policy Office (Belspo) launched in 1997 a research programme in support of a sustainable development policy. This umbrella programme included the Belgian Scientific Programme on Antarctic Research. The International Polar Foundation, an organization led by the civil engineer and explorer Alain Hubert, was commissioned by the Belgian Federal government in 2004 to design, construct and operate a new Belgian Antarctic Research Station as an element under this umbrella programme. The station was to be designed as a central location for investigating the characteristic sequence of Antarctic geographical regions (polynia, coast, ice shelf, ice sheet, marginal mountain area and dry valleys, inland plateau) within a radius of 200 kilometers (approx.124 miles) of a selected site. The station was also to be designed as "state of the art" with respect to sustainable development, energy consumption, and waste disposal, with a minimum lifetime of 25 years. The goal of the project was to build a station and enable science. So first we needed some basic requirements, which I have listed here; plus we had to finance the station ourselves. Our most important requirement was that we decided to make it a zero emissions station. This was both a philosophical choice as we thought it more consistent with Antarctic Treaty obligations and it was also a logistical advantage. If you are using renewable energy sources, you do not have to bring in all the fuel.

Utilization of waste heat from electricity generating stations

International Nuclear Information System (INIS)

Robertson, R.F.S.

1977-06-01

Historically the nuclear power station has been designed solely as an electricity producer. But in Canada today only 15 percent of our energy consumption is as electricity. The non-electrical needs today are supplied almost entirely by natural gas and oil. There is an incentive to see whether a nuclear station could supply energy for some of these non-electrical needs, thus freeing gas and oil for uses for which they may be more valuable and suitable, especially in transportation. A group located at the Whiteshell Nuclear Research Establishment undertook a series of studies to examine this problem. These studies were done in sufficient depth to provide technological and economic answers, and as a result several reports have been published on various topics. In this report, the findings from these studies are drawn together in an assessment of the potential in Canada for using waste heat. (author)

Revision of IRIS/IDA Seismic Station Metadata

Science.gov (United States)

Xu, W.; Davis, P.; Auerbach, D.; Klimczak, E.

2017-12-01

Trustworthy data quality assurance has always been one of the goals of seismic network operators and data management centers. This task is considerably complex and evolving due to the huge quantities as well as the rapidly changing characteristics and complexities of seismic data. Published metadata usually reflect instrument response characteristics and their accuracies, which includes zero frequency sensitivity for both seismometer and data logger as well as other, frequency-dependent elements. In this work, we are mainly focused studying the variation of the seismometer sensitivity with time of IRIS/IDA seismic recording systems with a goal to improve the metadata accuracy for the history of the network. There are several ways to measure the accuracy of seismometer sensitivity for the seismic stations in service. An effective practice recently developed is to collocate a reference seismometer in proximity to verify the in-situ sensors' calibration. For those stations with a secondary broadband seismometer, IRIS' MUSTANG metric computation system introduced a transfer function metric to reflect two sensors' gain ratios in the microseism frequency band. In addition, a simulation approach based on M2 tidal measurements has been proposed and proven to be effective. In this work, we compare and analyze the results from three different methods, and concluded that the collocated-sensor method is most stable and reliable with the minimum uncertainties all the time. However, for epochs without both the collocated sensor and secondary seismometer, we rely on the analysis results from tide method. For the data since 1992 on IDA stations, we computed over 600 revised seismometer sensitivities for all the IRIS/IDA network calibration epochs. Hopefully further revision procedures will help to guarantee that the data is accurately reflected by the metadata of these stations.

UK 2009-2010 repeat station report

Directory of Open Access Journals (Sweden)

Thomas J.G. Shanahan

2013-03-01

Full Text Available The British Geological Survey is responsible for conducting the UK geomagnetic repeat station programme. Measurements made at the UK repeat station sites are used in conjunction with the three UK magnetic observatories: Hartland, Eskdalemuir and Lerwick, to produce a regional model of the local field each year. The UK network of repeat stations comprises 41 stations which are occupied at approximately 3-4 year intervals. Practices for conducting repeat station measurements continue to evolve as advances are made in survey instrumentation and as the usage of the data continues to change. Here, a summary of the 2009 and 2010 UK repeat station surveys is presented, highlighting the measurement process and techniques, density of network, reduction process and recent results.

