Analysis methodology for the post-trip return to power steam line break event

Energy Technology Data Exchange (ETDEWEB)

Lee, Chul Shin; Kim, Chul Woo; You, Hyung Keun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

An analysis for Steam Line Break (SLB) events which result in a Return-to-Power (RTP) condition after reactor trip was performed for a postulated Yonggwang Nuclear Power Plant Unit 3 cycle 8. Analysis methodology for post-trip RTP SLB is quite different from that of non-RTP SLB and is more difficult. Therefore, it is necessary to develop a methodology to analyze the response of the NSSS parameters to the post-trip RTP SLB events and the fuel performance after the total reactivity exceeds the criticality. In this analysis, the cases with and without offsite power were simulated crediting 3-D reactivity feedback effect due to a local heatup in the vicinity of stuck CEA and compared with the cases without 3-D reactivity feedback with respect to post-trip fuel performance. Departure-to Nucleate Boiling Ratio (DNBR) and Linear Heat Generation Rate (LHGR). 36 tabs., 32 figs., 11 refs. (Author) .new.

Benchmark analysis of three main circulation pump sequential trip event at Ignalina NPP

International Nuclear Information System (INIS)

Uspuras, E.; Kaliatka, A.; Urbonas, R.

2001-01-01

The Ignalina Nuclear Power Plant is a twin-unit with two RBMK-1500 reactors. The primary circuit consists of two symmetrical loops. Eight Main Circulation Pumps (MCPs) at the Ignalina NPP are employed for the coolant water forced circulation through the reactor core. The MCPs are joined in groups of four pumps for each loop (three for normal operation and one on standby). This paper presents the benchmark analysis of three main circulation pump sequential trip event at RBMK-1500 using RELAP5 code. During this event all three MCPs in one circulation loop at Unit 2 Ignalina NPP were tripped one after another, because of inadvertent activation of the fire protection system. The comparison of calculated and measured parameters led us to establish realistic thermal hydraulic characteristics of different main circulation circuit components and to verify the model of drum separators pressure and water level controllers.(author)

Preliminary analysis of beam trip and beam jump events in an ADS prototype

International Nuclear Information System (INIS)

D'Angelo, A.; Bianchini, G.; Carta, M.

2001-01-01

A core dynamics analysis relevant to some typical current transient events has been carried out on an 80 MW energy amplifier prototype (EAP) fuelled by mixed oxides and cooled by lead-bismuth. Fuel and coolant temperature trends relevant to recovered beam trip and beam jump events have been preliminary investigated. Beam trip results show that the drop in temperature of the core outlet coolant would be reduced a fair amount if the beam intensity could be recovered within few seconds. Due to the low power density in the EAP fuel, the beam jump from 50% of the nominal power transient evolves benignly. The worst thinkable current transient, beam jump with cold reactor, mainly depends on the coolant flow conditions. In the EAP design, the primary loop coolant flow is assured by natural convection and is enhanced by a particular system of cover gas injection into the bottom part of the riser. If this system of coolant flow enhancement is assumed in function, even the beam jump with cold reactor event evolves without severe consequences. (authors)

Power supply trip control for nuclear reactor

International Nuclear Information System (INIS)

Hager, R.E.; Gutman, Jerzy.

1987-01-01

A control system for a trip coil in a switchgear mechanism controls the supply of electrical power to a process control device and ensures de-energization of the trip coil shortly after the trip coil is energized. The trip coil is energized not by an independent dc source as in prior art, but from rectified power from a step down transformer supplied from the switchgear output side. The transformer feeds a rectifier which is connected to the trip coil via a trip activation device. The output of the rectifier can be monitored using an optical converter to determine the ability of the control system to activate the trip coil and the condition of the power supplied to the process control device. The control device may be a rod positioner in a pressurised water nuclear reactor. (author)

A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

Energy Technology Data Exchange (ETDEWEB)

Alves, Antonio Carlos Pinto Dias

1993-12-31

The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

Energy Technology Data Exchange (ETDEWEB)

Alves, Antonio Carlos Pinto Dias

1994-12-31

The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

Boiling water reactor turbine trip (TT) benchmark

International Nuclear Information System (INIS)

2001-06-01

In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as for current nuclear applications Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for the coupling of core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for the purpose. The present volume describes the specification of such a benchmark. The transient addressed is a turbine trip (TT) in a BWR involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the plant make the present benchmark very valuable. The data used are from events at the Peach Bottom 2 reactor (a GE-designed BWR/4). (authors)

Realistic thermal transient margin analysis of 'MONJU' based on plant performance measurements. Reactor vessel outlet nozzle and evaporator feed water inlet tube sheet of the manual reactor plant trip

International Nuclear Information System (INIS)

Yamada, Fumiaki; Mori, Takero

2005-01-01

In order to develop technologies and achieve safe and stable operation of Monju' as well as realize optimized design and construction of safe and economically competitive fast breeder reactors, the authors are evaluating design approach applied to 'Monju' based on actually measured behavioral data during plant operations. This report uses actual measured characteristic data of 'Monju' during a plant trip test obtained at a commissioning stage with up to 40% power output and introduces plant thermal hydraulic behavior analysis in a representative thermal transient event, i.e. a manual plant trip. Thermal transient driven loads incurred by the reactor vessel outlet nozzle and by the evaporator feed water inlet tube sheet were further derived by structural analyses and were compared with the previously derived values in the design stage and with the limit values. Though the reactor vessel outlet nozzle was exposed to larger temperature change in the trip test than the analytical prediction, the newly calculated mechanical load was about 50% of the previous evaluation in the design stage. Also, the newly analyzed mechanical load incurred by the evaporator feed water inlet tube sheet in this event had a large margin against the limit value of cumulative damage cycle fraction, although the observed temperature disturbance in a steam blow test was wilder than the analytical prediction. Thus we concluded that the Monju' plant has an assured safety margin against thermal transient in plant trip events. (author)

Report on safety related occurrences and reactor trips July 1, 1979 - December 31, 1979

International Nuclear Information System (INIS)

Olsson, S.; Andermo, L.

1980-01-01

This is a report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1979 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 76 safety related occurrences and 27 reactor trips have been reported to the Nuclear Power Inspectorate. It is to the greatest extent conventional components such as valves and pumps which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant system and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips are normal. The average value for these 6 months is 4.5 trips/unit. Approximetely one half of the reactor trips happened at zero or very low power operation. The fact that even small deviations from prescribed operation result in an automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences without safety significance. (author)

Boiling water reactor turbine trip (TT) benchmark

International Nuclear Information System (INIS)

2005-01-01

In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts as well as for current applications. Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose. The present report is the second in a series of four and summarises the results of the first benchmark exercise, which identifies the key parameters and important issues concerning the thermalhydraulic system modelling of the transient, with specified core average axial power distribution and fission power time transient history. The transient addressed is a turbine trip in a boiling water reactor, involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the Peach Bottom 2 reactor (a GE-designed BWR/4) make the present benchmark particularly valuable. (author)

Control rod trip failures; Salem 1, the cause, response, and potential fixes

International Nuclear Information System (INIS)

Hall, R.E.; Boccio, J.L.; Luckas, W.J.

1984-01-01

This chapter presents a systems and reliability analysis of recent nuclear reactor control rod failure-to-trip (or scram) events that have been experienced in the US commercial nuclear industry. The operational factors of hardware, procedures, and human error are considered in the analysis of transients without scram. The 1980 Browns Ferry 3 scram system failure is analyzed to contrast the two 1983 Salem 1 events. The details of the Salem control rod failure to trip are investigated and used to calculate the reactor protection system unavailabilities. The internal reactor trip breaker logic is reviewed as related to the Westinghouse DB-50 breaker application. The impact of test and maintenance on system challenges is discussed. It is concluded that although the failure to trip or scram represents a single class of initiators, the actual events of each transient are operationally unique and require individual human responses

Report on safety related occurrences and reactor trips July 1, 1977 - December 31, 1977

International Nuclear Information System (INIS)

Andermo, L.; Sundman, B.

1974-04-01

This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1977 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 48 safety related occurrences and 49 reactor trips have been reported to the Nuclear Power Inspectorate. Included is also one incident June 21 in Barsebaeck 2 which was not included in the last compilation of occurrences. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips have increased nearly 30% since the last period. Those occurred during power operation however, were less. More than 50% of the reactor trips happened in the shutdown condition. The fact that even small deviations from prescribed operation result in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences withou02068NRM 0000169 450

Report on safety related occurrences and reactor trips January 1 - June 30, 1984

International Nuclear Information System (INIS)

1984-01-01

This is a systematically arranged report on all safety-related occurrences and reactor trips in Swedish nuclear power plants in operation during the period from January 1 to June 30 1984. It is based on the reports submitted by the utilities to the Swedish Nuclear Inspectorate according to Technical Specifications. Twice a year since 1974 the Inspectorate has issued a compilation on such reported occurrences and reactor trips. Starting with the compilation of the second half of 1982 some new features have been introduced. The most important change is that the volume of information has been increased. The full text, provided by the utilities when reporting the incidents, is now attached to the codified information and also the layout has been altered to facilitate reading. As in the previous reports the occurrences and reactor trips are arranged both alphabetically by facility name and chronologically by report number for each facility. Electricity generation charts for each facility are also presented. The primary purpose of this report is thus to present all the information furnished by the utilities when they submit their reports according to Technical Specifications. The only evaluation made by the Inspectorate is the categorization on the incidents. Like the previous reports this one also presents frequency of incidents as related to affected component, cause of incident etc. The difference is that only information reported by the utilities is used. This is the reason why a considerable proportion of the incidents are categorized as other component or other fault. Sometime in the future, however, the Inspectorate plants to put out a special report containing its own analyses of the most interesting events along with processed statistics and other information. (author)

Development of field programmable gate array-based reactor trip functions using systems engineering approach

Energy Technology Data Exchange (ETDEWEB)

Jung, Jae Cheon; Ahmed, Ibrahim [Nuclear Power Plant Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-08-15

Design engineering process for field programmable gate array (FPGA)-based reactor trip functions are developed in this work. The process discussed in this work is based on the systems engineering approach. The overall design process is effectively implemented by combining with design and implementation processes. It transforms its overall development process from traditional V-model to Y-model. This approach gives the benefit of concurrent engineering of design work with software implementation. As a result, it reduces development time and effort. The design engineering process consisted of five activities, which are performed and discussed: needs/systems analysis; requirement analysis; functional analysis; design synthesis; and design verification and validation. Those activities are used to develop FPGA-based reactor bistable trip functions that trigger reactor trip when the process input value exceeds the setpoint. To implement design synthesis effectively, a model-based design technique is implied. The finite-state machine with data path structural modeling technique together with very high speed integrated circuit hardware description language and the Aldec Active-HDL tool are used to design, model, and verify the reactor bistable trip functions for nuclear power plants.

Abnormal Events for Emergency Trip in HANARO

Energy Technology Data Exchange (ETDEWEB)

Ahn, Guk Hun; Choi, M. J.; Park, S. I.; Kim, H. W.; Kim, S. J.; Park, J. H.; Kwon, I. C

2006-12-15

This report gathers abnormal events related to emergency trip of HANARO that happened during its operation over 10 years since the first criticality on February 1995. The collected examples will be utilized to the HANARO's operators as a useful guide.

Report on safety related occurrences and reactor trips July 1, 1976-December 31, 1976

International Nuclear Information System (INIS)

Andermo, L.

1977-04-01

This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1, 1976 to December 31, 1976 inclusive. The facilities involved are Oskarshamn 1 and 2, Ringhals 1 and 2 and Barsebaeck 1. During this period of the 6 months 37 safety related occurrences and 34 reactor trips have been reported to the Nuclear Power Inspectorate. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The fact that even small deviations from prescribed operation results in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The number of reactor trips are almost as low as during the last period, which is a drastic reduction compared to earlier time periods. The greatest outages are caused by occurrences without safety significance.(author)

Microprocessor tester for the treat upgrade reactor trip system

International Nuclear Information System (INIS)

Lenkszus, F.R.; Bucher, R.G.

1984-01-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations

Reevaluation of steam generator level trip set point

Energy Technology Data Exchange (ETDEWEB)

Shim, Yoon Sub; Soh, Dong Sub; Kim, Sung Oh; Jung, Se Won; Sung, Kang Sik; Lee, Joon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

The reactor trip by the low level of steam generator water accounts for a substantial portion of reactor scrams in a nuclear plant and the feasibility of modification of the steam generator water level trip system of YGN 1/2 was evaluated in this study. The study revealed removal of the reactor trip function from the SG water level trip system is not possible because of plant safety but relaxation of the trip set point by 9 % is feasible. The set point relaxation requires drilling of new holes for level measurement to operating steam generators. Characteristics of negative neutron flux rate trip and reactor trip were also reviewed as an additional work. Since the purpose of the trip system modification for reduction of a reactor scram frequency is not to satisfy legal requirements but to improve plant performance and the modification yields positive and negative aspects, the decision of actual modification needs to be made based on the results of this study and also the policy of a plant owner. 37 figs, 6 tabs, 14 refs. (Author).

Uncertainty analysis of one Main Circulation Pump trip event at the Ignalina NPP

International Nuclear Information System (INIS)

Vileiniskis, V.; Kaliatka, A.; Uspuras, E.

2004-01-01

One Main Circulation Pump (MCP) trip event is an anticipated transient with expected frequency of approximately one event per year. There were a few events when one MCP was inadvertently tripped. The throughput of the rest running pumps in the affected Main Circulation Circuit loop increased, however, the total coolant flow through the affected loop decreased. The main question arises whether this coolant flow rate is sufficient for adequate core cooling. This paper presents an investigation of one MCP trip event at the Ignalina NPP. According to international practice, the transient analysis should consist of deterministic analysis by employing best-estimate codes and uncertainty analysis. For that purpose, the plant's RELAP5 model and the GRS (Germany) System for Uncertainty and Sensitivity Analysis package (SUSA) were employed. Uncertainty analysis of flow energy loss in different parts of the Main Circulation Circuit, initial conditions and code-selected models was performed. Such analysis allows to estimate the influence of separate parameters on calculation results and to find the modelling parameters that have the largest impact on the event studied. On the basis of this analysis, recommendations for the further improvement of the model have been developed. (author)

Technical evaluation of the proposed deletion of a reactor trip on a turbine trip below 50-percent power for the Beaver Valley nuclear power plant, Unit 1

International Nuclear Information System (INIS)

Reeves, W.E.

1979-12-01

This report documents the technical evaluation of the Duquesne Light Company's proposed license amendment for the deletion of a reactor trip on a turbine trip below 50% power for the Beaver Valley nuclear power plant, Unit 1. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Program being conducted for the US Nuclear Regulatory Commission by Lawrence Livermore Laboratory

Probabilistic methods in a study of trip setpoints

International Nuclear Information System (INIS)

Kaulitz, D. E.

2012-01-01

Most early vintage Boiling Water Reactors have a high head and high capacity High Pressure Coolant Injection (HPCI) pump to keep the core covered following a loss of coolant accident (LOCA). However, the protection afforded by the HPCI pump for mitigating a LOCA introduces the potential that a spurious start of the HPCI pump could oversupply the reactor vessel and lead to an automatic trip of the main turbine due to high water level. A turbine trip and associated increase in moderator density could challenge the bases of fuel integrity operating limits. To prevent turbine trip during spurious operation of the HPCI pump, the reactor protection system includes instrumentation and logic to sense high water level and automatically trip the HPCI pump prior to reaching the turbine trip setpoint. This paper describes an analysis that was performed to determine if existing reactor vessel water level trip instrumentation, logic and setpoints result in a high probability that the HPCI pump will trip prior to actuation of the turbine trip. Using nominal values for the initial water level and for the HPCI pump and turbine trip setpoints, and using the probability distribution functions for measurement uncertainty in these setpoints, a Monte Carlo simulation was employed to determine probabilities of successfully tripping the HPCI pump prior to tripping of the turbine. The results of the analysis established that the existing setpoints, instrumentation and logic would be expected to reliably prevent a trip of the main turbine. (authors)

Initiating Event Rates at U.S. Nuclear Power Plants. 1988 - 2013

International Nuclear Information System (INIS)

Schroeder, John A.; Bower, Gordon R.

2014-01-01

Analyzing initiating event rates is important because it indicates performance among plants and also provides inputs to several U.S. Nuclear Regulatory Commission (NRC) risk-informed regulatory activities. This report presents an analysis of initiating event frequencies at U.S. commercial nuclear power plants since each plant's low-power license date. The evaluation is based on the operating experience from fiscal year 1988 through 2013 as reported in licensee event reports. Engineers with nuclear power plant experience staff reviewed each event report since the last update to this report for the presence of valid scrams or reactor trips at power. To be included in the study, an event had to meet all of the following criteria: includes an unplanned reactor trip (not a scheduled reactor trip on the daily operations schedule), sequence of events starts when reactor is critical and at or above the point of adding heat, occurs at a U.S. commercial nuclear power plant (excluding Fort St. Vrain and LaCrosse), and is reported by a licensee event report. This report displays occurrence rates (baseline frequencies) for the categories of initiating events that contribute to the NRC's Industry Trends Program. Sixteen initiating event groupings are trended and displayed. Initiators are plotted separately for initiating events with different occurrence rates for boiling water reactors and pressurized water reactors. p-values are given for the possible presence of a trend over the most recent 10 years.

Installation of a second trip system

International Nuclear Information System (INIS)

Bessada, E.