As Blood Alcohol Content (BAC) Increases, So Does Impairment | NIH MedlinePlus the Magazine

Science.gov (United States)

... 30% Severe Impairment – 0.16 - 0.30% Speech, memory, coordination, attention, reaction time, balance significantly impaired All driving-related skills dangerously impaired Judgement and decision making dangerously impaired Blackouts (amnesia) Vomiting and other signs ...

Prevalence of low back pain among filling stations attendants in ...

African Journals Online (AJOL)

... 12 months prevalence of LBP prior to the study, severity of LBP, history of LBP, causative factors and management of LBP and its effect on filling stations attendant's activities and activities of daily living. It was adapted from a previous study Titled Prevalence of Low Back Pain among Traffic Warden in Lagos state. The data ...

Influence of Great East Japan Earthquake on neutron source station in J-PARC

International Nuclear Information System (INIS)

Sakai, Kenji; Sakamoto, Shinichi; Kinoshita, Hidetaka; Seki, Masakazu; Haga, Katsuhiro; Kogawa, Hiroyuki; Wakui, Takashi; Naoe, Takashi; Kasugai, Yoshimi; Tatsumoto, Hideki; Aso, Tomokazu; Hasegawa, Shoichi; Maekawa, Fujio; Oikawa, Kenichi; Ooi, Motoki; Watanabe, Akihiko; Teshigawara, Makoto; Meigo, Shin-ichiro; Ikezaki, Kiyomi; Akutsu, Atsushi; Harada, Masahide; Takada, Hiroshi; Futakawa, Masatoshi

2012-03-01

This report investigates the behavior, damage and restoration of each component in a neutron source station of the Materials and Life Science Experimental Facility (MLF) of J-PARC at the time of the Great East Japan Earthquake (M9.0) and verified the safety design for emergency accidents in the neutron source station. The neutron source station of the MLF at the J-PARC generates neutrons by injecting proton beams into a mercury target, and supplies to user experimental apparatuses. It consists of the mercury target, three moderators filled with supercritical hydrogen, reflectors, water cooling shields, a vessel filled with helium gas, neutron beam shutters, biological-shields and so on. In case of loss of their external electric power supply, a control function for the source station is kept by an emergency power supply. According to interlock sequences in an emergency, a signal for terminating the beam operation is transmitted, the circulators shut down automatically, and the hydrogen gas is released out of the building. On March 11 in 2011, strong shocks caused by the earthquake were observed all over Ibaraki prefecture. At the date, a status of the source station was ready for the restart of beam operation. In the MLF, after strong quakes were detected at the several instruments, the external power supply was lost, all of the circulators shut down automatically, and the hydrogen gas was released. The leakages of mercury, hydrogen and radio-activation gases did not occur. While, the quakes made gaps between the shield blocks and ruptured external pipe lines for compressed air and water by subsidence around the building. But significant damages to the components were not found though the pressure drop of compressed air lines influenced on the mercury target trolley lock system and pneumatic operation values. These results substantiated the validity of the safety design for emergency accidents in the neutron source station in the MLF, and suggested several points

Thermal management of space stations

Institute of Scientific and Technical Information of China (English)

无

2001-01-01

Thermal management aims at making full use of energy resources available in the space station to reduce energy consumption, waste heat rejection and the weight of the station. It is an extension of the thermal control. This discussion introduces the concept and development of thermal management, presents the aspects of thermal management and further extends its application to subsystems of the space station.

Jet grout strut for deep station boxes of the north/south metro line Amsterdam : Design and back analysis

NARCIS (Netherlands)

Delfgaauw, S.; Buykx, S.M.; Bosch, J.W.

2009-01-01

The new North/South metro line in Amsterdam contains several stations, some of which with an excavation depth of 30 m below surface. The design of these station boxes is very much determined by the adjacent historic buildings, high water table and the relatively soft soil. A lot of effort is put

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

Science.gov (United States)

TANAKA, Shun-ichi

2012-01-01

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

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

International Nuclear Information System (INIS)

Tanaka, Shun-ichi

2012-01-01

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

Analysis Methodology for Optimal Selection of Ground Station Site in Space Missions

Science.gov (United States)

Nieves-Chinchilla, J.; Farjas, M.; Martínez, R.

2013-12-01

Optimization of ground station sites is especially important in complex missions that include several small satellites (clusters or constellations) such as the QB50 project, where one ground station would be able to track several spatial vehicles, even simultaneously. In this regard the design of the communication system has to carefully take into account the ground station site and relevant signal phenomena, depending on the frequency band. To propose the optimal location of the ground station, these aspects become even more relevant to establish a trusted communication link due to the ground segment site in urban areas and/or selection of low orbits for the space segment. In addition, updated cartography with high resolution data of the location and its surroundings help to develop recommendations in the design of its location for spatial vehicles tracking and hence to improve effectiveness. The objectives of this analysis methodology are: completion of cartographic information, modelling the obstacles that hinder communication between the ground and space segment and representation in the generated 3D scene of the degree of impairment in the signal/noise of the phenomena that interferes with communication. The integration of new technologies of geographic data capture, such as 3D Laser Scan, determine that increased optimization of the antenna elevation mask, in its AOS and LOS azimuths along the horizon visible, maximizes visibility time with spatial vehicles. Furthermore, from the three-dimensional cloud of points captured, specific information is selected and, using 3D modeling techniques, the 3D scene of the antenna location site and surroundings is generated. The resulting 3D model evidences nearby obstacles related to the cartographic conditions such as mountain formations and buildings, and any additional obstacles that interfere with the operational quality of the antenna (other antennas and electronic devices that emit or receive in the same bandwidth

Evaluation of Shoreham Nuclear Power Station, Unit 1 technical specifications

International Nuclear Information System (INIS)

Baxter, D.E.; Bruske, S.J.

1985-08-01

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Shoreham Nuclear Power Station Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the assumptions of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Shoreham T/S. Several discrepancies were identified and subsequently resolved through discussions with the cognizant NRC reviewer, NRC staff reviewers and/or utility representatives. The Shoreham Nuclear Power Station Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

Evaluation of Waterford Steam Electric Station Unit 3 technical specifications

International Nuclear Information System (INIS)

Baxter, D.E.; Bruske, S.J.

1985-09-01

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Waterford Steam Electric Station Unit 3 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the requirements of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Waterford T/S. Several discrepancies were identified and subsequently resolved by the cognizant NRC reviewer. Pending completion of the resolutions noted in Part 3 of this report, the Waterford Steam Electric Station Unit 3 T/S, to the extent reviewed, are in conformance with the FSAR and SER

Beaver Valley Power Station and Shippingport Atomic Power Station. 1984 Annual environmental report, radiological. Volume 2

International Nuclear Information System (INIS)

1985-01-01

This report describes the Radiological Environmental Monitoring Program conducted during 1984 in the vicinity of the Beaver Valley Power Station and the Shippingport Atomic Power Station. The Radiological Environmental Program consists of on-site sampling of water and gaseous effluents and off-site monitoring of water, air, river sediments, soils, food pathway samples, and radiation levels in the vicinity of the site. This report discusses the results of this monitoring during 1984. The environmental program outlined in the Beaver Valley Power Station Technical Specifications was followed throughout 1984. The results of this environmental monitoring program show that Shippingport Atomic Power Station and Beaver Valley Power Station operations have not adversely affected the surrounding environment. 23 figs., 18 tabs