1997-01-01

Since its first criticality in 1957, the NRU reactor has been operating safely and efficiently supporting the CANDU reactor's research and development programs and producing radioisotopes for medical use. To ensure that the reactor continues to operate safely and effectively, Atomic Energy of Canada Limited (AECL) commissioned a team in 1989 to conduct a systematic review and assessment of the reactor condition. The outcome of the study indicated that the overall condition of the reactor is good and that it is being operated safely. The study also produced recommendations as to where safety can be improved. These recommendations are the basis of the upgrade program currently being implemented in the reactor. The Second Trip System (STS) is part of the upgrade program. It is a stand alone seismically qualified trip system that operates independently from the existing first trip system (FST) to shutdown the reactor. This paper discusses the design, installation and the inactive commissioning of the system, and the process used to ensure that the system can be retrofitted to the reactor without affecting its safety or its operational requirements. (author)

Expert system for the CPCS-initiated trip analysis

International Nuclear Information System (INIS)

Sohn, Sedo; Im, Inyoung; Kuh, Jungeui

1991-01-01

In Yonggwang nuclear units 3 and 4, the core protection calculator system (CPCS) performs various protection logics against many transients and certain accidents. The CPCS is a real-time computer system calculating the departure from nucleate boiling ratio (DNBR), and local power density, and other protection logics. It takes process variables such as neutron flux, hot-leg temperature, cold-leg temperature, control element assembly positions, and reactor coolant pump shaft speed. Since the CPCS protection logics are quite complex, it is difficult for an operator to tell immediately which parameter is the major cause of the reactor trip. Thus, whenever the reactor trip signal is generated, the process input variables and calculated results, including selected intermediate variables, are frozen in the specified computer memory for later analysis. These frozen variables are called the trip buffer. Analysis of the trip buffer requires an expert in the CPCS and related documents containing algorithms and a data base for algorithms. The Trip Buffer Analysis Program (TBAP) is an expert system that pinpoints the causes of the CPCS initiated reactor trip, thus relieving the operator from the burden of analyzing the trip buffer

Single Point Vulnerability Analysis of Automatic Seismic Trip System

International Nuclear Information System (INIS)

Oh, Seo Bin; Chung, Soon Il; Lee, Yong Suk; Choi, Byung Pil

2016-01-01

Single Point Vulnerability (SPV) analysis is a process used to identify individual equipment whose failure alone will result in a reactor trip, turbine generator failure, or power reduction of more than 50%. Automatic Seismic Trip System (ASTS) is a newly installed system to ensure the safety of plant when earthquake occurs. Since this system directly shuts down the reactor, the failure or malfunction of its system component can cause a reactor trip more frequently than other systems. Therefore, an SPV analysis of ASTS is necessary to maintain its essential performance. To analyze SPV for ASTS, failure mode and effect analysis (FMEA) and fault tree analysis (FTA) was performed. In this study, FMEA and FTA methods were performed to select SPV equipment of ASTS. D/O, D/I, A/I card, seismic sensor, and trip relay had an effect on the reactor trip but their single failure will not cause reactor trip. In conclusion, ASTS is excluded as SPV. These results can be utilized as the basis data for ways to enhance facility reliability such as design modification and improvement of preventive maintenance procedure

Single Point Vulnerability Analysis of Automatic Seismic Trip System

Energy Technology Data Exchange (ETDEWEB)

Oh, Seo Bin; Chung, Soon Il; Lee, Yong Suk [FNC Technology Co., Yongin (Korea, Republic of); Choi, Byung Pil [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

Single Point Vulnerability (SPV) analysis is a process used to identify individual equipment whose failure alone will result in a reactor trip, turbine generator failure, or power reduction of more than 50%. Automatic Seismic Trip System (ASTS) is a newly installed system to ensure the safety of plant when earthquake occurs. Since this system directly shuts down the reactor, the failure or malfunction of its system component can cause a reactor trip more frequently than other systems. Therefore, an SPV analysis of ASTS is necessary to maintain its essential performance. To analyze SPV for ASTS, failure mode and effect analysis (FMEA) and fault tree analysis (FTA) was performed. In this study, FMEA and FTA methods were performed to select SPV equipment of ASTS. D/O, D/I, A/I card, seismic sensor, and trip relay had an effect on the reactor trip but their single failure will not cause reactor trip. In conclusion, ASTS is excluded as SPV. These results can be utilized as the basis data for ways to enhance facility reliability such as design modification and improvement of preventive maintenance procedure.

Analyses of anticipated transient without scram events in SMART

International Nuclear Information System (INIS)

Kim, Hyung Rae; Chun, Ji Han; Kim, Soo Hyoung; Yang, Soo Hyung; Bae, Kyoo Hwan

2012-01-01

SMART is a small integral reactor, which was developed at KAERI and acquired standard design approval in 2012. SMART works like a pressurized light water reactor in principle though it is more compact than loop type large commercial reactors. ATWS(Anticipated Transient Without Scram) event is an AOO(Anticipated Operational Occurrence) where RPS fails to trip the reactor when requested. SMART incorporated a DPS(diverse protection system) to protect the reactor system when RPS(reactor protection system) fails to trip the reactor. The results of transient analyses show that DPS in SMART effectively mitigates the consequence of ATWS

Report on safety related occurrences and reactor trips January 1 - June 30, 1985

International Nuclear Information System (INIS)

1986-01-01

This is a systematically arranged report on all safety-related occurrences and reacotr trips in Swedish nuclear power plants in operation during the period from January 1 to June 30 1985. It is based on the reports submitted by the utilities to the Swedish Nuclear power Inspectorate according to Technical Specifications. Twice a year since 1974 the Inspectorate has issued a compilation on such reported occurrences and reactor trips. Starting with the compilation of the second half of 1982 some new features have been introduced. The most important change is that the volume of information has been increased. The full test, provided by the utilities when reporting the incidents, is now attached to the codified information and also the layout has been altered to facilitate reading. As in the previous reports the occurrences and reactor trips are arranged both alphabetically by facility name and chronologically by report number for each facility. Electricity generation charts for each facility are also presented. The primary purpose of this report is thus to present all the information furnished by utlities when they submit their reports according the Technical Specifications. The only evaluation made by the Inspecotrate is the categorization on the incidents. Like the previous reports this one also presents frequency of incidents as related to affected component, cause of incident etc. The difference is that only information reported by the utilities is used. This is the reason why a considerable proportion of the incidents are categorized as 'other fault'. (author)

The chemical monitoring and control during temporary turbine trip or reactor scram of nuclear power plant

International Nuclear Information System (INIS)

Liu Heng

2012-01-01

During normal operation, a malfunction of equipment or improper operation sometimes results in a turbine trip or reactor scram or even cold shutdown. Because present chemical control strategy and programs aimed at the situation of normal operation and planed refueling outage, no integrate emergency program of radiochemical and chemical control had been developed to focus on this urgent and unexpected situation. After many years of practice and experience feedback, chemists have created an emergency collaborative program of radiochemical and chemical control which aims at these unexpected situations such as unplanned unit down power, turbine trip, or reactor scram. The program defines different radiochemical and chemical control measures and steps during different status to monitor primary loop dose rate variation, fuel assembly integrity and water chemical excursion to prevent components from corrosion. (author)

Reactor protection system software test-case selection based on input-profile considering concurrent events and uncertainties

International Nuclear Information System (INIS)

Khalaquzzaman, M.; Lee, Seung Jun; Cho, Jaehyun; Jung, Wondea

2016-01-01

Recently, the input-profile-based testing for safety critical software has been proposed for determining the number of test cases and quantifying the failure probability of the software. Input-profile of a reactor protection system (RPS) software is the input which causes activation of the system for emergency shutdown of a reactor. This paper presents a method to determine the input-profile of a RPS software which considers concurrent events/transients. A deviation of a process parameter value begins through an event and increases owing to the concurrent multi-events depending on the correlation of process parameters and severity of incidents. A case of reactor trip caused by feedwater loss and main steam line break is simulated and analyzed to determine the RPS software input-profile and estimate the number of test cases. The different sizes of the main steam line breaks (e.g., small, medium, large break) with total loss of feedwater supply are considered in constructing the input-profile. The uncertainties of the simulation related to the input-profile-based software testing are also included. Our study is expected to provide an option to determine test cases and quantification of RPS software failure probability. (author)

Evaluation of the root cause for MSR high level trip in Maanshan

International Nuclear Information System (INIS)

Liao, L.-Y.; Ferng, Y.-M.; Jange, S.J.; Ko, C.M.

2004-01-01

Reactor trip due to Moisture Separator Reheater (MSR) high water level has been a long time issue for Maanshan nuclear power plant. The operating experience shows that there are five reactor trips due to MSR high water level. Four out of the five reactor trips are generated when Combined Intermediate valve (CIV) no. 1 is closed during CIV closure test. The fifth reactor trip occurs when the reactor power is increasing from 99% to 100%. An extensive root cause analysis has been performed by Taipower Company. It is concluded that the water accumulated in the cross under leg between the exhaust of high pressure turbine and the inlet of MSR was the water source contributing to the MSR high level trip. Although, Maanshan does not have similar trip after the root cause analysis, it is interested to evaluate the proposed root cause from thermal hydraulic point of view. It is also hoped that some useful guidelines can be established. This paper includes a description of the scenario of reactor trips, a summary of the root cause analysis done by Taipower Company, an examination of possible mechanisms, an identification of key parameters and a presentation of major findings. In addition, the applicability of RELAP5/MOD3 under this condition is discussed. (author)

Sensitivity analysis on the effect of software-induced common cause failure probability in the computer-based reactor trip system unavailability

International Nuclear Information System (INIS)

Kamyab, Shahabeddin; Nematollahi, Mohammadreza; Shafiee, Golnoush

2013-01-01

Highlights: ► Importance and sensitivity analysis has been performed for a digitized reactor trip system. ► The results show acceptable trip unavailability, for software failure probabilities below 1E −4 . ► However, the value of Fussell–Vesley indicates that software common cause failure is still risk significant. ► Diversity and effective test is founded beneficial to reduce software contribution. - Abstract: The reactor trip system has been digitized in advanced nuclear power plants, since the programmable nature of computer based systems has a number of advantages over non-programmable systems. However, software is still vulnerable to common cause failure (CCF). Residual software faults represent a CCF concern, which threat the implemented achievements. This study attempts to assess the effectiveness of so-called defensive strategies against software CCF with respect to reliability. Sensitivity analysis has been performed by re-quantifying the models upon changing the software failure probability. Importance measures then have been estimated in order to reveal the specific contribution of software CCF in the trip failure probability. The results reveal the importance and effectiveness of signal and software diversity as applicable strategies to ameliorate inefficiencies due to software CCF in the reactor trip system (RTS). No significant change has been observed in the rate of RTS failure probability for the basic software CCF greater than 1 × 10 −4 . However, the related Fussell–Vesley has been greater than 0.005, for the lower values. The study concludes that consideration of risk associated with the software based systems is a multi-variant function which requires compromising among them in more precise and comprehensive studies

Development of INSTEC(INformation System of Trip Event Cases)

International Nuclear Information System (INIS)

Lee, Jeong Woon; Shim, Bong Sik; Park, Keun Oak; Cheon, Se Woo

1996-09-01

In this research, we established an incident analysis procedure based on the concept of interaction between plant components and developed INSTEC(INformation System of Trip Event Cases) which can manage data obtained as the result of incident analysis. The analysis procedure is consisted of the following steps; reconfiguration of incident context, identification of the paths and contents of the interaction between plant components, identification of unit event obstructing normal plant operation, identification of possible erroneous actions, decision of error modes, identification of likely causes, summarization of analysis results. INSTEC was developed to effectively present the result of incident analysis. This system offers the analyzed information such as analysis results of human error cases, operating issues and problems, recommendations to prevent a similar incident, etc. 24 tabs., 18 figs., 10 refs. (Author)

Risk assessment to determine the advisability of seismic trip systems

International Nuclear Information System (INIS)

Cummings, G.E.; Wells, J.E.

1977-01-01

Seismic trip (scram) systems have been used for many years on certain research, test, and production reactors, but not on commercial power reactors. An assessment is made of the risks associated with the presence and absence of such trip systems on power reactors. An attempt was made to go beyond the reactor per se and to consider the risks to society as a whole; for example, the advantages of tripping to avoid an earthquake-caused accident were weighed against the disadvantages associated with interrupting electric power in a time when it would be needed for emergency services. The comparative risk assessment was performed by means of fault tree analysis

Application-specific integrated circuit design for a typical pressurized water reactor pressure channel trip

International Nuclear Information System (INIS)

Battle, R.E.; Manges, W.W.; Emery, M.S.; Vendermolen, R.I.; Bhatt, S.

1994-01-01

This article discusses the use of application-specific integrated circuits (ASICs) in nuclear plant safety systems. ASICs have certain advantages over software-based systems because they can be simple enough to be thoroughly tested, and they can be tailored to replace existing equipment. An architecture to replace a pressurized water reactor pressure channel trip is presented. Methods of implementing digital algorithms are also discussed

Analysis of Peach Bottom turbine trip tests

International Nuclear Information System (INIS)

Cheng, H.S.; Lu, M.S.; Hsu, C.J.; Shier, W.G.; Diamond, D.J.; Levine, M.M.; Odar, F.

1979-01-01

Current interest in the analysis of turbine trip transients has been generated by the recent tests performed at the Peach Bottom (Unit 2) reactor. Three tests, simulating turbine trip transients, were performed at different initial power and coolant flow conditions. The data from these tests provide considerable information to aid qualification of computer codes that are currently used in BWR design analysis. The results are presented of an analysis of a turbine trip transient using the RELAP-3B and the BNL-TWIGL computer codes. Specific results are provided comparing the calculated reactor power and system pressures with the test data. Excellent agreement for all three test transients is evident from the comparisons

Probabilistic study of primary pump trip in a P.W.R. reactor: use of response surface methodology

International Nuclear Information System (INIS)

Bars, C.; Duchemin, B.; Maigret, N.; Peltier, J.; Rostan, O.; Villeneuve, M.J. de; Lanore, J.M.

1981-09-01

This paper describes a probabilistic study about the consequences of the trip or blockage of one of the three PWR reactor primary pumps. The distribution of the input parameters is taken into account and the resulting distribution of the consequence (number of failed fuel rods) is assessed. The necessity to do this study with the response surface methodology and the precautions to take are outlined. The results show that the probability to have failed fuel rods is about 10 -4 for pump trip and 0.16 for blockage with, in this case, a mean of 196 failed rods, that is 0.5 % of total number of rods

Failure mode and effects analysis on typical reactor trip system

International Nuclear Information System (INIS)

Eisawy, E.A.

2010-01-01

An updated failure mode and effects analysis, FMEA , has been performed on a typical reactor trip system. This upgrade helps to avoid system damage and ,as a result, extends the system service life. It also provides for simplified maintenance and surveillance testing. The operating conditions under which the system is to carry out its function and the operational profile expected for the system have been determined. The results of the FMEA have been given in terms of operating states of the subsystem.The results are given in form of table which is set up such that for a given failure one can read across it and determine which items remain operating in the system. From this data one can identify the number of components operating in the system for monitors pressure exceeds the setpoint pressure.

Design and implementation of STD32-BUS based reactor protection trip unit on FPGA imbaby

International Nuclear Information System (INIS)

Mahmoud, I.; Elnokity, O.A.; Refai, M.K.