Arduino adventures escape from Gemini station

CERN Document Server

Kelly, James Floyd

2013-01-01

Arduino Adventures: Escape from Gemini Station provides a fun introduction to the Arduino microcontroller by putting you (the reader) into the action of a science fiction adventure story.  You'll find yourself following along as Cade and Elle explore Gemini Station-an orbiting museum dedicated to preserving and sharing technology throughout the centuries. Trouble ensues. The station is evacuated, including Cade and Elle's class that was visiting the station on a field trip. Cade and Elle don't make it aboard their shuttle and are trapped on the station along with a friendly artificial intellig

Hydrogen vehicle fueling station

Energy Technology Data Exchange (ETDEWEB)

Daney, D.E.; Edeskuty, F.J.; Daugherty, M.A. [Los Alamos National Lab., NM (United States)] [and others

1995-09-01

Hydrogen fueling stations are an essential element in the practical application of hydrogen as a vehicle fuel, and a number of issues such as safety, efficiency, design, and operating procedures can only be accurately addressed by a practical demonstration. Regardless of whether the vehicle is powered by an internal combustion engine or fuel cell, or whether the vehicle has a liquid or gaseous fuel tank, the fueling station is a critical technology which is the link between the local storage facility and the vehicle. Because most merchant hydrogen delivered in the US today (and in the near future) is in liquid form due to the overall economics of production and delivery, we believe a practical refueling station should be designed to receive liquid. Systems studies confirm this assumption for stations fueling up to about 300 vehicles. Our fueling station, aimed at refueling fleet vehicles, will receive hydrogen as a liquid and dispense it as either liquid, high pressure gas, or low pressure gas. Thus, it can refuel any of the three types of tanks proposed for hydrogen-powered vehicles -- liquid, gaseous, or hydride. The paper discusses the fueling station design. Results of a numerical model of liquid hydrogen vehicle tank filling, with emphasis on no vent filling, are presented to illustrate the usefulness of the model as a design tool. Results of our vehicle performance model illustrate our thesis that it is too early to judge what the preferred method of on-board vehicle fuel storage will be in practice -- thus our decision to accommodate all three methods.

Low earth orbit environmental effects on the space station photovoltaic power generation systems

International Nuclear Information System (INIS)

Nahra, H.K.

1977-01-01

A summary of the Low Earth Orbital Environment, its impact on the photovoltaic power systems of the space station and the solutions implemented to resolve the environmental concerns or issues are described. Low Earth Orbital Environment (LEO) presents several concerns to the photovoltaic power systems of the space station. These concerns include atomic oxygen interaction with the polymeric substrate of the solar arrays, ionized environment effects on the array operating voltage, the effects of the meteoroids and debris impacts and penetration through the different layers of the solar cells and their circuits, and the high energy particle and radiation effects on the overall solar array performance. Potential solutions to some of the degrading environmental interactions that will provide the photovoltaic power system of the space station with the desired life are also summarized

Application for the Tape Station

CERN Document Server

Solero, A

2003-01-01

The Tape Station is used as an Isolde facility to observe the variations of intensity and the lifespan of certain isotopes. A Siemens Simatic FM-352-5 module controls the Tape Station in a PLC system then a DSC controls the PLC, which will be controlled the Tape station program. During the Isolde consolidation project, the Tape Station has been rebuilt, and the control system has been fully integrated in the PS control. Finally, a new application has been written in JAVA Development kit 1.4 and the PS Java environment. The main purpose of this note is to explain how to use this program.

Beaver Valley Power Station and Shippingport Atomic Power Station. 1977 annual environmental report: radiological. Volume 2

International Nuclear Information System (INIS)

1978-01-01

The environmental monitoring conducted during 1977 in the vicinity of the Beaver Valley Power Station and the Shippingport Atomic Power Station is described. The environmental monitoring program consists of onsite sampling of water, gaseous, and air effluents, as well as offsite monitoring of water, air, river sediments, and radiation levels in the vicinity of the site. The report discusses releases of small quantities of radioactivity to the Ohio River from the Beaver Valley Power Station and Shippingport Atomic Power Station during 1977

Work/control stations in Space Station weightlessness

Science.gov (United States)

Willits, Charles

1990-01-01

An ergonomic integration of controls, displays, and associated interfaces with an operator, whose body geometry and dynamics may be altered by the state of weightlessness, is noted to rank in importance with the optimal positioning of controls relative to the layout and architecture of 'body-ported' work/control stations applicable to the NASA Space Station Freedom. A long-term solution to this complex design problem is envisioned to encompass the following features: multiple imaging, virtual optics, screen displays controlled by a keyboard ergonomically designed for weightlessness, cursor control, a CCTV camera, and a hand-controller featuring 'no-grip' vernier/tactile positioning. This controller frees all fingers for multiple-switch actuations, while retaining index/register determination with the hand controller. A single architectural point attachment/restraint may be used which requires no residual muscle tension in either brief or prolonged operation.

Hydrogen Fuelling Stations

DEFF Research Database (Denmark)

Rothuizen, Erasmus Damgaard

. A system consisting of one high pressure storage tank is used to investigate the thermodynamics of fuelling a hydrogen vehicle. The results show that the decisive parameter for how the fuelling proceeds is the pressure loss in the vehicle. The single tank fuelling system is compared to a cascade fuelling......This thesis concerns hydrogen fuelling stations from an overall system perspective. The study investigates thermodynamics and energy consumption of hydrogen fuelling stations for fuelling vehicles for personal transportation. For the study a library concerning the components in a hydrogen fuelling...... station has been developed in Dymola. The models include the fuelling protocol (J2601) for hydrogen vehicles made by Society of Automotive Engineers (SAE) and the thermodynamic property library CoolProp is used for retrieving state point. The components in the hydrogen fuelling library are building up...

Establishing a mobile automatic monitoring station for emergency response

International Nuclear Information System (INIS)

Fang, Hsin-Fa

2008-01-01

Full text: A radiological/nuclear emergency event may cause environmental contamination. The emergency response works always need to plan an environmental survey programme incorporating the assessment results to see what is happening. The places where are assessed to have the highest radioactive contamination/radiation dose will catch more concern and need continuous monitoring. It will cause unnecessary dangers and dose to command that personnel conduct surveying in such places when the radiological/nuclear accident become more severe. A mobile automatic monitoring station has been established for emergency response by INER (Institute of Nuclear Energy Research) to solve the problem practically. The monitoring station involves a HPIC to monitor radiation dose, an anemometer to monitor wind speed and direction, a GPS to get position data, a GPRS/3G communication module to send monitoring and positioning data to the monitoring centre where can show the monitoring result directly on a map shown on the computer. These instruments are integrated in a trailer easy to be towed to the place need to be monitored. The electric power of the station is supplied by s a solar power energy system. It can supply the station working at least 10 days without extra electric power supply designed based on the expected time length of a nuclear power plant event. The HPIC is very sensitive and stable that can discriminate a 10 nSv/hr increasing of dose rate with the monitoring time period every ten seconds. Where the radiological dispersion device events happened is not predictable, it is difficult to get suitable wind monitoring data to assess the result of radiological dispersion device events. The anemometer added on the station can provide the real time wind monitoring data to help assessment works. (author)

Tobruk power station

Energy Technology Data Exchange (ETDEWEB)

Boergardts, B

1978-01-01

In February of 1975, the Electricity Corporation Benghazi (ECB) awarded a contract for the construction of a turnkey power station and seawater desalination plant in Tobruk, Libya to a consortium under the leadership of BBC Mannheim. This power station has an output of 129 MW and supplies about 24,000 m/sup 3/ of drinking water daily. It went into operation in 1977, two and a half years after the contract was awarded.

National Seismic Station

International Nuclear Information System (INIS)

Stokes, P.A.