2007-01-01

This paper presents a way to design and implement the Trip Unit of a Reactor Protection System (RPS) using a Field Programmable Gate Arrays (FPGA). Instead of the traditional embedded Microprocessor based interface design method, a proposed tailor made FPGA based circuit is built to substitute the Trip Unit (TL1) existing in Egypt's 2' ' Research reactor ETRR-2. The existing embedded system is built around the STD32 field Computer Bus which used in industrial and process control applications. It is modular, rugged, reliable, and easy-to-use and is able to support a large mix of I/O cards and to easily change its configuration in the future. Therefore, the state machine of this bus is extracted from its timing diagrams and implemented in VHDL to interface the designed TU circuit. The proposed designed circuit implemented using ALTERA EPF10K10LC84-3 chip replaces the Single Board Computer which have the embedded SAY program of the TU providing the same integrated HAV and SAV functions implemented in FPGA Chip housed in an printed circuit board, which uses the same shape and specifications of STD32 boards. H/W implementation of both TU and STD32 Bus in VHDL addresses the issues of safety and reusability

Relap5 simulation for severe accident analysis of RSG-GAS Reactor

International Nuclear Information System (INIS)

Andi Sofrany Ekariansyah; Endiah P-Hastuti; Sudarmono

2018-01-01

The research reactor in the world is to be known safer than power reactor due to its simpler design related to the core and operational characteristics. Nevertheless, potential hazards of research reactor to the public and the environment can not be ignored due to several special features. Therefore the level of safety must be clearly demonstrated in the safety analysis report (SAR) using safety analysis, which is performed with various approaches and methods supported by computational tools. The purpose of this research is to simulate several accidents in the Indonesia RSG-GAS reactor, which may lead to the fuel damage, to complement the severe accident analysis results that already described in the SAR. The simulation were performed using the thermal hydraulic code of RELAP5/SCDAP/Mod3.4 which has the capability to model the plate-type of RSG-GAS fuel elements. Three events were simulated, which are loss of primary and secondary flow without reactor trip, blockage of core subchannels without reactor trip during full power, and loss of primary and secondary flow followed by reactor trip and blockage of core subchannel. The first event will harm the fuel plate cladding as showed by its melting temperature of 590 °C. The blockage of one or more subchannels in the one fuel element results in different consequences to the fuel plates, in which at least two blocked subchannels will damage one fuel plate, even more the blockage of one fuel element. The combination of loss of primary and secondary flow followed by reactor trip and blockage of one fuel element has provided an increase of fuel plate temperature below its melting point meaning that the established natural circulation and the relative low reactor power is sufficient to cool the fuel element. (author)

Investigation of the loss of forced cooling test by using the high temperature engineering test reactor (HTTR) (Contract research)

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Inaba, Yoshitomo; Goto, Minoru; Tochio, Daisuke

2007-09-01

The three gas circulators trip test and the vessel cooling system stop test as the safety demonstration test by using the High Temperature engineering Test Reactor (HTTR) are under planning to demonstrate inherent safety features of High Temperature Gas-cooled Reactor. All three gas circulators to circulate the helium gas as the coolant are stopped to simulate the loss of forced cooling in the three gas circulators trip test. The stop of the vessel cooling system located outside the reactor pressure vessel to remove the residual heat of the reactor core follows the stop of all three gas circulators in the vessel cooling system stop test. The analysis of the reactor transient for such tests and abnormal events postulated during the test was performed. From the result of analysis, it was confirmed that the three gas circulators trip test and the vessel cooling system stop test can be performed within the region of the normal operation in the HTTR and the safety of the reactor facility is ensured even if the abnormal events would occur. (author)

Development of a new model to evaluate the probability of automatic plant trips for pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Shimada, Yoshio [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan); Kawai, Katsunori; Suzuki, Hiroshi [Mitsubishi Heavy Industries Ltd., Tokyo (Japan)

2001-09-01

In order to improve the reliability of plant operations for pressurized water reactors, a new fault tree model was developed to evaluate the probability of automatic plant trips. This model consists of fault trees for sixteen systems. It has the following features: (1) human errors and transmission line incidents are modeled by the existing data, (2) the repair of failed components is considered to calculate the failure probability of components, (3) uncertainty analysis is performed by an exact method. From the present results, it is confirmed that the obtained upper and lower bound values of the automatic plant trip probability are within the existing data bound in Japan. Thereby this model can be applicable to the prediction of plant performance and reliability. (author)

BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

International Nuclear Information System (INIS)

1987-11-01

This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

The application and design of distributed control system in reactor shutdown system of Qinshan phase III

International Nuclear Information System (INIS)

Su Guoquan; Liu Wangtian; Yu Yijun; Xiong Weihua

2006-03-01

The design, commissioning and running of the reactor trip parameter monitoring system used in Qinshan Phase III are introduced. The applying technology of Distributed Control System realized trip parameter monitoring and realized the function of trip parameters quick data acquisitioning, transferring, saving, alarm, query. The applying of trip parameters monitoring system improved the abilities of plant status monitoring and event analyzing, and increased the security and economy of nuclear power plant. (authors)

Instrumentation and control for reactor power setback in PFBR

International Nuclear Information System (INIS)

Upadhyay, Chandra Kant; Vasal, Tanmay; Nagaraj, C.P.; Madhusoodanan, K.

2013-01-01

In Prototype Fast Breeder Reactor (PFBR), a 500 MWe plant, Reactor Power Setback is a special operation envisaged for bulk power reduction on occurrence of certain events in Balance of Plant. The bulk power reduction requires a large negative reactivity perturbation if reactor is operating on nominal power. This necessitates a reliable monitoring system with fault tolerant I and C architecture in order to inhibit reactor SCRAM on negative reactivity trip signal. The impact of above events on the process is described. Design of a functional prototype module to carry out RPSB logic operation and its interface with other instruments has been discussed. (author)

Study on thermalhydraulics of natural circulation decay heat removal in FBR. Experiment with water of typical reactor trip in the demonstration FBR

International Nuclear Information System (INIS)

Koga, Tomonari; Murakami, Takahiro; Eguchi, Yuzuru

2010-01-01

Intending to enhance safety and to reduce costs, an FBR plant is being developed in Japan. In relies solely on natural circulation of the primary cooling loop to remove a decay heat of the core after reactor trips. A water test was carried out to advance the development. The test used a 1/10 reduced scale model simulating the core and cooling systems. The experiments simulated representative accidents from steady state to decay heat removal through reactor trip and clarified thermal-hydraulic issues on the thermal circulation performance. Some modifications of the system design were proposed for solving serious problems of natural circulation. An improved design complying with the suggestions will make it possible for natural circulation of the cooling systems to remove the decay heat of the core without causing and unstable or unpredictable change. (author)

Management of operational events in research reactor

International Nuclear Information System (INIS)

Zhong Heping; Yang Shuchun; Peng Xueming

2001-01-01

The author describes the tracing management process post-operational event in a research reactor based on nuclear safety code, under the background of the research reactor in Nuclear Power Institute of China. It presorts the definite measures to the event tracing and it up its management factors

Initiating Events for Multi-Reactor Plant Sites

Energy Technology Data Exchange (ETDEWEB)

Muhlheim, Michael David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-09-01

Inherent in the design of modular reactors is the increased likelihood of events that initiate at a single reactor affecting another reactor. Because of the increased level of interactions between reactors, it is apparent that the Probabilistic Risk Assessments (PRAs) for modular reactor designs need to specifically address the increased interactions and dependencies.

Event tree analysis for the system of hybrid reactor

International Nuclear Information System (INIS)

Yang Yongwei; Qiu Lijian

1993-01-01

The application of probabilistic risk assessment for fusion-fission hybrid reactor is introduced. A hybrid reactor system has been analysed using event trees. According to the character of the conceptual design of Hefei Fusion-fission Experimental Hybrid Breeding Reactor, the probabilities of the event tree series induced by 4 typical initiating events were calculated. The results showed that the conceptual design is safe and reasonable. through this paper, the safety character of hybrid reactor system has been understood more deeply. Some suggestions valuable to safety design for hybrid reactor have been proposed

Power reactor events, May-June 1986

International Nuclear Information System (INIS)

Massaro, S.A.

1986-12-01

Power Reactor Events is a bi-monthly newsletter that compiles operating experience information about commercial nuclear power plants. This includes summaries of noteworthy events and listings and/or abstracts of USNRC and other documents that discuss safety-related or possible generic issues. It is intended to feed back some of the lessons learned from operational experience to the various plant personnel, i.e., managers, licensed reactor operators, training coordinators, and support personnel. Events at the following plants are reported: McGuire Unit 1; Susquehanna Units 1 and 2; Browns Ferry Units 1, 2, and 3; and River Bend Unit 1

Reactor protection system

International Nuclear Information System (INIS)

Fairbrother, D.B.; Lesniak, L.M.; Orgera, E.G.

1977-10-01

The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

Analysis of thermal fatigue events in light water reactors

Energy Technology Data Exchange (ETDEWEB)

Okuda, Yasunori [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

2000-09-01

Thermal fatigue events, which may cause shutdown of nuclear power stations by wall-through-crack of pipes of RCRB (Reactor Coolant Pressure Boundary), are reported by licensees in foreign countries as well as in Japan. In this paper, thermal fatigue events reported in anomalies reports of light water reactors inside and outside of Japan are investigated. As a result, it is clarified that the thermal fatigue events can be classified in seven patterns by their characteristics, and the trend of the occurrence of the events in PWRs (Pressurized Water Reactors) has stronger co-relation to operation hours than that in BWRs (Boiling Water Reactors). Also, it is concluded that precise identification of locations where thermal fatigue occurs and its monitoring are important to prevent the thermal fatigue events by aging or miss modification. (author)

Summary of the First Workshop on OECD/NRC boiling water reactor turbine trip benchmark

International Nuclear Information System (INIS)

2000-11-01

The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. This first workshop was focused on technical issues connected with the first draft of

Reactor protection system. Revision 1

International Nuclear Information System (INIS)

Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

1975-04-01

The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

Summary of the OECD/NRC Boiling Water Reactor Turbine Trip Benchmark - Fourth Workshop (BWR-TT4)

International Nuclear Information System (INIS)

2002-01-01

The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. The purpose of this fourth workshop was to present and discuss final results of

Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

International Nuclear Information System (INIS)

Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

2000-01-01

A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

Flow protection trip limits operational charge-discharge facility -- C Reactor

Energy Technology Data Exchange (ETDEWEB)

Van Wormer, F.W.

1958-09-19

Because of wide variations in the venturi throat pressure, well beyond the panellit gage trip range, that occur during the sequence of operational charge-discharge, the panellit gage cannot be included in the scram safety circuit during the period of time that charge- discharge operations are being performed. In its stead, the function of the panellit gage is replaced in an overlapping manner by a tube inlet pressure monitor that is equipped with high and low pressure trip mechanisms that may be included in the scram safety circuit during the time that the panellit gage must be by-passed. The tube inlet pressure monitor is then used to provide the protection from unstable flow that is normally obtained with the panellit gage. This memorandum describes the manner in which the tube inlet pressure monitor trip points are to be determined and used.

Echoes from the Field: An Ethnographic Investigation of Outdoor Science Field Trips

Science.gov (United States)

Boxerman, Jonathan Zvi

As popular as field trips are, one might think they have been well-studied. Nonetheless, field trips have not been heavily studied, and little research has mapped what actually transpires during field trips. Accordingly, to address this research gap, I asked two related research questions. The first question is a descriptive one: What happens on field trips? The second question is explanatory: What field trip events are memorable and why? I employed design research and ethnographic methodologies to study learning in naturally occurring contexts. I collaborated with middle-school science teachers to design and implement more than a dozen field trips. The field trips were nested in particular biology and earth sciences focal units. Students were tasked with making scientific observations in the field and then analyzing this data during classroom activities. Audio and video recording devices captured what happened during the field trips, classroom activities and discussions, and the interviews. I conducted comparative microanalysis of videotaped interactions. I observed dozens of events during the field trips that reverberated across time and place. I characterize the features of these events and the objects that drew interest. Then, I trace the residue across contexts. This study suggests that field trips could be more than one-off experiences and have the potential to be resources to seed and enrich learning and to augment interest in the practice of science.

Estimation of acceptable beam trip frequencies of accelerators for ADS and comparison with performances of existing accelerators

International Nuclear Information System (INIS)

Takei, Hayanori; Tsujimoto, Kazufumi; Nishihara, Kenji; Furukawa, Kazuro; Yano, Yoshiharu; Ogawa, Yujiro; Oigawa, Hiroyuki

2009-09-01

Frequent beam trips as experienced in existing high power proton accelerators may cause thermal fatigue problems in ADS components which may lead to degradation of their structural integrity and reduction of their lifetime. Thermal transient analyses were performed to investigate the effects of beam trips on the reactor components, with the objective of formulating ADS design that had higher engineering possibilities and determining the requirements for accelerator reliability. These analyses were made on the thermal responses of four parts of the reactor components; the beam window, the cladding tube, the inner barrel and the reactor vessel. Our results indicated that the acceptable frequency of beam trips ranged from 50 to 2x10 4 times per year depending on the beam trip duration. As the beam trips for durations exceeding five minutes were assumed to make the plant shut down and restart, the plant availability was estimated to be 70%. In order to consider measures to reduce the frequency of beam trips on the high power accelerator for ADS, we compared the acceptable frequency of beam trips with the operation data of existing accelerators. The result of this comparison showed that for typical conditions the beam trip frequency for durations of 10 seconds or less was within the acceptable level, while that exceeding five minutes should be reduced to about 1/30 to satisfy the thermal stress conditions. (author)

Operating experience feedback from safety significant events at research reactors

Energy Technology Data Exchange (ETDEWEB)

Shokr, A.M. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor; Rao, D. [Bhabha Atomic Research Centre, Mumbai (India)

2015-05-15

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

Operating experience feedback from safety significant events at research reactors

International Nuclear Information System (INIS)

Shokr, A.M.

2015-01-01

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

Computation of a BWR Turbine Trip with CATHARE-CRONOS2-FLICA4 Coupled Codes

International Nuclear Information System (INIS)

Mignot, G.; Royer, E.; Rameau, B.; Todorova, N.

2004-01-01

The CEA/DEN modeling and computation results with the CATHARE, CRONOS2, and FLICA4 codes of the Organisation for Economic Co-operation and Development boiling water reactor turbine trip benchmark are presented. The first exercise of the benchmark to model the whole reactor thermal hydraulics with specified power has been performed with the CATHARE system code. Exercise 2, devoted to core thermal-hydraulic neutronic analysis with provided boundary conditions and neutronic cross sections, has been carried out with the CRONOS2 and FLICA4 codes. Finally, exercise 3, combining system thermal hydraulics and core three-dimensional thermal-hydraulics-neutronics, was computed with the three coupled codes: CATHARE, CRONOS2, and FLICA4.Our one-dimensional thermal-hydraulic reactor computation agrees well with the benchmark reference data and demonstrates the capacities of CATHARE to model a turbine trip transient. Coupled three-dimensional thermal-hydraulic and neutronic analysis displays a high sensitivity of the power peak to the core thermal-hydraulic model. The use of at least 100 channels is recommended to achieve reasonable results for integral and local parameters. Deviations between experimental data and exercise 3 results are discussed: timing of events, core pressure drop, and neutronic model. Finally, analysis of extreme scenarios as sensitivity studies on the transient to assess the effect of the scram, the bypass relief valve, and the steam relief valves is presented

Development, Dedication and Application of an Automatic Seismic Trip System for Nuclear Power Plants of Taiwan Power Company

International Nuclear Information System (INIS)

Liao, Hsin-kai; Lee, Chung-lin; Chen, Chang-kuo; Hsu, Yao-tung; Shyu, Shian-shing

2011-01-01

This paper describes the setups of Automatic Seismic Trip System (ASTS), including development, dedication and implementation, for Nuclear Power Plants (NPPs) of Taiwan Power Company (TPC). The purposed ASTS was designed to trip the reactor when big earthquake occurs. These ASTS were classified as class 1E equipment. They were developed and dedicated for safety applications in accordance with IEEE 323-1983, IEEE 344-1987, IEEE 383-1974 and Reg. Guide 1.180 R1. In order to meet the technical specification required by TPC, three sub-units in the ASTS were developed: Earthquake sensors: Kinemetrices FBA-23 triaxial accelerometers are selected since they were successfully used in Taiwan for seismic monitoring for more than 10 years. Signal conditioning module: It is designed to reduce noise from motion accelerometer (FBA-23) and then transmit seismic signal to the set-point and trip unit via instrument amplify circuit, 0.1 to 10Hz band pass filter circuit, absolute-value converter and voltage to current converter. Trip control module: after comparing the seismic signal level and set-point, the result will decide whether to drive the output relay or not. The output relay is used as the interface between ASTS and the reactor protection system in NPP. For the commercial grade item dedication for safety application, five processes were conducted. Those processes are Seismic test: to use plant specific required response spectrum (RRS), the test required spectrum should envelop RRS: Seismic auto-trip accuracy test: must not trip when filtered PA below set point minus 0.05g, and must trip when filtered PA exceeds set point over 0.05g. Trip signals occurred within 10 second interval are considered as same events: NEMA4 water proof test for sensor box: Anti-radiation test: 8.76x100 rads over 40 years: EMI/EMC test: follow RG 1.180 requirement. The ASTS were installed in three NPPs, six units in total, without connection to RPS in 2006. After one year reliable operation, the

Root-cause Investigation for No Setback Initiation at Liquid Zone Control Unit Perturbation in CANDU6 Reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Donghwan; Kim, Youngae; Kim, Sungmin [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Liquid zone control system (LZCS) is one of the indigenous systems in CANDU type reactor for reactor reactivity control. The LZCS is filled with light water and used to provide a continuous fine control of the reactivity and the reactor power level. This system is also designed to accomplish spatial control of the power distribution, automatically, which prevents xenon induced power oscillations. As the tilt control term is phased out, it is replaced by a level control term, which tends to drive the individual zone levels towards the average level of all the zones. Most of CANDU reactors have been experienced these events. Generally setback or stepback conditions are on when variables of spatial control off, high zone power, etc. are reached to the initiating conditions before ROP trip. But the condition of setback or stepback is not initiated before ROP trip sometime. In this study the root-causes for this event are investigated, and the impact assessment is performed by physics computational modeling. To investigate the root-cause of ROP trip before initiating setback at abnormal operating condition, some LZC perturbation models were simulated and investigated the neutron flux readings of zone detector and ROP detector. Two root-causes were founded. The first, flux variation by water level change is more gradual than other zones due to design characteristics in zone 03. The second, ROP detector (SDS no. 2 3G) in the near zone 03 is very sensitive below 40% of water level due to ROP detector installed position. Even though setback is initiated earlier than ROP trip in case of zone 03 perturbation, ROP trip will be occurred because power decreasing rate is very slow(0.1%/sec) on setback condition.

The development of cause analysis system for CPCS trip using the rule-base deduction

International Nuclear Information System (INIS)

Park, Hee Seok; Kim, Dong Hoon; Seo, Ho Joon; Koo, In Soo; Park, Suk Joon

1992-01-01

The Core Protection Calculator System(CPCS) was developed to initiate a Reactor Trip under the circumstance of certain transients by Combustion Engineering Company. The major function of the CPCS is to generate contact outputs for the Departure from Nucleate Boiling Ratio(DNBR) Trip and Local Power Density(LPD) Trip. But in CPCS the trip causes can not be identified, only trip status is displayed. It may take much time and efforts for plant operator to analyse the trip causes of CPCS. So, the Cause Analysis System for CPCS(CASCPCS) has been developed using the rule-base deduction method to aid the operators in Nuclear Power Plant

An Improved Setpoint Determination Methodology for the Plant Protection System Considering Beyond Design Basis Events

International Nuclear Information System (INIS)

Lee, C.J.; Baik, K.I.; Baek, S.M.; Park, K.-M.; Lee, S.J.