1982-06-01

The National Seismic Station was developed to meet the needs of regional or worldwide seismic monitoring of underground nuclear explosions to verify compliance with a nuclear test ban treaty. The Station acquires broadband seismic data and transmits it via satellite to a data center. It is capable of unattended operation for periods of at least a year, and will detect any tampering that could result in the transmission of unauthentic seismic data

Effect of mobile phone station on micronucleus frequency and chromosomal aberrations in human blood cells.

Science.gov (United States)

Yildirim, M S; Yildirim, A; Zamani, A G; Okudan, N

2010-01-01

The use of mobile telephones has rapidly increased worldwide as well as the number of mobile phone base stations that lead to rise low level radiofrequency emissions which may in turn have possible harm for human health. The national radiation protection board has published the known effects of radio waves exposure on humans living close to mobile phone base stations. However, several studies have claimed that the base station has detrimental effects on different tissues. In this study, we aimed to evaluate the effects of mobile phone base stations on the micronucleus (MN) frequency and chromosomal aberrations on blood in people who were living around mobile phone base stations and healthy controls. Frequency of MN and chromosomal aberrations in study and control groups was 8.96 +/- 3.51 and 6.97 +/- 1.52 (p: 0.16); 0.36 +/- 0.31 and 0.75 +/- 0.61 (p: 0.07), respectively. Our results show that there was not a significant difference of MN frequency and chromosomal aberrations between the two study groups. The results claim that cellular phones and their base stations do not produce important carcinogenic changes.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Passive system with steam-water injector for emergency supply of NPP steam generators

International Nuclear Information System (INIS)

Il'chenko, A.G.; Strakhov, A.N.; Magnitskij, D.N.

2009-01-01

The calculation results of reliability indicators of emergency power supply system and emergency feed-water supply system of serial WWER-1000 unit are presented. To ensure safe water supply to steam generators during station blackout it was suggested using additional passive emergency feed-water system with a steam-water injector working on steam generators dump steam. Calculated analysis of steam-water injector operating capacity was conducted at variable parameters of steam at the entrance to injector, corresponding to various moments of time from the beginning of steam-and-water damping [ru

A reliability model of the Angra 1 power system by the device of stages optimized by genetic algorithms

International Nuclear Information System (INIS)

Crossetti, Patricia Guimaraes

2006-01-01

This thesis proposes a probabilistic model to perform the reliability analysis of nuclear power plant systems under aging. This work analyses the Angra 1 power system. Systems subject to aging consist of components whose failure rates are not all constant, thus generating Non-Markovian models. Genetic algorithms were used for optimizing the application of the device of stages. Two approaches were used in the optimization, MCEF and MCEV. The results obtained for the Angra 1 power system show that the probability of a station blackout is negligible. (author)

Shippingport Station Decommissioning Project

International Nuclear Information System (INIS)

McKernan, M.L.

1989-01-01

The Shippingport Atomic Power Station was located on the Ohio River in Shippingport Borough (Beaver County), Pennsylvania, USA. The US Atomic Energy Commission (AEC) constructed the plant in the mid-1950s on a seven and half acre parcel of land leased from Duquesne Light Company (DLC). The purposes were to demonstrate and to develop Pressurized Water Recovery technology and to generate electricity. DLC operated the Shippingport plant under supervision of (the successor to AEC) the Department of Energy (DOE)-Naval Reactors (NR) until operations were terminated on October 1, 1982. NR concluded end-of-life testing and defueling in 1984 and transferred the Station's responsibility to DOE Richland Operations Office (RL), Surplus Facility Management Program Office (SFMPO5) on September 5, 1984. SFMPO subsequently established the Shippingport Station Decommissioning Project and selected General Electric (GE) as the Decommissioning Operations Contractor. This report is intended to provide an overview of the Shippingport Station Decommissioning Project

20 years of power station master training

International Nuclear Information System (INIS)

Schwarz, O.