2013-06-01

According to the nuclear regulations and industry standards, the trip setpoint and allowable value for the plant protection system have been determined by considering design basis events. In order to improve the safety of a nuclear power plant, an attempt has been made to develop an improved setpoint determination methodology for the plant protection system trip parameter considering not only a design basis event but also a beyond design basis event. The results of a quantitative evaluation performed for the Advanced Power Reactor 1400 nuclear power plant in Korea are presented herein. The results confirmed that the proposed methodology is able to improve the nuclear power plant's safety by determining more reasonable setpoints that can cover beyond design basis events. (authors)

Small break LOCA analysis for YGN 5 and 6 RCP trip strategy in power mode operation

International Nuclear Information System (INIS)

Kim, Tech Mo; Choi, Han Rim

2001-01-01

A continued operation of Reactor Coolant Pumps(RCPs) during a Small Break Loss of Coolant Accident(SBLOCA) in all operation mode may increase unnecessary inventory loss from the Reactor Coolant System(RCS) causing a severe core uncovery which might lead to fuel failure. After Three Mile Island Unit 2(TMI-2) accident, the Combustion Engineering Owner Group(CEOG) developed RCP trip strategy called 'Trip-Two/Leave-Two' (T2/L2). The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis demonstrates the inherent safety of RCP trip strategy during an SBLOCA for Youggwang Nuclear Power Plant Unit 5 and 6(YGN 5 and 6). The trip setpoint of the first two RCPs for YGN 5 and 6 is calculated to be 1721 psia in pressurizer pressure based on the limiting SBLOCA with 0.15 ft 2 break size in the hot leg. The analysis results show that YGN 5 and 6 can maintain the core coolability even if the operator fails to trip the second two RCPs or trips at the worst time of minimum liquid inventory

Assessment of the MDNBR enhancement methodologies for the SMART control rods banks withdrawal event

International Nuclear Information System (INIS)

Yang, Soo Hyung; Chung, Young-Jong; Kim, Hee Cheol

2005-01-01

For an electricity generation and seawater desalination, a 330 MW System-integrated Modular Advanced ReacTor (SMART) was developed by KAERI. The safety level of the SMART is enhanced when compared to that of the typical commercial reactors, with the aid of an elimination of a large break loss of coolant accident by placing the major components of the primary system in a reactor vessel and the adoption of a new technology and a passive design concept into the safety system. However, the events related to reactivity and power distribution anomalies have been evaluated as vulnerable points when compared to the other initiating events in the SMART, since the reactivity worth of the control rods (CR) banks is quite large due to the boron free core concept. Especially, safety margins, i.e., minimum departure from nucleate boiling ratio (MDNBR), are significantly threatened during the CR banks withdrawal event. Therefore, MDNBR enhancement methodology for the CR banks withdrawal event should be considered to further enhance the safety level of the SMART design. Two methodologies have been suggested to enhance the MDNBR during the CR banks withdrawal event: the application of a DNBR trip function into a core protection system and a turbine trip delay methodology. Sensitivity studies are performed to evaluate the two MDNBR enhancement methodologies and show that the suggested methodologies could enhance the MDNBR during the CR banks withdrawal event of the SMART

External event Probabilistic Risk Assessment for the High Flux Isotope Reactor (HFIR)

International Nuclear Information System (INIS)

Flanagan, G.F.; Johnson, D.H.; Buttemer, D.; Perla, H.F.; Chien, S.H.

1989-01-01

The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 x 10 -4 . In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 50% of the internal event initiated contribution and is dominated by seismic events

Questionnaire-based person trip visualization and its integration to quantitative measurements in Myanmar

Science.gov (United States)

Kimijiama, S.; Nagai, M.

2016-06-01

With telecommunication development in Myanmar, person trip survey is supposed to shift from conversational questionnaire to GPS survey. Integration of both historical questionnaire data to GPS survey and visualizing them are very important to evaluate chronological trip changes with socio-economic and environmental events. The objectives of this paper are to: (a) visualize questionnaire-based person trip data, (b) compare the errors between questionnaire and GPS data sets with respect to sex and age and (c) assess the trip behaviour in time-series. Totally, 345 individual respondents were selected through random stratification to assess person trip using a questionnaire and GPS survey for each. Conversion of trip information such as a destination from the questionnaires was conducted by using GIS. The results show that errors between the two data sets in the number of trips, total trip distance and total trip duration are 25.5%, 33.2% and 37.2%, respectively. The smaller errors are found among working-age females mainly employed with the project-related activities generated by foreign investment. The trip distant was yearly increased. The study concluded that visualization of questionnaire-based person trip data and integrating them to current quantitative measurements are very useful to explore historical trip changes and understand impacts from socio-economic events.

The Effect of Current-Limiting Reactors on the Tripping of Short Circuits in High-Voltage Electrical Equipment

International Nuclear Information System (INIS)

Volkov, M. S.; Gusev, Yu. P.; Monakov, Yu. V.; Cho, Gvan Chun

2016-01-01

The insertion of current-limiting reactors into electrical equipment operating at a voltage of 110 and 220 kV produces a change in the parameters of the transient recovery voltages at the contacts of the circuit breakers for disconnecting short circuits, which could be the reason for the increase in the duration of the short circuit, damage to the electrical equipment and losses in the power system. The results of mathematical modeling of the transients, caused by tripping of the short circuit in a reactive electric power transmission line are presented, and data are given on the negative effect of a current-limiting resistor on the rate of increase and peak value of the transient recovery voltages. Methods of ensuring the standard requirements imposed on the parameters of the transient recovery voltages when using current-limiting reactors in the high-voltage electrical equipment of power plants and substations are proposed and analyzed

Reactor protection system design using micro-computers

International Nuclear Information System (INIS)

Fairbrother, D.B.

1977-01-01

Reactor Protection Systems for Nuclear Power Plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock and Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. This paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System

Reactor protection system design using micro-computers

International Nuclear Information System (INIS)

Fairbrother, D.B.

1976-01-01

Reactor protection systems for nuclear power plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock and Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. The paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System

Impact of Pre-Initiators on PSA in Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Ochirbat, Chimedtseren [KAIST, Daejeon (Korea, Republic of); Kim, Sok Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-10-15

Most of nuclear power plants had already conducted PSA work to examine their plant safety for identifying vulnerability and preparing the mitigating strategies for severe accident. However, the PSA for research reactor has been conducted limitedly comparing with nuclear power plants due to lack of awareness and resources. Most of PSA results demonstrated that human failure events (HFEs) take a major role of risk contributor in terms of core damage frequency. HFEs are categorized as the following three types: pre-initiating event interaction (e.g., maintenance of errors, testing errors, calibration errors), initiating event related interactions (e.g., human error causing loss of power, human error causing system trip), and post-initiating event (e.g., all action actuating manual safety system backup of an automatic system). Lack of resources and utilization of research reactor calls a vicious circle in terms of safety degradation. The safety degradation poses the vulnerability of human failure during research reactor utilization process. Typically, evaluation of pre-initiators related to test and maintenance are not taking into account in PSA for research reactors. This paper aims to investigate the impact of pre-initiating events related to test and maintenance activities on PSA results in terms of core damage frequency for a research reactor.

Impact of Pre-Initiators on PSA in Research Reactor

International Nuclear Information System (INIS)

Ochirbat, Chimedtseren; Kim, Sok Chul

2014-01-01

Most of nuclear power plants had already conducted PSA work to examine their plant safety for identifying vulnerability and preparing the mitigating strategies for severe accident. However, the PSA for research reactor has been conducted limitedly comparing with nuclear power plants due to lack of awareness and resources. Most of PSA results demonstrated that human failure events (HFEs) take a major role of risk contributor in terms of core damage frequency. HFEs are categorized as the following three types: pre-initiating event interaction (e.g., maintenance of errors, testing errors, calibration errors), initiating event related interactions (e.g., human error causing loss of power, human error causing system trip), and post-initiating event (e.g., all action actuating manual safety system backup of an automatic system). Lack of resources and utilization of research reactor calls a vicious circle in terms of safety degradation. The safety degradation poses the vulnerability of human failure during research reactor utilization process. Typically, evaluation of pre-initiators related to test and maintenance are not taking into account in PSA for research reactors. This paper aims to investigate the impact of pre-initiating events related to test and maintenance activities on PSA results in terms of core damage frequency for a research reactor

Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

Energy Technology Data Exchange (ETDEWEB)

Baek, Joo S.; Diamond, David

2016-12-06

A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in the analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.

Power supply with nuclear reactor

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

Each parameter of the processes of a nuclear reactor and components operatively associated therewith is monitored by a set of four like sensors. A trip system normally operates on a 'two out of four' configuration; i.e., to trip the reactor it is necessary that at least two sensors of a set sense an off-normal parameter. This assumes that all sensors are in normal operating condition. However, when a sensor is in test or is subject to maintenance or is defective or disabled, the 'two out of four' configuration would be reduced to a 'one out of three' configuration because the affected sensor is taken out of service. This would expose the system to the possibility that a single sensor failure, which may be spurious, will cause a trip of the reactor. To prevent this, it is necessary that the affected sensor be bypassed. If only one sensor is bypassed, the system operates on a 'two out of three' configuration. With two sensors bypassed, the sensing of an off-normal parameter by a third sensor trips the reactor

Seismic damage sensing of bridge structures with TRIP reinforcement steel bars

Science.gov (United States)

Adachi, Yukio; Unjoh, Shigeki

2001-07-01

Intelligent reinforced concrete structures with transformation-induced-plasticity (TRIP) steel rebars that have self-diagnosis function are proposed. TRIP steel is special steel with Fe-Cr based formulation. It undergoes a permanent change in crystal structure in proportion to peak strain. This changes from non-magnetic to magnetic steel. By using the TRIP steel rebars, the seismic damage level of reinforced concrete structures can be easily recognized by measuring the residual magnetic level of the TRIP rebars, that is directly related to the peak strain during a seismic event. This information will be most helpful for repairing the damaged structures. In this paper, the feasibility of the proposed intelligent reinforced concrete structure for seismic damage sensing is experimentally studied. The relation among the damage level, peak strain of rebars, and residual magnetic level of rebars of reinforced concrete beams implemented with TRIP steel bars was experimentally studied. As the result of this study, this intelligent structure can diagnose accumulated strain/damage anticipated during seismic event.

C-Reactor I and E loading instability limits

Energy Technology Data Exchange (ETDEWEB)

Hess, K.W.

1957-01-24

The pilot charging of I & E fuel elements has been implemented at C-Reactor under Production Test IP-19-A. It was necessary to provide adequate tube protection against flow interruption by establishing proper trip setting on the Panellit pressure gauges. the administration of these Panellit trip settings is done by trip-before- boiling tube outlet temperature limits, which are similar in principle to the current instability limits. Trip-before-boiling limits for C-Reactor I & E fuel elements loadings are presented in this document.

Nuclear reactor safety protection device

International Nuclear Information System (INIS)

Okido, Fumiyasu; Noguchi, Atomi; Matsumiya, Shoichi; Furusato, Ken-ichiro; Arita, Setsuo.

1994-01-01

The device of the present invention extremely reduces a probability of causing unnecessary scram of a nuclear reactor. That is, four control devices receive signals from each of four sensors and output four trip signals respectively in a quardruplicated control device. Each of the trip signals and each of trip signals via a delay circuit are inputted to a logical sum element. The output of the logical sum circuit is inputted to a decision of majority circuit. The decision of majority circuit controls a scram pilot valve which conducts scram of the reactor by way of a solenoid coils. With such procedures, even if surge noises of a short pulse width are mixed to the sensor signals and short trip signals are outputted, there is no worry that the scram pilot valve is actuated. Accordingly, factors of lowering nuclear plant operation efficiency due to erroneous reactor scram can be reduced. (I.S.)

Primary heat transport pump trip by ground fault (deterioration of insulation in the cable quick disconnect)

International Nuclear Information System (INIS)

Chun, C.-Y.

1991-01-01

At 08:29 Sept. 1, 1988, Wolsong unit 1 was operating at 100% full power when a primary heat transport pump was suddenly tripped by breaker trip due to ground fault in the power distribution connector assembly. Soon after the pump trip, the reactor was shut down automatically on low heat transport flow. Operators tried to restart the pump twice but failed. A field operator reported to the shift supervisor that he found an electrical spark and smoke at the vicinity of the pump when the pump started to run. Inspection showed that a power distribution connector assembly for making fast and easy power connections to the PHT pump motor, 3312-PM2, was damaged severely by thermal shock. Particularly, broken parts of the insulating plug flew away across the boiler room and dropped to the floor. Direct causes of the failure were bad contact and deterioration of integrity along the creep paths between the insulating plug and the connector housing. The failed connector assembly had been used for more than 7 years. Its status had been checked infrequently during the in-service period. The standard torque value was not applied to the installation of connectors. Therefore, we concluded that long term inservice in combinations of application of improper torque value induced failure of insulation. This paper describes the scenarios, causes of the event and corrective actions to prevent recurrence of this event. (author)

Primary heat transport pump trip by ground fault (deterioration of insulation in the cable quick disconnect)

Energy Technology Data Exchange (ETDEWEB)

Chun, C -Y [Wolsong Nuclear Power Plant, Korea Electric Power Corporation, Wolsong (Korea, Republic of)

1991-04-01

At 08:29 Sept. 1, 1988, Wolsong unit 1 was operating at 100% full power when a primary heat transport pump was suddenly tripped by breaker trip due to ground fault in the power distribution connector assembly. Soon after the pump trip, the reactor was shut down automatically on low heat transport flow. Operators tried to restart the pump twice but failed. A field operator reported to the shift supervisor that he found an electrical spark and smoke at the vicinity of the pump when the pump started to run. Inspection showed that a power distribution connector assembly for making fast and easy power connections to the PHT pump motor, 3312-PM2, was damaged severely by thermal shock. Particularly, broken parts of the insulating plug flew away across the boiler room and dropped to the floor. Direct causes of the failure were bad contact and deterioration of integrity along the creep paths between the insulating plug and the connector housing. The failed connector assembly had been used for more than 7 years. Its status had been checked infrequently during the in-service period. The standard torque value was not applied to the installation of connectors. Therefore, we concluded that long term inservice in combinations of application of improper torque value induced failure of insulation. This paper describes the scenarios, causes of the event and corrective actions to prevent recurrence of this event. (author)

The LEP RF Trip and Beam Loss Diagnostics System

CERN Document Server

Arnaudon, L; Beetham, G; Ciapala, Edmond; Juillard, J C; Olsen, R

2002-01-01

During the last years of operation the number of operationally independent RF stations distributed around LEP reached a total of 40. A serious difficulty when running at high energy and high beam intensities was to establish cause and effect in beam loss situations, where the trip of any single RF station would result in beam loss, rapidly producing further multiple RF station trips. For the last year of operation a fast post-mortem diagnostics system was developed to allow precise time-stamping of RF unit trips and beam intensity changes. The system was based on eight local DSP controlled fast acquisition and event recording units, one in each RF sector, connected to critical RF control signals and fast beam intensity monitors and synchronised by GPS. The acquisition units were armed and synchronised at the start of each fill. At the end of the fill the local time-stamped RF trip and beam intensity change history tables were recovered, events ordered and the results stored in a database for subsequent analys...

Effect of reactor conditions on MSIV-ATWS power level

International Nuclear Information System (INIS)

Diamond, D.J.

1987-01-01

In a boiling water reactor (BWR) when there is closure of the main steam isolation valves (MSIVs), the energy generated in the core will be transferred to the pressure suppression pool (PSP) via steam that flows out of the relief valves. The pool has limited capacity as a heat sink and hence, if there is no reactor trip [an anticipated transient without scram (ATWS) event], there is the possibility that the pool temperature may rise beyond acceptable limits. The present study was undertaken to determine how the initial reactor conditions affect the power level during an MSIV-ATWS event. The time of interest is the 20- to 30-min period when it is assumed that the reactor is in a quasi equilibrium condition with the water level and pressure fixed, natural circulation conditions and no control rod movement or significant boron in the core. The initial conditions of interest are the time of the cycle and the operating state

The analysis of the initiating events in thorium-based molten salt reactor

International Nuclear Information System (INIS)

Zuo Jiaxu; Song Wei; Jing Jianping; Zhang Chunming

2014-01-01

The initiation events analysis and evaluation were the beginning of nuclear safety analysis and probabilistic safety analysis, and it was the key points of the nuclear safety analysis. Currently, the initiation events analysis method and experiences both focused on water reactor, but no methods and theories for thorium-based molten salt reactor (TMSR). With TMSR's research and development in China, the initiation events analysis and evaluation was increasingly important. The research could be developed from the PWR analysis theories and methods. Based on the TMSR's design, the theories and methods of its initiation events analysis could be researched and developed. The initiation events lists and analysis methods of the two or three generation PWR, high-temperature gascooled reactor and sodium-cooled fast reactor were summarized. Based on the TMSR's design, its initiation events would be discussed and developed by the logical analysis. The analysis of TMSR's initiation events was preliminary studied and described. The research was important to clarify the events analysis rules, and useful to TMSR's designs and nuclear safety analysis. (authors)

A study on the regulatory approach of KNGR multiple failure events

Energy Technology Data Exchange (ETDEWEB)

Chang, Keun Sun; Kweon, Y. C.; Kang, H. J.; Lee, S. J.; Lee, Y. S.; Moon, J. J.; Lee, M. K. [Sumoon Univ., Asan (Korea, Republic of); Jeong, Ji Hwan [Baekseok College, Cheonan (Korea, Republic of); Yang, S. H. [Korea Association for Nuclear Technology, Taejon (Korea, Republic of)

2000-02-15

This project is to provide the regulatory direction of 3 major technical issues for the Korean Next Generation Reactors, which are parts of major technical issues resulted from the safety regulation R and D on the KNGR. The outstanding results are as follows : through comparison and analysis of domestic/international requirements related to SBO, additional items, which are considered in SSRs, are identified. According to investigation, procedure and training should be included in SSRs, and plant-specific capability analysis requirement contains initial condition, acceptance and addition analysis on the leak rake through RCP seal, etc. In addition, state of the art on the major items related to SBO requirement are described. Several safety analysis requirements are suggested that are needed to be used in the analyses which are aiming to show the ability of the SDVS to cope with TLOFW event. The suggested requirements include suggestions in BE method, reactor thermal power and decay heat, time to reactor trip, time to RCP trip, operator response time, pressurizer and steam generator, and thermal-hydraulic models related to TLOFW event. It is recommended that Moody model mentioned in 10CFR50 appendix K should be excluded in calculation of discharge flow through bleed valves in case of a TLOFW event. Some requirements on initial and boundary conditions are suggested to be used in the analyses of NPPs during MSGTR events. The suggestion includes requirements on analysis method, number of reptured tubes, repture location, operator response time, primary coolant leak flow, and acceptance criteria. As there has been no occurrence of MSGTR event and little literatures reporting analysis results of the event, some items need more study. In addition, some analyses are needed in order to fine the rupture location which gives the most conservative consequence.

A study on the regulatory approach of KNGR multiple failure events

International Nuclear Information System (INIS)

Chang, Keun Sun; Kweon, Y. C.; Kang, H. J.; Lee, S. J.; Lee, Y. S.; Moon, J. J.; Lee, M. K.; Jeong, Ji Hwan; Yang, S. H.

2000-02-01

This project is to provide the regulatory direction of 3 major technical issues for the Korean Next Generation Reactors, which are parts of major technical issues resulted from the safety regulation R and D on the KNGR. The outstanding results are as follows : through comparison and analysis of domestic/international requirements related to SBO, additional items, which are considered in SSRs, are identified. According to investigation, procedure and training should be included in SSRs, and plant-specific capability analysis requirement contains initial condition, acceptance and addition analysis on the leak rake through RCP seal, etc. In addition, state of the art on the major items related to SBO requirement are described. Several safety analysis requirements are suggested that are needed to be used in the analyses which are aiming to show the ability of the SDVS to cope with TLOFW event. The suggested requirements include suggestions in BE method, reactor thermal power and decay heat, time to reactor trip, time to RCP trip, operator response time, pressurizer and steam generator, and thermal-hydraulic models related to TLOFW event. It is recommended that Moody model mentioned in 10CFR50 appendix K should be excluded in calculation of discharge flow through bleed valves in case of a TLOFW event. Some requirements on initial and boundary conditions are suggested to be used in the analyses of NPPs during MSGTR events. The suggestion includes requirements on analysis method, number of reptured tubes, repture location, operator response time, primary coolant leak flow, and acceptance criteria. As there has been no occurrence of MSGTR event and little literatures reporting analysis results of the event, some items need more study. In addition, some analyses are needed in order to fine the rupture location which gives the most conservative consequence

N reactor external events probabilistic risk assessment

International Nuclear Information System (INIS)

Baxter, J.T.

1989-01-01

An external events probabilistic risk assessment of the N Reactor has been completed. The methods used are those currently being proposed for external events analysis in NUREG-1150. Results are presented for the external hazards that survived preliminary screening. They are earthquake, fire, and external flood. Core damage frequencies for these hazards are shown to be comparable to those for commercial pressurized water reactors. Dominant fire sequences are described and related to 10 CFR 50, Appendix R design requirements. Potential remedial measures that reduce fire core damage risk are described including modifications to fire protection systems, procedure changes, and addition of new administrative controls. Dominant seismic sequences are described. The effect of non-safety support system dependencies on seismic risk is presented

The Steam Generating Heavy Water Reactor

International Nuclear Information System (INIS)

Middleton, J.E.

1975-01-01

An account is given of the SGHWR, the prototype of which was built by the United Kingdom Atomic Energy Authority at Winfrith, under the following headings: Introduction; origin of the SGHWR concept; conceptual design (choice of reactor type, steam cycle, reactor coolant system, nuclear behaviour, fuel design, core design, and protective, auxiliary and containment systems); operation and control (integrity of core cooling, reactivity control, power trimming, long term reactivity control, xenon override, load following, power shaping, spatial stability control, void coefficient); protective systems (breached coolant circuit trip, intact coolant circuits trip, power set-back trip); dynamic characteristics; reactor control; station control (decoupled control system, coupled control system, rate of response); Winfrith prototype (design and safety philosophy, conceptual features and parameters, reactor coolant system, protective systems, emergency core cooling, core structure, fuel design, vented containment). (U.K.)

Review of operational experience with the gas-cooled Magnox reactors of the United Kingdom Central Electricity Generating Board

International Nuclear Information System (INIS)

Cave, L.; Clarke, A.W.

1984-01-01

The paper provides a review, which is mainly of a statistical nature, of 260 reactor years of operating experience which the (United Kingdom) Central Electricity Generating Board (CEGB) has obtained with its gas-cooled, graphite moderated Magnox reactors. The main emphasis in the review is on safety rather than on availability. Data are provided on the overall incidence and frequencies of faults and it is shown that the plant items which are predominantly responsible for recorded faults are the gas circulators and the turbo-alternators. Analysis of the reactor trip experience shows that the incidence of events which necessitate an automatic shutdown of the reactor has been about one per reactor year and that of other events leading to a reactor trip has not been much higher (1.4 per reactor year). As would be expected from the length of the operating experience, some relatively rare events have occurred (expected frequency 10 -2 per reactor year, or less) but on each occasion the reactor shutdown system and decay heat removal systems functioned satisfactorily. No overheating of, or damage to, the fuel occurred as a result of these rare events or of other, more frequent, faults. Analysis of the trend of failure rates has shown an improvement with time in nearly all safety-related items and external inspection of the primary coolant circuits has shown no significant deterioration with time. However, some derating of the reactors has been necessary to reduce the effects of oxidation of mild steel in CO 2 , in order to obtain optimum service lives. In spite of major differences between the systems, a comparison of the failure rates of analogous systems and plant items in PWRs and the Magnox reactors show a considerable similarity. Overall, the review of CEGB's operational experience with its Magnos reactors has shown that the frequencies of faults in systems and plant items has been satisfyingly low. (author)

Method for controlling FBR type reactor

International Nuclear Information System (INIS)

Tamano, Toyomi; Iwashita, Tsuyoshi; Sakuragi, Masanori

1991-01-01

The present invention provides a controlling method for moderating thermal transient upon trip in an FBR type reactor. A flow channel for bypassing an intermediate heat exchanger is disposed in a secondary Na system. Then, bypassing flow rate is controlled so as to suppress fluctuations of temperature at a primary exit of the intermediate heat exchanger. Bypassing operation by using the bypassing flow channel is started at the same time with plant trip, to reduce the flow rate of secondary Na flown to the intermediate heat exchanger, so that the imbalance between the primary and the secondary Na flowrates is reduced. Accordingly, fluctuations of the temperature at the primary exit of the intermediate heat exchanger upon trip is suppressed. In view of the above, thermal transient applied to the reactor container upon plant trip can be moderated. As a result, the working life of the reactor can be extended, to improve plant integrity and safety. (I.S.)

Summary of the OECD/NRC Boiling Water Reactor Turbine Trip Benchmark - Fifth Workshop (BWR-TT5)

International Nuclear Information System (INIS)

2003-01-01

The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. The purpose of this fifth workshop was to discuss the results from Phase III (best

The Results of a Site Repair after a High Vibration Trip of a Secondary Cooling Fan in HANARO

International Nuclear Information System (INIS)

Park, Yong-Chul; Kim, Yang-Gon; Lee, Yong-Sub; Jung, Hawn-Seong; Lim, In-Cheol

2007-01-01

HANARO, an open-tank-in-pool type research reactor of 30 MWth power in Korea, which is different from a power plant reactor, exhausts a heat generated from the reactor core into the atmosphere through a secondary cooling tower instead of an electric power production from the heat. After a cooling tower overhaul, No. 2 cooling fan of the cooling tower was stopped by a high vibration trip while HANARO was operating normally. This paper describes the development of a high vibration trip of the cooling fan and the results of a site repair of the cooling fan

Nuclear plant-aging research on reactor protection systems

International Nuclear Information System (INIS)

Meyer, L.C.

1988-01-01

This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed

Process Control Logic Modification to Mitigate Transient Following Tripping of a Primary Circulating Pump for a 540 MWe PHWR Power Plant

International Nuclear Information System (INIS)

Contractor, Ankur D; Gaikwad, Avinash J.; Kumar, Rajesh; Chakraborty, G.; Vhora, S.F.

2006-01-01

The 540 MWe Indian Pressurised Heavy Water Reactor (PHWR) incorporates many new features as compared to the earlier 220 MWe PHWRs. To evaluate the new design features like Primary Heat Transport (PHT) system configuration with two loops, four Primary Circulating Pumps (PCPs) and four passes through core, addition of a Pressurizer (surge Tank) in the PHT system along with Feed/Bleed system and their safety related implications, simulation model have been developed. A reactor step-back is proposed following one PCP trip. The corresponding PCP in the healthy loop is tripped to avoid asymmetrical flow and pressure distribution in the two identical loops. In spite of such elaborate provisions, the margins from high/low PHT pressure are small following tripping of one PCP. Mathematical models for all the major components and sub-systems of the proposed 540 MWe PHWR were developed based on the conservation equations of mass, momentum, energy and equation of state. All the associated control systems are also modeled. The PHT system includes the reactor core with nuclear fuel, PCP, PHT system pressure controller with feed/bleed system and Pressurizer (Surge Tank). The secondary system includes mainly the Steam Generators (SGs), the SG level and pressure controllers, apart from the various steam cycle components. All these models are integrated together to form the Plant Transient Analysis Computer Code Dyna540. The scenario following one PCP trips leads to different states (high/low pressure in Reactor Outlet Header (ROH)) depending upon the banks in which the PCP trips. The pressurizer is connected to two ROHs on one side of the reactor. The system pressure is controlled based on average of four ROHs pressure. In the case of asymmetrical pump operation, this logic leads to a situation where individual ROH pressure goes very near the low/high PHT system pressure trip set point, even though the controlled average pressure is very close to the set pressure. The PHT high

External events analysis for the Savannah River Site K reactor

International Nuclear Information System (INIS)

Brandyberry, M.D.; Wingo, H.E.

1990-01-01

The probabilistic external events analysis performed for the Savannah River Site K-reactor PRA considered many different events which are generally perceived to be ''external'' to the reactor and its systems, such as fires, floods, seismic events, and transportation accidents (as well as many others). Events which have been shown to be significant contributors to risk include seismic events, tornados, a crane failure scenario, fires and dam failures. The total contribution to the core melt frequency from external initiators has been found to be 2.2 x 10 -4 per year, from which seismic events are the major contributor (1.2 x 10 -4 per year). Fire initiated events contribute 1.4 x 10 -7 per year, tornados 5.8 x 10 -7 per year, dam failures 1.5 x 10 -6 per year and the crane failure scenario less than 10 -4 per year to the core melt frequency. 8 refs., 3 figs., 5 tabs

Safety design/analysis and scenario for prevention of CDA with ECCS in lead-bismuth-cooled fast reactor

International Nuclear Information System (INIS)

Minoru, Takahashi; Vaclav, Dostal; Abu Khalid, Rivai; Novitrian; Yumi, Yamada

2007-01-01

Safety design has been developed to show safety feature of Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR). The core is designed to have negative void reactivity even if the entire core and upper plenum are voided by steam intrusion from above. In-vessel type control rod driving mechanisms are used to prevent control rods from accidental ejection due to high pressure in the reactor vessel. In cases of coolant leakage from reactor vessel and feed water pipes, Pb-Bi coolant level in the reactor vessel is kept at the required level for decay heat removal by means of closed type guard vessel. Dual pipes are adopted to avoid leak of water in the feedwater system. Pump trip in feedwater systems initiates loss of coolant flow (LOF) event, although there is no concern of loss of flow accident due to primary pump trip. Injection of high pressure water slows down the flow-coast-down of feedwater at the LOF event. It has been evaluated that the fuel temperature is kept lower than safety limits at the unprotected loss of flow and heat sink (ATWS). A scenario for prevention of the core disruptive accident (CDA) with the emergency core cooling system (ECCS) is examined. The reactor becomes super-critical when the reactor vessel is filled with water. It is necessary to use water with boric acid for the ECC system, and additional backup rods for sub-critical core in water injection. (authors)

Application of IAEA's International Nuclear Event Scale to events at testing/research reactors in Japan

International Nuclear Information System (INIS)

Nozawa, Masao; Watanabe, Norio

1999-01-01

The International Nuclear Event Scale (INES) is a means for providing prompt, clear and consistent information related to nuclear events and facilitating communication between the nuclear community, the media and the public on such events. This paper describes the INES rating process for events at testing/research reactors and nuclear fuel processing facilities and experience on the application of the INES scale in Japan. (author)

Reactor limitation system improves the safety and availability of the Angra 2 nuclear power plant

International Nuclear Information System (INIS)

Souza Mendes, J.E. de

1987-01-01

Beyond the classic Reactor Protection System and Reactor Control System, nuclear plant Angra 2 has a third system called Reactor Limitation System which combines the intelligence features of the control systems with the high reliability of the protection systems. In determined events, which are not controlled by the control system (e.g.: load rejection, failure of one main reactor coolant pump), the Reactor Limitation System actuates automatically in order to lead the plant to a safe operating condition and so it avoids the actuation of the Reactor Protection System and consequently the reactor trip. This increases safety and availability of the plant and reduces component stresses. After the safe operating condition is reached, the process guidance automatically returns to the control systems. (Author) [pt

Retran simulation of Oyster Creek generator trip startup test

International Nuclear Information System (INIS)

Alammar, M.A.

1987-01-01

RETRAN simulation of Oyster Creek generator trip startup test was carried out as part of Oyster Creek RETRAN model qualification program for reload licensing applications. The objective of the simulation was to qualify the turbine model and its interface with the control valve and bypass systems under severe transients. The test was carried out by opening the main breakers at rated power. The turbine speed governor closed the control valves and the pressure regulator opened the bypass valves within 0.5 sec. The stop valves closed by a no-load turbine trip, before the 10 percent overspeed trip was reached and the reactor scrammed on high APRM neutron flux. The simulation resulted in qualifying a normalized hydraulic torque for the turbine model and a 0.3 sec, delay block for the bypass model to account for the different delays in the hydraulic linkages present in the system. One-dimensional kinetics was used in this simulation

On line test of trip channels and actuators in primary shutdown system for RAPP-3,4/KAIGA-1,2 reactors

International Nuclear Information System (INIS)

Pramanik, M.; Gupta, P.K.; Ravi Prakash

1997-01-01

Several types of system design and logic arrangements have been used for reactor shutdown systems to avoid the possibility that a single failure within the trip channels/shutdown system actuators can prevent a shutdown system actuation. The trip channels and the logic arrangements associated with the shutdown systems use redundancy to allow them to continue to operate successfully even after having a certain number of failures. A periodic test is thus needed to detect and repair/replace failed elements to prevent accumulation and eventual system failure. The test must be capable of detecting the first failure. The design initiates shutdown system actuation by deenergising the logic relays and turning off the power to the final electrical actuators. Thus, the systems are fail safe with respect to loss of electrical power to the instruments, logic channels and the actuators. Several system/logic arrangements are used to reduce the chances of spurious actuation caused by the loss of a single power supply and other single failures. In general, the systems use coincidence of instrument channel trips and have separate power supplies for the individual instrument channel and dual power supplies where a single final control element is used. These features also permit on line test of instrument channels and logic train. On line test detects component failures not found by other means. The test determines whether gross failure has occurred rather than perform a calibration. As far as practicable the whole channel from sensors to logic and final control element is to be tested. (author)

The C language auto-generation of reactor trip logic caused by steam generator water level using CASE tools

International Nuclear Information System (INIS)

Kim, Jang Yeol; Lee, Jang Soo

1999-01-01

The purpose is to produce a model of nuclear reactor trip logic caused by the steam generator water level of Wolsung 2/3/4 unit through an activity chart and a statechart and to produce C language automatically using statechart-based formalism and statemate MAGNUM toolset suggested by David Harel Formalism. It was worth attempting auto-generation of C language through we manually made Software Requirement specification(SRS) for safety-critical software using statechart-based formalism. Most of the phase of the software life-cycle except the software requirement specification of an analysis phase were generated automatically by Computer Aided Software Engineering(CASE) tools. It was verified that automatically produced C language has high productivity, portability, and quality through the simulation. (Author). 6 refs., 6 figs

Analysis of unprotected overcooling events in the Integral Fast Reactor

International Nuclear Information System (INIS)

Vilim, R.B.