1977-01-01

In the early fifties, the VGB working group 'Power station master training' elaborated plans for systematic and uniform training of power station operating personnel. In 1957, the first power station master course was held. In the meantime, 1.720 power station masters are in possession of a master's certificate of a chamber of commerce and trade. Furthermore, 53 power station masters have recently obtained in courses of the 'Kraftwerksschule e.V.' the know-how which enables them to also carry out their duty as a master in nuclear power stations. (orig.) [de

A report on upgraded seismic monitoring stations in Myanmar: Station performance and site response

Science.gov (United States)

Thiam, Hrin Nei; Min Htwe, Yin Myo; Kyaw, Tun Lin; Tun, Pa Pa; Min, Zaw; Htwe, Sun Hninn; Aung, Tin Myo; Lin, Kyaw Kyaw; Aung, Myat Min; De Cristofaro, Jason; Franke, Mathias; Radman, Stefan; Lepiten, Elouie; Wolin, Emily; Hough, Susan E.

2017-01-01

Myanmar is in a tectonically complex region between the eastern edge of the Himalayan collision zone and the northern end of the Sunda megathrust. Until recently, earthquake monitoring and research efforts have been hampered by a lack of modern instrumentation and communication infrastructure. In January 2016, a major upgrade of the Myanmar National Seismic Network (MNSN; network code MM) was undertaken to improve earthquake monitoring capability. We installed five permanent broadband and strong‐motion seismic stations and real‐time data telemetry using newly improved cellular networks. Data are telemetered to the MNSN hub in Nay Pyi Taw and archived at the Incorporated Research Institutions for Seismology Data Management Center. We analyzed station noise characteristics and site response using noise and events recorded over the first six months of station operation. Background noise characteristics vary across the array, but indicate that the new stations are performing well. MM stations recorded more than 20 earthquakes of M≥4.5 within Myanmar and its immediate surroundings, including an M 6.8 earthquake located northwest of Mandalay on 13 April 2016 and the Mw 6.8 Chauk event on 24 August 2016. We use this new dataset to calculate horizontal‐to‐vertical spectral ratios, which provide a preliminary characterization of site response of the upgraded MM stations.

EUPOS - Satellite multifunctional system of reference stations in Central and Eastern Europe

Science.gov (United States)

Sledzinski, J.

2003-04-01

The European project EUPOS (European Position Determination System) of establishment of a system of multifunctional satellite reference stations in Central and Eastern Europe is described in the paper. Fifteen countries intend to participate in the project: Bulgaria, Croatia, Czech Republic, Estonia, Germany, Hungary, Latvia, Lithuania, Macedonia, Poland, Romania, Russia, Serbia, Slovak Republic and Slovenia. One common project will be prepared for all countries, however it will include the existing or developed infrastructure in particular countries. The experiences of establishing and operating of the German network SAPOS as well as experiences gained by other countries will be used. The European network of stations will be compatible with the system SAPOS and future European system Galileo. The network of reference stations will provide signal for both positioning of the geodetic control points and for land, air and marine navigation. Several levels of positioning accuracy will be delivered.

Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

Directory of Open Access Journals (Sweden)

Siniša Šadek

2010-01-01

Full Text Available RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

Application of a quantitative histological health index for Antarctic rock cod (Trematomus bernacchii) from Davis Station, East Antarctica.

Science.gov (United States)

Corbett, Patricia A; King, Catherine K; Mondon, Julie A

2015-08-01

A quantitative Histological Health Index (HHI) was applied to Antarctic rock cod (Trematomus bernacchii) using gill, liver, spleen, kidney and gonad to assess the impact of wastewater effluent from Davis Station, East Antarctica. A total of 120 fish were collected from 6 sites in the Prydz Bay region of East Antarctica at varying distances from the wastewater outfall. The HHI revealed a greater severity of alteration in fish at the wastewater outfall, which decreased stepwise with distance. Gill and liver displayed the greatest severity of alteration in fish occurring in close proximity to the wastewater outfall, showing severe and pronounced alteration respectively. Findings of the HHI add to a growing weight of evidence indicating that the current level of wastewater treatment at Davis Station is insufficient to prevent impact to the surrounding environment. The HHI for T. bernacchii developed in this study is recommended as a useful risk assessment tool for assessing in situ, sub-lethal impacts from station-derived contamination in coastal regions throughout Antarctica. Copyright © 2015 Elsevier Ltd. All rights reserved.