1989-01-01

Simple analytic models are developed for predicting the response of a metal fueled, liquid-metal cooled reactor to unprotected overcooling events in the balance of plant. All overcooling initiators are shown to fall into two categories. The first category contains these events for which there is no final equilibrium state of constant overcooling, as in the case for a large steam leak. These events are analyzed using a non-flow control mass approach. The second category contains those events which will eventually equilibrate, such as a loss of feedwater heaters. A steady flow control volume analysis shows that these latter events ultimately affect the plant through the feedwater inlet to the steam generator. The models developed for analyzing these two categories provide upper bounds for the reactor's passive response to overcooling accident initiators. Calculation of these bounds for a prototypic plant indicate that failure limits -- eutectic melting, sodium boiling, fuel pin failure -- are not exceeded in any overcooling event. 2 refs

Nuclear reactors

International Nuclear Information System (INIS)

Yoshioka, Michiko.

1985-01-01

Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

Initiating Event Analysis of a Lithium Fluoride Thorium Reactor

Science.gov (United States)

Geraci, Nicholas Charles

The primary purpose of this study is to perform an Initiating Event Analysis for a Lithium Fluoride Thorium Reactor (LFTR) as the first step of a Probabilistic Safety Assessment (PSA). The major objective of the research is to compile a list of key initiating events capable of resulting in failure of safety systems and release of radioactive material from the LFTR. Due to the complex interactions between engineering design, component reliability and human reliability, probabilistic safety assessments are most useful when the scope is limited to a single reactor plant. Thus, this thesis will study the LFTR design proposed by Flibe Energy. An October 2015 Electric Power Research Institute report on the Flibe Energy LFTR asked "what-if?" questions of subject matter experts and compiled a list of key hazards with the most significant consequences to the safety or integrity of the LFTR. The potential exists for unforeseen hazards to pose additional risk for the LFTR, but the scope of this thesis is limited to evaluation of those key hazards already identified by Flibe Energy. These key hazards are the starting point for the Initiating Event Analysis performed in this thesis. Engineering evaluation and technical study of the plant using a literature review and comparison to reference technology revealed four hazards with high potential to cause reactor core damage. To determine the initiating events resulting in realization of these four hazards, reference was made to previous PSAs and existing NRC and EPRI initiating event lists. Finally, fault tree and event tree analyses were conducted, completing the logical classification of initiating events. Results are qualitative as opposed to quantitative due to the early stages of system design descriptions and lack of operating experience or data for the LFTR. In summary, this thesis analyzes initiating events using previous research and inductive and deductive reasoning through traditional risk management techniques to

Operating Experience from Events Reported to the IAEA Incident Reporting System for Research Reactors

International Nuclear Information System (INIS)

2015-03-01

Operating experience feedback is an effective mechanism in providing lessons learned from events and the associated corrective actions to prevent them, helping to improve safety at nuclear installations. The Incident Reporting System for Research Reactors (IRSRR), which is operated by the IAEA, is an important tool for international exchange of operating experience feedback for research reactors. The IRSRR reports contain information on events of safety significance with their root causes and lessons learned which help in reducing the occurrence of similar events at research reactors. To improve the effectiveness of the system, it is essential that national organizations demonstrate an appropriate interest for the timely reporting of events important to safety and share the information in the IRSRR database. At their biennial technical meetings, the IRSRR national coordinators recommended collecting the operating experience from the events reported to the IRSRR and disseminating it in an IAEA publication. This publication highlights the root causes, safety significance, lessons learned, corrective actions and the causal factors for the events reported to the IRSRR up to September 2014. The publication also contains relevant summary information on research reactor events from sources other than the IRSRR, operating experience feedback from the International Reporting System for Operating Experience considered relevant to research reactors, and a description of the elements of an operating experience programme as established by the IAEA safety standards. This publication will be of use to research reactor operating organizations, regulators and designers, and any other organizations or individuals involved in the safety of research reactors

RETRAN analysis of San Onofre Unit 2 turbine trip from 100% power

International Nuclear Information System (INIS)

Ting, Y.P.

1985-01-01

During the San Onofre Nuclear Generating Station Unit (SONGS 2) startup test, the plant experienced a turbine trip from 100% power on June 16, 1983. The trip was initiated by the condenser pressure switch malfunctioning. The plant computers were operating and recorded many plant key parameters. The resulting trip behaved as if it has been manually initiated and it was considered equivalent to a preplanned turbine trip test. A RETRAN-02 model was developed to simulate the SONGS 2 June 16 turbine trip event. The RETRAN analysis of the trip is a continuing effort of in-house SONGS 2 RETRAN model development to benchmark the calculations against the plant startup test data. The overall agreement between measured data and the RETRAN calculations was very good, providing confidence in the capability of the model and the RETRAN program. Comparative data are presented

OECD/NRC BWR Turbine Trip Transient Benchmark as a Basis for Comprehensive Qualification and Studying Best-Estimate Coupled Codes

International Nuclear Information System (INIS)

Ivanov, Kostadin; Olson, Andy; Sartori, Enrico

2004-01-01

An Organisation for Economic Co-operation and Development (OECD)/U.S. Nuclear Regulatory Commission (NRC)-sponsored coupled-code benchmark has been initiated for a boiling water reactor (BWR) turbine trip (TT) transient. Turbine trip transients in a BWR are pressurization events in which the coupling between core space-dependent neutronic phenomena and system dynamics plays an important role. In addition, the available real plant experimental data make this benchmark problem very valuable. Over the course of defining and coordinating the BWR TT benchmark, a systematic approach has been established to validate best-estimate coupled codes. This approach employs a multilevel methodology that not only allows for a consistent and comprehensive validation process but also contributes to the study of different numerical and computational aspects of coupled best-estimate simulations. This paper provides an overview of the OECD/NRC BWR TT benchmark activities with emphasis on the discussion of the numerical and computational aspects of the benchmark

PRISM reactor system design and analysis of postulated unscrammed events

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.C.

1991-01-01

Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. (author)

Analysis of 'human element related trip case book in Korean NPPs' using organizational factors

International Nuclear Information System (INIS)

Kim, S. Y.; Kim, Y. I.; Lee, Y. S.; Kim, C. S.; Jung, C. H.; Jung, W. D.

2002-01-01

There have been no studies appling organizational factors to data analysis in Korean NPPs. In this paper, data in 'human element related trip case book in Korean NPPs' are analyzed and categorized by the 20 organizational factors of NRC-BNL according to the cause of reactor trip. These inform us how organizational factors affected on the safety of Korean NPPs. Consequently important organizational factor are identified through which it is known that NPP organization would have a tendency

Verification of CTF/PARCSv3.2 coupled code in a Turbine Trip scenario

International Nuclear Information System (INIS)

Abarca, A.; Hidalga, P.; Miro, R.; Verdu, G.; Sekhri, A.

2017-01-01

Multiphysics codes had revealed as a best-estimate approach to simulate core behavior in LWR. Coupled neutronics and thermal-hydraulics codes are being used and improved to achieve reliable results for reactor safety transient analysis. The implementation of the feedback procedure between the coupled codes at each time step allows a more accurate simulation and a better prediction of the safety limits of analyzed scenarios. With the objective of testing the recently developed CTF/PARCSv3.2 coupled code, a code-to-code verification against TRACE has been developed in a BWR Turbine Trip scenario. CTF is a thermal-hydraulic subchannel code that features two-fluid, three-field representation of the two-phase flow, while PARCS code solves the neutronic diffusion equation in a 3D nodal distribution. PARCS features allow as well the use of extended sets of cross section libraries for a more precise neutronic performance in different formats like PMAX or NEMTAB. Using this option the neutronic core composition of KKL will be made taking advantage of the core follow database. The results of the simulation will be verified against TRACE results. TRACE will be used as a reference code for the validation process since it has been a recommended code by the USNRC. The model used for TRACE includes a full core plus relevant components such as the steam lines and the valves affecting and controlling the turbine trip evolution. The coupled code performance has been evaluated using the Turbine Trip event that took place in Kern Kraftwerk Leibstadt (KKL), at the fuel cycle 18. KKL is a Nuclear Power Plant (NPP) located in Leibstadt, Switzerland. This NPP operates with a BWR developing 3600 MWt in fuel cycles of one year. The Turbine Trip is a fast transient developing a pressure peak in the reactor followed by a power decreasing due to the selected control rod insertion. This kind of transient is very useful to check the feedback performance between both coupled codes due to the fast

Reactor safety protection system

International Nuclear Information System (INIS)

Nishi, Hiroshi; Yokoyama, Tsuguo.

1989-01-01

A plurality of neutron detectors are disposed around a reactor core and detection signals from optional two neutron detectors are inputted into a ratio calculation device. If the ratio between both of the neutron flux level signals exceeds a predetermined value, a reactor trip signal is generated from an alarm setting device. Further, detection signals from all of the neutron detection devices are inputted into an average calculation device and the reactor trip signal is generated also in a case where the average value exceeds a predetermined set value. That is, when the reactor core power is increased locally, the detection signal from the neutron detector nearer to the point of power increase is greater than the increase rate for the entire reactor core power, while the detection signal from the neutron detector remote from the point of power increase is smaller. Thus, the local power increase ratio in the FBR reactor core can be detected efficiently by calculating the ratio for the neutron flux level signals from two neutron detectors, thereby enabling to exactly recognize the local power increase rate in the reactor core. (N.H.)

Prism reactor system design and analysis of postulated unscrammed events

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.C.

1991-08-01

Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

PRISM reactor system design and analysis of postulated unscrammed events

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.C.; Rosztoczy, Z.; Lane, J.

1991-01-01

Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristics and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

RF Trip and Beam Loss Diagnostics in LEP using GPS timing

CERN Document Server

Arnaudon, L; Beetham, G; Ciapala, Edmond; Juillard, J C; Olsen, R; CERN. Geneva. SPS and LEP Division

2000-01-01

A fast diagnostics system has been installed in LEP to allow precise time-stamping of RF unit trips. The system also monitors the fast decay of current when a beam loss occurs. From the information gathered it is now possible to determine which RF units have provoked a beam loss at high energy and which have tripped as a result. The system uses GPS equipment installed at all of the even points of LEP together with fast local DSP acquisition and event recording units in each RF sector. An overall control application driven by the LEPExec arms the system at the start of each fill, calculates and displays RF and trip beam loss events in sequence, then stores the results in a database. The system installation was completed in time for the LEP 2000 startup and initial problems were quickly resolved. Throughout the year it has proved invaluable for high energy running. The experience gained will also be very useful for similar diagnostics applications in LHC.

Safety design of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)

International Nuclear Information System (INIS)

Takahashi, Minoru; Uchida, Shoji; Yamada, Yumi; Koyama, Kazuya

2008-01-01

In Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb-Bi coolant above the core, and Pb-Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb-Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. (author)

Initiating events and accidental sequences taken into account in the CAREM reactor design

International Nuclear Information System (INIS)

Kay, J.M.; Felizia, E.R.; Navarro, N.R.; Caruso, G.J.

1990-01-01

The advance made in the nuclear security evaluation of the CAREM reactor is presented. It was carried out using the Security Probabilistic Analysis (SPA). The latter takes into account the different phases of identification and solution of initiating events and the qualitative development of event trees. The method of identification of initiating events is the Master Logical Diagram (MLD), whose deductive basis makes it appropriate for a new design like the one described. The qualitative development of the event trees associated to the identified initiating events, allows identification of those accidental sequences which are to have the security systems in the reactor. (Author) [es

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

Directory of Open Access Journals (Sweden)

Matthew Bucknor

2017-03-01

Full Text Available Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general for the postulated transient event.

Advanced reactor passive system reliability demonstration analysis for an external event

Energy Technology Data Exchange (ETDEWEB)

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin [Argonne National Laboratory, Argonne (United States)

2017-03-15

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

Advanced reactor passive system reliability demonstration analysis for an external event

International Nuclear Information System (INIS)

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin

2017-01-01

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event

Field Trips. Beginnings Workshop.

Science.gov (United States)

Cartwright, Sally; Aronson, Susan S.; Stacey, Susan; Winbush, Olga

2001-01-01

Five articles highlight benefits and organization of field trips: (1) "Field Trips Promote Child Learning at Its Best"; (2) "Planning for Maximum Benefit, Minimum Risk"; (3) "Coaching Community Hosts"; (4) "The Story of a Field Trip: Trash and Its Place within Children's Learning and Community"; and (5) "Field Trip Stories and Perspectives" (from…

Kuosheng BWR/6 recirculation pump trip transient analysis with the RETRAN02/MOD5 code

International Nuclear Information System (INIS)

Wang, J.R.; Shih, C.

1992-01-01

A recirculation pump trip (RPT) event results in a reduction in recirculation flow, which reduces the core coolant flow rate. A reduction in core flow results in an increase in core void fraction and hence a decrease in core power due to negative void reactivity feedback. Although this category of events is less severe than others and generally considered as nonlimiting, core instability still may occur such as that at LaSalle on March 9, 1988. This paper focuses on the RPT transient analysis of Kuosheng Nuclear Power Plant (KNPP), which has two units of General Electric-designed boiling water reactor (BWR)/6 with rated core thermal power of 2894 MW and rated core flow of 10645 kg/s (23472 lb m /s). The approach to investigating the RPT transient of KNPP consists of two steps. The first step is to develop a plant-specific model using the RETRAN02/MOD5 code. In this step, various plant-specific information, including design documentation, drawings, safety analysis reports, and other information supplied by vendors were collected for model development. The RPT startup test at 68% power was used for system model benchmarking to ensure the adequacy of this model and identify several sensitive parameters. The second step is to assess whether similar power oscillation phenomena may occur at KNPP because of an RPT with isolated feedwater heater event. Two transient analyses (with or without reactor scram) of the KNPP RPT with isolated feedwater heater were investigated

Reactor protection systems for the Replacement Research Reactor, ANSTO

International Nuclear Information System (INIS)

Morris, C.R.

2003-01-01

The 20-MW Replacement Research Reactor Project which is currently under construction at ANSTO will have a combination of a state of the art triplicated computer based reactor protection system, and a fully independent, and diverse, triplicated analogue reactor protection system, that has been in use in the nuclear industry, for many decades. The First Reactor Protection System (FRPS) consists of a Triconex triplicated modular redundant system that has recently been approved by the USNRC for use in the USA?s power reactor program. The Second Reactor Protection System is a hardwired analogue system supplied by Foxboro, the Spec 200 system, which is also Class1E qualified. The FRPS is used to drop the control rods when its safety parameter setpoints have been reached. The SRPS is used to drain the reflector tank and since this operation would result in a reactor poison out due to the time it would take to refill the tank the FRPS trip setpoints are more limiting. The FRPS and SRPS have limited hardwired indications on the control panels in the main control room (MCR) and emergency control centre (ECC), however all FRPS and SRPS parameters are capable of being displayed on the reactor control and monitoring system (RCMS) video display units. The RCMS is a Foxboro Series I/A control system which is used for plant control and monitoring and as a protection system for the cold neutron source. This paper will provide technical information on both systems, their trip logics, their interconnections with each other, and their integration into the reactor control and monitoring system and control panels. (author)

Experience with reactor power cutback system at Palo Verde nuclear generating station

International Nuclear Information System (INIS)

Chari, D.R.; Rec, J.R.; Simoni, L.P.; Eimar, R.L.; Sowers, G.W.

1987-01-01

Palo Verde Nuclear Generating Station (PVNGS) is a three unit site which illustrates System 80 nuclear steam supply system (NSSS) design. The System 80 NSSS is the Combustion Engineering (C-E) standard design rated at 3817 Mwth. PVNGS Units 1 and 2 achieved commercial operation on February 13, 1986 and September 22, 1986, respectively, while Unit 3 has a forecast date for commercial operation in the third quarter of 1987. The System 80 design incorporates a reactor power cutback system (RPCS) feature which reduces plant trips caused by two common initiating events: loss of load/turbine trip (LOL) and loss of one main feedwater pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety system

RA-6 reactor's probabilistic safety evaluation. Identification and selection of starting events

International Nuclear Information System (INIS)

Kay, J.; Chiossi, C.; Felizia, E.; Vallerga, H.; Kalejman, G.; Navarro, R.; Caruso, G.J.

1987-01-01

A summary of the 'Identification and selection of starting events' stage of the previous probabilistic safety evaluation of RA-6 reactor is presented. This evaluation was performed to verify if the safety criteria required for the licensing of RA-6 are met and to promote the diffusion of its meaning and usefulness with educational purposes. At this stage the starting events of RA-6 are determined and the probability that such events occur is calculated. The identification and selection of starting events is performed in two steps: determination of proposed starting events and determination of postulated starting events. The proposed starting events are determined by means of the master logic diagram (MLD) method, while the postulated starting events are obtained by grouping the proposed starting events. The simplifying hypothesis required for the application of MLD to the reactor are also formulated. The probability that the proposed and postulated starting events occur is afterwards calculated, adopting different fault models, in accordance with the nature of events that are considered. Conservative hypothesis on the characteristics of these events and the uncertainty of parameter values of those models are also formulated. The numerical values of the above mentioned probabilities are obtained by giving the parameters suitable values that are extracted from specialized publications. (Author)

Safety Assessment for transient event occurred during the ASTS test of Hanbit Unit 2

International Nuclear Information System (INIS)

Yang, Changkeun; Kim, Yohan; Ha, Sangjun

2014-01-01

Safety Injection has been actuated during the ASTS (Automatic Seismic Trip System) test of Hanbit Unit 2 on Feb. 28, 2014. It could be bad effect on system integrity. KHNP has been performed safety assessment of system for effect of Safety Injection (SI) actuation occurred during the ASTS test of hanbit Unit 2. Stable state of nuclear power plant system has been confirmed according to Safety Injection and reactor trip event occurred during the ASTS test of hanbit Unit 2. In the result of system safety assessment, major variables of nuclear power plant are located in optimal range and not exceed safety limit. It remains nuclear fuel and the integrity of the power plant is in a safe condition were conformed. After ASTS action, thermal elimination has been processed throughout the turbine until turbine signal occurrence because ASTS is connected to M-G set in the present hanbit Unit 2. Therefore, Safety Injection signal has been actuated by rapid reduction of Steam Generator pressure. In this paper, it is concluded that consideration of equipment and setpoint is needed for that Safety Injection has been not occurred under the unnecessary situation. Stable state of nuclear power plant system has been confirmed for Safety Injection and reactor trip event occurred during the ASTS test of hanbit Unit 2. In the result of system safety assessment, major variables of nuclear power plant are located in optimal range and not exceed safety limit. It remains nuclear fuel and the integrity of the plant is in a safe condition were conformed. It is concluded that consideration of equipment and setpoint is needed for that Safety Injection has been not occurred under the unnecessary situation

Effects of delayed RCP trip during SBLOCA in PWR

International Nuclear Information System (INIS)

Montero-Mayorga, J.; Queral, C.; Gonzalez-Cadelo, J.

2014-01-01

Highlights: • Review of RCP trip issue in case of SBLOCA showing adequacy of present EOPs. • Risk assessment of a SBLOCA deterministic safety analysis by means of ISA methodology. • Evaluation of the probability of damage considering uncertainties in operator actuation times. • Application of ISA methodology to probabilistic safety analysis. • Obtaining of RCP trip available time as function of break size. - Abstract: After the Three Mile Island (TMI) accident, the issue of when to trip the Reactor Coolant Pumps (RCPs) in case of a Small Break Loss of Coolant Accident (SBLOCA) became very important. Several analyses were performed during the 1980s leading to the current Emergency Operating Procedures (EOPs). However these analyses have not been reviewed taking into account that several improvements have been performed in the last thirty years with respect to two phase-flow models, thermal–hydraulics codes and safety assessment methodologies. In this sense, this work has two main objectives: First of all, an assessment of the analyses carried out by Pressurizer Water Reactor (PWR) vendors after the TMI-2 accident with a model of Almaraz Nuclear Power Plant (NPP) for TRACE code (V 5.0 patch 1). On the other hand, Integrated Safety Assessment (ISA) methodology is applied to explore this matter. Such methodology has been developed by the Spanish Nuclear Safety Council (CSN) and it is an adequate method to perform analyses in nuclear safety in which the uncertainties in operator actuation time play an important role. The main conclusions obtained from this work are that, the current EOPs are adequate to manage a SBLOCA sequence in a suitable manner and that ISA methodology is a powerful tool that provides accurate information to the analyst in order to verify the robustness of the EOPs and to perform the safety assessment of both, deterministic and probabilistic safety analysis

Master Logic Diagram: An Approach to Identify Initiating Events of HTGRs

Science.gov (United States)

Purba, J. H.

2018-02-01

Initiating events of a nuclear power plant being evaluated need to be firstly identified prior to applying probabilistic safety assessment on that plant. Various types of master logic diagrams (MLDs) have been proposedforsearching initiating events of the next generation of nuclear power plants, which have limited data and operating experiences. Those MLDs are different in the number of steps or levels and different in the basis for developing them. This study proposed another type of MLD approach to find high temperature gas cooled reactor (HTGR) initiating events. It consists of five functional steps starting from the top event representing the final objective of the safety functions to the basic event representing the goal of the MLD development, which is an initiating event. The application of the proposed approach to search for two HTGR initiating events, i.e. power turbine generator trip and loss of offsite power, is provided. The results confirmed that the proposed MLD is feasiblefor finding HTGR initiating events.

Abnormal Events for Reactor System and Facilities in HANARO

Energy Technology Data Exchange (ETDEWEB)

Choi, Ho Young; Lee, B. H.; Lee, M.; Kang, I. H.; Lee, U. G.; Sin, H. C.; Park, C. Y.; Song, B. S.; Lee, S. H.; Han, J. S

2006-12-15

This report gathers abnormal events related to reactor system and facilities of HANARO that happened during its operation over 10 years since the first criticality on February 1995. The collected examples will be utilized to the HANARO's operators as a useful guide.

RIA Analysis of Unprotected TRIGA Reactor

Directory of Open Access Journals (Sweden)

M.H. Altaf

2017-07-01

Full Text Available An RIA (reactivity initiated accident analysis has been carried out for the TRIGA Mark II research reactor considering both step and ramp reactivity ranges within 0.5 % dk/k (< $1 to 2.0 % dk/k (>$2. The insertion time was set at 10 s. Based on the fact that a reactor becomes unprotected if scram does not work at the event of danger, to define unprotected conditions, the time to actuate scram (trip was taken as close to total simulation time. In this long duration of scram inactivity, it is obtained from the present analysis that the reactor remained safe to up to 1.8 % dk/k ($2.57 for step reactivity and 1.99 % dk/k ($2.84 for ramp reactivity. In addition to negative temperature coefficient of reativity, probably the longer time of reactivity insertion keeps TRIGA safe even at larger magnitudes of reactivity during unprotected reactor transients. Coupled point kinetics, neutronics, and thermal hydraulics code EUREKA-2/R has been utilized for this work. It appears that EUREKA-2/RR predicts the sequence of unprotected transient scenario of TRIGA core with good approximation and the results will definitely be helpful for the reactor operators.

Reactor feedwater pump control device

International Nuclear Information System (INIS)

Nishiyama, Hiroyuki.

1990-01-01

An amount of feedwater necessary for ensuring reactor inventory after scram is ensured automatically based on the reactor output before scram of a BWR type reactor. That is, if scram should occur, a feedwater flow rate just before the scram is stored by reactor output signals. Further, the amount of feedwater required after the scram is determined based on the output of the memory. The reactor power after the scram based on a feedwater flow rate and a main steam flow rate is inputted to an integrator, to calculate and output the amount of the feedwater flow rate (1) injected after the scram for the inventory. A coast down flowrate (2) in a case of pump trip is forecast by the output signals. Automatic trip is outputted to all turbine driving feedwater pumps when the sum of (1) and (2) exceeds a necessary and sufficient amount of feedwater required for ensuring inventory. For motor driving feedwater pumps, only a portion, for example, one of the pumps is automatically started while other pumps are stopped their operation, only in this case, to prevent excess water feeding. (I.S.)

Development of a cause analysis system for a CPCS trip by using the rule-base deduction method.

Science.gov (United States)

Park, Je-Yun; Koo, In-Soo; Sohn, Chang-Ho; Kim, Jung-Seon; Cho, Gi-Ho; Park, Hee-Seok

2009-07-01

A Core Protection Calculator System (CPCS) was developed to initiate a Reactor Trip under the circumstance of certain transients by a Combustion Engineering Company. The major function of the Core Protection Calculator System is to generate contact outputs for the Departure from Nucleate Boiling Ratio (DNBR) Trip and a Local Power Density (LPD) Trip. But in a Core Protection Calculator System, a trip cause cannot be identified, thus only trip signals are transferred to the Plant Protection System (PPS) and only the trip status is displayed. It could take a considerable amount of time and effort for a plant operator to analyze the trip causes of a Core Protection Calculator System. So, a Cause Analysis System for a Core Protection Calculator System (CASCPCS) has been developed by using the rule-base deduction method to assist operators in a Nuclear Power Plant. CASCPCS consists of three major parts. Inference engine has a role of controlling the searching knowledge base, executing the rules and tracking the inference process by using the depth-first searching method. Knowledge base consists of four major parts: rules, data base constants, trip buffer variables and causes. And a user interface is implemented by using menu-driven and window display techniques. The advantage of CASCPCS is that it saves time and effort to diagnose the trip causes of a Core Protection Calculator System, it increases a plant's availability and reliability, and it makes it easy to manage CASCPCS because of using only a cursor control.

Westinghouse Small Modular Reactor passive safety system response to postulated events

International Nuclear Information System (INIS)

Smith, M. C.; Wright, R. F.

2012-01-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000 R reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to

Loss of main and auxiliary feedwater event at the Davis-Besse Plant on June 9, 1985

International Nuclear Information System (INIS)

1985-07-01

On June 9, 1985, Toledo Edison Company's Davis-Besse Nuclear Power Plant, located in Ottawa County, Ohio, experienced a partial loss of feedwater while the plant was operating at 90% power. Following a reactor trip, a loss of all feedwater occurred. The event involved a number of equipment malfunctions and extensive operator actions, including operator actions outside the control room. Several operator errors also occurred during the event. This report documents the findings of an NRC Team sent to Davis-Besse by the NRC Executive Director for Operations in conformance with the staff-proposed Incident Investigation Program

Evidence Based Prevention of Occupational Slips, Trips and Falls

DEFF Research Database (Denmark)

Jensen, Olaf Chresten

2009-01-01

It is estimated that about one third of the compensated occupational injuries and half of the most serious occupational injuries in merchant seafaring are related to slips, trips and falls (STF)-events. Among the elderly, STF is the risk factor that causes the largest number of inpatient days...

Event management in research reactors

International Nuclear Information System (INIS)

Perrin, C.D.

2006-01-01

In the Radiological and Nuclear Safety field, the Nuclear Regulatory Authority of Argentina controls the activities of three investigation reactors and three critical groups, by means of evaluations, audits and inspections, in order to assure the execution of the requirements settled down in the Licenses of the facilities, in the regulatory standards and in the documentation of mandatory character in general. In this work one of the key strategies developed by the ARN to promote an appropriate level of radiological and nuclear safety, based on the control of the administration of the abnormal events that its could happen in the facilities is described. The established specific regulatory requirements in this respect and the activities developed in the entities operators are presented. (Author)

Environmental impacts of radiological consequences during the anticipated transients without scram (ATWS) events in nuclear power reactors

International Nuclear Information System (INIS)

El-Kafas, A.A.

2011-01-01

Anticipated transients without scram (ATWS), is one of the (worst case) accidents could happen if the system that provides a highly reliable means of shutting down the reactor (scram system )fails to work during a reactor event (anticipated transient).It has two general characteristics: (1) Initiation by a transient anticipated to occur one or more times in the life of reactor and ,(2) Assumed to proceed without scram.The types of events considered are those used for designing the plant .The evaluation of the radiological consequences during the assessment of the nuclear events,especially ATWS in nuclear power reactors, is very essential for environmental studies and public safety. In this paper, the root cases for nuclear events and dose calculation are presented. Scenario of accident sequences together with radiological impacts is illustrated for loss of coolant accident (LOCA) for a typical pressurized water reactor nuclear power plant. Recommendations for mitigating or preventing the release of radiation and high radioactive materials to environment are presented.

Reactor control device

International Nuclear Information System (INIS)

Fukami, Haruo; Morimoto, Yoshinori.

1981-01-01

Purpose: To operate a reactor always with safety operation while eliminating the danger of tripping. Constitution: In a reactor control device adapted to detect the process variants of a reactor, control a control rod drive controlling system based on the detected signal to thereby control the driving the control rods, control the reactor power and control the electric power generated from an electric generator by the output from the reactor, detection means is provided for the detection of the electric power from said electric generator, and a compensation device is provided for outputting control rod driving compensation signals to the control rod driving controlling system in accordance with the amount of variation in the detected value. (Seki, T.)

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

Validation of reactor core protection system

International Nuclear Information System (INIS)

Lee, Sang-Hoon; Bae, Jong-Sik; Baeg, Seung-Yeob; Cho, Chang-Ho; Kim, Chang-Ho; Kim, Sung-Ho; Kim, Hang-Bae; In, Wang-Kee; Park, Young-Ho

2008-01-01

Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

Fail-safe logic elements for use with reactor safety systems

International Nuclear Information System (INIS)

Bobis, J.P.; McDowell, W.P.

1976-01-01

A complete fail-safe trip circuit is described which utilizes fail-safe logic elements. The logic elements used are analog multipliers and active bandpass filter networks. These elements perform Boolean operations on a set of AC signals from the output of a reactor safety-channel trip comparator

Computer based systems for fast reactor core temperature monitoring and protection

International Nuclear Information System (INIS)

Wall, D.N.

1991-01-01

Self testing fail safe trip systems and guardlines have been developed using dynamic logic as a basis for temperature monitoring and temperature protection in the UK. The guardline and trip system have been tested in passive operation on a number of reactors and a pulse coded logic guardline is currently in use on the DIDO test reactor. Acoustic boiling noise and ultrasonic systems have been developed in the UK as diverse alternatives to using thermocouples for temperature monitoring and measurement. These systems have the advantage that they make remote monitoring possible but they rely on complex signal processing to achieve their output. The means of incorporating such systems within the self testing trip system architecture are explored and it is apparent that such systems, particularly that based on ultrasonics has great potential for development. There remain a number of problems requiring detailed investigation in particular the verification of the signal processing electronics and trip software. It is considered that these problems while difficult are far from insurmountable and this work should result in the production of protection and monitoring systems suitable for deployment on the fast reactor. 6 figs

Safety assessment of the advanced CANDU reactor in postulated LOCA/LOECC events

International Nuclear Information System (INIS)

Hazen Hezhi Fan; Zoran Bilanovic

2005-01-01

The Advanced CANDU Reactor TM (ACR TM ) retains the proven strengths and features of CANDU reactors, and incorporates innovative new features and state-of-the-art technology. In addition to the enhanced emergency core cooling system, the reserve water system is designed to be available to inject reserve water by gravity into the reactor inlet headers after a postulated loss-of-coolant accident (LOCA). To assist in the ACR design and analysis of beyond the design basis events, simulations are needed to demonstrate the effectiveness of these two independent systems on core cooling, and to assess the consequences of the postulated accident coincident with the impairment of either of the two systems. The current paper is subject to an assessment of a postulated large LOCA coincident with loss of the emergency core cooling (LOECC) system. A postulated LOCA/LOECC has very low probability, in the range usually associated with severe core damage events. However, in the CANDU design, including ACR, the presence of moderator water surrounding the fuel channels acts as an effective heat sink, together with other safety features, to prevents severe core damage following a postulated LOCA/LOECC. Therefore, it is possible to analyse LOCA/LOECC using the same deterministic tools that are used for analysis of events with much higher frequencies, in the design basis event range. The assessment is conducted based on the current ACR-700 design. However, the analysis methodology, scope, computer tools, and the results in principle, are applicable to larger ACR designs. This assessment includes system (circuit), fuel channel, and fuel analyses. Some assessment results are needed in subsequent moderator analysis and containment analysis. In the assessment, several simulations were performed to analyse the full circuit and individual fuel channel transient behaviours, as well as the fission product release behaviour. The assessment has captured the key responses of the reactor heat

Analysis of postulated events for the revised ALMR/PRISM design

International Nuclear Information System (INIS)

Slovik, G.C.; Van Tuyle, G.J.

1991-01-01

The PRISM reactor is presently under pre-application licensing review by the NRC, with Brookhaven National Laboratory (BNL) providing technical assistance. In this paper, the authors describe the latest set of results from their SSC and MINET code calculations. Series of postulated events were analyzed, for the most current PRISM design, to evaluate the system performance under unscrammed conditions. The PRISM reactor utilizes a metal fuel composed of 27% Pu, 10% Zr, and 63% U, based on the fuel developed by Argonne National Laboratory as part of the Integral Fast Reactor program. The fuel has a small power and temperature defect as a result of the high fuel thermal conductivity of metal and a hard neutron spectrum. The reactivity feedbacks of the core are designed to provide a negative response to off-normal, high temperature conditions, by invoking negative responses from the thermal expansion of the fuel, control rods, and core radial dimensions. The core restraint system incorporates the limited free bowing feature, which generates an outward bow of the in-core portion of the fuel assembly when a temperature gradient exits. The core also has three Gas Expansion Modules (GEMs) placed around the periphery, which will remove a combined worth of -69 cents of reactivity when the pumps are tripped from full power conditions. In evaluating this passive shutdown response, in which the PRISM reactor power should decrease significantly in response to overheated conditions, one considers three classes of unscrammed events. These include the unscrammed loss of flow (ULOF), loss of heat sink (ULOHS), and transient over power (UTOP) events

Effects of RCP trip when recovering HPSI during LOCA in a Westinghouse PWR

Energy Technology Data Exchange (ETDEWEB)

Montero-Mayorga, Javier, E-mail: fj.montero@alumnos.upm.es; Queral, César; Rivas-Lewicky, Julio; González-Cadelo, Juan

2014-12-15

Highlights: • If HPSI is recovered during SBLOCA and RCPs are tripped core damage can be reached. • If the RCPs are tripped once the accumulators have injected the damage can be avoided. • If only 2 out of 3 RCPs are tripped the damage can be also avoided. • Improvements are proposed to the EOPs in order to avoid possible damage. - Abstract: Current Westinghouse Emergency Operating Procedures (EOPs) indicate initially that the operator must keep the reactor coolant pumps (RCPs) running during a Small Break Loss of Coolant Accident (SBLOCA) if there is unavailability of high pressure safety injection (HPSI) system in order to cool the core by forced convection. However, the crew must follow different EOPs along the transient depending on its evolution. In these EOPs there are several conditions which indicate the necessity of tripping one or more RCPs when HPSI is recovered. In this paper the occurrence of a SBLOCA with unavailability of HPSI has been analyzed with a model of Almaraz Nuclear Power Plant (Westinghouse 3 Loop) for TRACE code V5.0 patch 1. Two different approaches have been considered: the first one, taking into account Optimal Recovery Guidelines (ORGs) and in the second approach, the transition to Function Restoration Guidelines (FRGs) due to inadequate core cooling (ICC) conditions is considered. Results of this paper lead to the implementation of an improvement in current EOPs regarding how many RCPs should be tripped during SBLOCA sequences.