47 CFR 22.313 - Station identification.

Science.gov (United States)

2010-10-01

... Telephone Radio Systems in the Rural Radiotelephone Service; (5) [Reserved] (6) Stations operating pursuant... Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) COMMON CARRIER SERVICES PUBLIC MOBILE SERVICES... of each station in the Public Mobile Services must ensure that the transmissions of that station are...

Space station operations management

Science.gov (United States)

Cannon, Kathleen V.

1989-01-01

Space Station Freedom operations management concepts must be responsive to the unique challenges presented by the permanently manned international laboratory. Space Station Freedom will be assembled over a three year period where the operational environment will change as significant capability plateaus are reached. First Element Launch, Man-Tended Capability, and Permanent Manned Capability, represent milestones in operational capability that is increasing toward mature operations capability. Operations management concepts are being developed to accomodate the varying operational capabilities during assembly, as well as the mature operational environment. This paper describes operations management concepts designed to accomodate the uniqueness of Space Station Freedoom, utilizing tools and processes that seek to control operations costs.

Locations of Sampling Stations for Water Quality Monitoring in Water Distribution Networks.

Science.gov (United States)

Rathi, Shweta; Gupta, Rajesh

2014-04-01

Water quality is required to be monitored in the water distribution networks (WDNs) at salient locations to assure the safe quality of water supplied to the consumers. Such monitoring stations (MSs) provide warning against any accidental contaminations. Various objectives like demand coverage, time for detection, volume of water contaminated before detection, extent of contamination, expected population affected prior to detection, detection likelihood and others, have been independently or jointly considered in determining optimal number and location of MSs in WDNs. "Demand coverage" defined as the percentage of network demand monitored by a particular monitoring station is a simple measure to locate MSs. Several methods based on formulation of coverage matrix using pre-specified coverage criteria and optimization have been suggested. Coverage criteria is defined as some minimum percentage of total flow received at the monitoring stations that passed through any upstream node included then as covered node of the monitoring station. Number of monitoring stations increases with the increase in the value of coverage criteria. Thus, the design of monitoring station becomes subjective. A simple methodology is proposed herein which priority wise iteratively selects MSs to achieve targeted demand coverage. The proposed methodology provided the same number and location of MSs for illustrative network as an optimization method did. Further, the proposed method is simple and avoids subjectivity that could arise from the consideration of coverage criteria. The application of methodology is also shown on a WDN of Dharampeth zone (Nagpur city WDN in Maharashtra, India) having 285 nodes and 367 pipes.

Ondergronds Station Blijdorp, Rotterdam

NARCIS (Netherlands)

Hijma, M.P.|info:eu-repo/dai/nl/266562426; Cohen, K.M.|info:eu-repo/dai/nl/185633374

2014-01-01

Het is in de herfst van 2005. Een lief meisje, Marieke, rijdt op haar vouwfiets door Rotterdam. Bij het Centraal Station is het al tijden een grote bouwplaats. Onder de nieuwe hal komt een veel groter metrostation en ook onder de Statenweg in Blijdorp is een grote bouwput voor een nieuw station.

Single spectrometer station for neutrino-tagging

International Nuclear Information System (INIS)

Nedyalkov, I.P.

1984-01-01

A neutrino tagging station built with respect to the following scheme is proposed. A beam of muons and kaons passes through a magnetic spectrometer, where the energy of each particle is measured. There are coordinate detectors behind the spectrometer in several planes, where the direction of the trajectory of a given particle is determined. Thus, mesons enter the decay point wth the known 4-momentum. Behind the decay point the direction of μ-meson generated by the decay of parent mesons is measured. It is shown that information is sufficient for determining the kind of parent particle (pion or kaon), the energy and the direction of trajectory of the neutrino