20% inlet header break analysis of Advanced Heavy Water Reactor

International Nuclear Information System (INIS)

Srivastava, A.; Gupta, S.K.; Venkat Raj, V.; Singh, R.; Iyer, K.

2001-01-01

The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. A passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power levels, with no primary coolant pumps. Loss of coolant due to failure of inlet header results in depressurization of primary heat transport (PHT) system and containment pressure rise. Depressurization activates various protective and engineered safety systems like reactor trip, isolation condenser and advanced accumulator, limiting the consequences of the event. This paper discusses the thermal hydraulic transient analysis for evaluating the safety of the reactor, following 20% inlet header break using RELAP5/MOD3.2. For the analysis, the system is discretized appropriately to simulate possible flow reversal in one of the core paths during the transient. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, steam quality and flow in different parts of the Primary Heat Transport (PHT) system. Flow and energy discharges into the containment are also estimated for use in containment analysis. (author)

A microprocessor based monitoring system for a small nuclear reactor facility

International Nuclear Information System (INIS)

Miller, G.E.; DeKeyser, C.F.

1980-01-01

An inexpensive microprocessor based system has been designed and constructed for our 250 kilowatt TRIGA reactor facility. The system, which is beginning operational testing, can monitor on a continuous basis the status of up to 54 devices and maintain a record of events. These devices include fixed radiation monitors, pool water level trips, security alarms and an access control unit. In the latter case, the unit permits selection of different levels of access permission based on the time of day. The system can alert security and other personnel in the event of abnormalities. Because of the inclusion of this in the security system, special reliability and failure mode operation. The unit must also be simple to install, program and operate. (author)

TRIP RATES FOR CONDOMINIUM CONSTRUCTION PROJECT

Directory of Open Access Journals (Sweden)

Wirach Hirun

2015-01-01

Full Text Available The number of large scale condominium construction projects had dramatically increased in Bangkok. Many projects had occurred in either densely populated areas or in central business districts, where traffic conditions were usually highly congested. To prevent traffic problems, a traffic impact study must be prepared and submitted for review by concerned public authorities. Unit trip generation rates were important data in traffic impact analysis. Without accurate unit trip generation rates, public agencies could not obtain accurate information on the traffic that will be generated. This study aimed to study trip rates and the factors affecting them for condominium construction project in Bangkok. The data were collected from 30 condominium construction sites located in 15 districts of Bangkok. The analysis used the linear regression method and was divided into three cases: 1 trip rates for all vehicles, 2 trip rates for classified vehicles, and 3 trip rates for all types of condominium. All case analyses considered weekdays, Saturday, and Sunday. The results found that trip rates related to the number of dwellings in the condominium. The trip rates for all vehicle types on weekdays, Saturday, and Sunday were 10.636, 4.647, and 9.294 vehicles per 100 dwelling units per day respectively. The trip rates for six-wheeled and ten-wheeled trucks on weekdays, Saturday, and Sunday were 2.046, 0.975, and 0.575 vehicles per 100 dwelling units per day respectively. The trip rate for four-wheeled trucks and passenger cars on weekdays was 1.960. Regarding condominium types, the trip rate for low rise condominiums for all vehicle types on weekdays was 5.315 while the trip rates for high rise condominiums for weekdays, Saturday, and Sunday were 3.965, 2.667, and 1.261 respectively.

Event and fault tree model for reliability analysis of the greek research reactor

International Nuclear Information System (INIS)

Albuquerque, Tob R.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes

2013-01-01

Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This work uses the methods of fault tree (FT) and event tree (ET) to perform the Probabilistic Safety Assessment (PSA) in research reactors. The PSA according to IAEA (International Atomic Energy Agency) is divided into Level 1, Level 2 and level 3. At Level 1, conceptually safety systems act to prevent the accident, at Level 2, the accident occurred and seeks to minimize the consequences, known as stage management of the accident, and at Level 3 are determined consequences. This paper focuses on Level 1 studies, and searches through the acquisition of knowledge consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR - 1, was used as a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from there were developed the possible accident sequences, using event tree, which could lead damage to the core. Furthermore, for each of the affected systems, the possible accidents sequences were made fault tree and evaluated the probability of each event top of the FT. The studies were conducted using a commercial computational tool SAPHIRE. The results thus obtained, performance or failure to act of the systems analyzed were considered satisfactory. This work is directed to the Greek Research Reactor due to data availability. (author)

Event and fault tree model for reliability analysis of the greek research reactor

Energy Technology Data Exchange (ETDEWEB)

Albuquerque, Tob R.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes, E-mail: atalbuquerque@ien.gov.br, E-mail: btony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2013-07-01

Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This work uses the methods of fault tree (FT) and event tree (ET) to perform the Probabilistic Safety Assessment (PSA) in research reactors. The PSA according to IAEA (International Atomic Energy Agency) is divided into Level 1, Level 2 and level 3. At Level 1, conceptually safety systems act to prevent the accident, at Level 2, the accident occurred and seeks to minimize the consequences, known as stage management of the accident, and at Level 3 are determined consequences. This paper focuses on Level 1 studies, and searches through the acquisition of knowledge consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR - 1, was used as a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from there were developed the possible accident sequences, using event tree, which could lead damage to the core. Furthermore, for each of the affected systems, the possible accidents sequences were made fault tree and evaluated the probability of each event top of the FT. The studies were conducted using a commercial computational tool SAPHIRE. The results thus obtained, performance or failure to act of the systems analyzed were considered satisfactory. This work is directed to the Greek Research Reactor due to data availability. (author)

Applying Bayesian neural networks to separate neutrino events from backgrounds in reactor neutrino experiments

International Nuclear Information System (INIS)

Xu, Y; Meng, Y X; Xu, W W

2008-01-01

A toy detector has been designed to simulate central detectors in reactor neutrino experiments in the paper. The samples of neutrino events and three major backgrounds from the Monte-Carlo simulation of the toy detector are generated in the signal region. The Bayesian Neural Networks (BNN) are applied to separate neutrino events from backgrounds in reactor neutrino experiments. As a result, the most neutrino events and uncorrelated background events in the signal region can be identified with BNN, and the part events each of the fast neutron and 8 He/ 9 Li backgrounds in the signal region can be identified with BNN. Then, the signal to noise ratio in the signal region is enhanced with BNN. The neutrino discrimination increases with the increase of the neutrino rate in the training sample. However, the background discriminations decrease with the decrease of the background rate in the training sample

New flux detectors for CANDU 6 reactors

International Nuclear Information System (INIS)

Cuttler, J.M.; Medak, N.

1992-06-01

CANDU reactors utilize large numbers of in-core self-powered detectors for control and protection. In the original design, the detectors (coaxial cables) were wound on carrier tubes and immersed in the heavy water moderator. Failures occurred due to corrosion and other factors, and replacement was very costly because the assemblies were not designed with maintenance in mind. A new design was conceived based on straight detectors, of larger diameter, in a sealed package of individual 'well' tubes. This protected the detectors from hostile environments and enabled individual failed sensors to be replaced by inserting spares in vacant neighbouring tubes. The new design was made retrofittable to older CANDU reactors. Provision was made for on-line scanning of the core with a miniature fission chamber. The modified detectors were tested in a lengthy development program and found to exhibit superior performance to that of the original detectors. Most of the CANDU reactors have now adopted the new design. In the case of the Gentilly-2 and Point Lepreau reactors, advantage was taken of the opportunity to redesign the detector layout (using better codes and the increased flexibility in positioning detectors) to achieve better coverage of abnormal events, leading to higher trip setpoints and wider operating margins

Healthy Ride Trip Data

Data.gov (United States)

Allegheny County / City of Pittsburgh / Western PA Regional Data Center — A dataset that shows trips taken using the Healthy Ride system by quarter. The dataset includes bike number, membership type, trip start and end timestamp, and...

Comparative evaluation of recent water hammer events in light water reactors

International Nuclear Information System (INIS)

House, R.K.; Sursock, J.P.; Kim, J.H.

1987-01-01

Water hammer events that occurred in commercial U.S. light water reactors in the five-year period from 1981 to 1985 were surveyed, and a preliminary evaluation of the events was conducted. The information developed supplements a previous study which evaluated water hammer events in the twelve-year period from 1969 to 1981. The current study of water hammer events in the 1980's confirms that the rate of events remains relatively constant (less than 0.25 events per plant year) in both PWRs and BWRs. Although water hammer events are not normally considered a safety issue, the economic impact of the events on plant operations can be significant. One particular severe water hammer event is estimated to have cost the plant owner $10 million for repair and evaluation alone. A variety of key characteristics of the recent water hammer events are summarized to establish a basis for further study of preventative methods

A study on design of the trip computer for ECCS based on dynamic safety system

International Nuclear Information System (INIS)

Kim, Seog Nam

2000-02-01

The Emergency Core Cooling system in current nuclear power plants typically has a considerable number of complex functions and largely cumbersome operator interfaces. Functions for initiation, switch-over between various phases of operation, interlocks, monitoring, and alarming are usually performed by relay and analog comparator logic which is difficult to maintain and test. To improve problems of an analog based ECC (Emergency Core Cooling) System, the trip computer for ECCS based on Dynamic Safety System is implemented. The Dynamic Safety System (DSS) is a computer based reactor protection system that has fail-safe nature and performs a dynamic self-testing. The most important feature of the DSS is the introduction of test signal that send the system into a tripped state. The test signals are interleaved between the plant signals to produce an output which switches between a tripped and health state. The dynamic operation is a key feature of the failsafe design of the system. In this thesis, a possible implementation of the DSS using PLC is presented for a CANDU reactor. ECC System of the CANDU Reactor is selected as the reference system. The function of the DSS is implemented In PLC with the CONCEPT language. CONCEPT was developed by GROUPE SCHNEIDER as a graphic user interface programming tool for the Quantum PLC. A MMI display for ECCS based on DSS is implemented with LOOKOUT as an object driven programming tool. The Validation test has been performed by S/W Input Simulator as per Validation Test Procedure. The result of the test was checked and displayed on the MMI display. From the test results, it is shown that the DSS based ECC System operates correctly in all conditions

Primary coolant pipe rupture event in liquid metal cooled reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2004-08-01

In liquid-metal cooled fast reactors (LMFR) the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). However, the primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors (Indira Gandhi Centre for Atomic Research, Kalpakkam, India, 13-17 January 2003) was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the technical meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the technical meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

Upgrading the electrical system of the IEA-R1 reactor to avoid triggering event of accidents

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2015-01-01

The IEA-R1 research reactor at the Institute of Energy and Nuclear Research (IPEN) is a research reactor open pool type, built and designed by the American firm 'Babcox and Wilcox', having as coolant and moderator demineralized light water and Beryllium and graphite, as reflectors. The power supply system is designed to meet the electricity demand required by the loads of the reactor (Security systems and systems not related to security) in different situations the plant can meet, such as during startup, normal operation at power, shutdown, maintenance, exchange of fuel elements and accident situations. Studies have been done on possible accident initiating events and deterministic techniques were applied to assess the consequences of such incidents. Thus, the methods used to identify and select the accident initiating events, the methods of analysis of accidents, including sequence of events, transient analysis and radiological consequences, have been described. Finally, acceptance criteria of radiological doses are described. Only a brief summary of the item concerning loss of electrical power will be presented. The loss of normal electrical power at the IEA-R1 reactor is very common. In the case of Electric External Power Loss, at the IEA-R1 reactor building, there may be different sequences of events, as described below. When the supply of external energy in the IEA-R1 facility fails, the Electrical Distribution Vital System, consisting of 4 (four) generators type 'UPS', starts operation, immediately and it will continue supplying power to the reactor control table, core cooling system and other security systems. To contribute to security, in the electric power failure, starts to operate the Emergency Cooling System (SRE). SRE has the function of removing residual heat from the core to prevent the melting of fuel elements in the event of loss of refrigerant to the core. Adding to the generators with batteries group system, new auxiliary

Assessment of full power turbine trip start-up test for C. Trillo 1 with RELAP5/MOD2

International Nuclear Information System (INIS)

Lozano, M.F.; Moreno, P.; de la Cal, C.; Larrea, E.; Lopez, A.; Santamaria, J.G.; Lopez, E.; Novo, M.

1993-07-01

C. Trillo I has developed a model of the plant with RELAP5/MOD2/36.04. This model will be validated against a selected set of start-up tests. One of the transients selected to that aim is the turbine trip, which presents very specific characteristics that make it significantly different from the same transient in other PWRs of different design, the main difference being that the reactor is not tripped: a reduction in primary power is carried out instead. Pre-test calculations were done of the Turbine Trip Test and compared against the actual test. Minor problems in the first model, specially in the Control and Limitation Systems, were identified and post-test calculations had been carried out. The results show a good agreement with data for all the compared variables

RELAP5/MOD3.3 Analysis of the Loss of External Power Event with Safety Injection Actuation

Directory of Open Access Journals (Sweden)

Andrej Prošek

2018-01-01

Full Text Available The code assessment typically comprises basic tests cases, separate effects test, and integral effects tests. On the other hand, the thermal hydraulic system codes like RELAP5/MOD3.3 are primarily intended for simulation of transients and accidents in light water reactors. The plant measured data come mostly from startup tests and operational events. Also, for operational events the measured plant data may not be sufficient to explain all details of the event. The purpose of this study was therefore besides code assessment to demonstrate that simulations can be very beneficial for deep understanding of the plant response and further corrective measures. The abnormal event with reactor trip and safety injection signal actuation was simulated with the latest RELAP5/MOD3.3 Patch 05 best-estimate thermal hydraulic computer code. The measured and simulated data agree well considering the major plant system responses and operator actions. This suggests that the RELAP5 code simulation is good representative of the plant response and can complement not available information from plant measured data. In such a way, an event can be better understood.

Model with Peach Bottom Turbine trip and thermal-Hydraulic code TRACE V5P3

International Nuclear Information System (INIS)

Mesado, C.; Miro, R.; Barrachina, T.; Verdu, G.

2014-01-01

This work is the continuation of the work presented previously in the thirty-ninth meeting annual of the Spanish Nuclear society. The semi-automatic translation of the Thermo-hydraulic model TRAC-BF1 Peach Bottom Turbine Trip to TRACE was presented in such work. This article is intended to validate the model obtained in TRACE, why compare the model results result from the translation with the Benchmark results: NEA/OECD BWR Peach Bottom Turbine Trip (PBTT), in particular is of the extreme scenario 2 of exercise 3, in which there is SCRAM in the reactor. Among other data present in the (transitional) Benchmark , are: total power, axial profile of power, pressure Dome, total reactivity and its components. (Author)

Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors (U)

International Nuclear Information System (INIS)

Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

1990-01-01

This paper reports on a full-scope probabilistic risk assessment (PRA) performed for the Savannah River Site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident

Trip internalization in multi-use developments.

Science.gov (United States)

2014-04-01

Internal trip capture refers to how the number trips to and from a development are reduced by the proximity of : complementary land uses within the development (e.g., residential to retail). Internal trips occur within the : development and do not en...

Trace coupled with PARCS benchmark against Leibstadt plant data during the turbine trip test

Energy Technology Data Exchange (ETDEWEB)

Sekhri, Abdelkrim; Baumann, Peter, E-mail: abdelkrim.sekhri@kkl.ch, E-mail: peter.Baumann@kkl.ch [KernkraftwerkLeibstadt AG, Leibstadt (Switzerland); Hidalga, Patricio; Morera, Daniel; Miro, Rafael; Barrachina, Teresa; Verdu, Gumersindo, E-mail: pathigar@etsii.upv.es, E-mail: dmorera@isirym.upv.es, E-mail: rmiro@isirym.upv.es, E-mail: tbarrachina@isirym.upv.es, E-mail: gverdu@isirym.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV), Valencia, (Spain). Institute for Industrial, Radiophysical and Environmental Safety

2013-07-01

In order to enhance the modeling of Nuclear Power Plant Leibstadt (KKL), the coupling of 3D neutron kinetics PARCS code with TRACE has been developed. To test its performance a complex transient of Turbine Trip has been simulated comparing the results with the existing plant data of Turbine Trip test. For this transient also Cross Sections have been generated and used by PARCS. The thermal-hydraulic TRACE model is retrieved from the already existing model. For the benchmarking the Turbine Trip transient has been simulated according to the test resulting in the closure of the turbine control valve (TCV) and the following opening of the bypass valve (TBV). This transient caused a pressure shock wave towards the Reactor Pressure Vessel (RPV) which provoked the decreasing of the void level and the consequent slight power excursion. The power control capacity of the system showed a good response with the procedure of a Selected Rod Insertion (SRI) and the recirculation loops performance which resulted in the proper thermal power reduction comparable to APRM data recorder from the plant. The comparison with plant data shows good agreement in general and assesses the performance of the coupled model. Due to this, it can be concluded that the coupling of PARCS and TRACE codes in addition with the Cross Section used works successfully for simulating the behavior of the reactor core during complex plant transients. Nevertheless the TRACE model shall be improved and the core neutronics corresponding to the test shall be used in the future to allow quantitative comparison between TRACE and plant recorded data. (author)

Steam generating system in LMFBR type reactors

International Nuclear Information System (INIS)

Kurosawa, Katsutoshi.

1984-01-01

Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

Development of advanced automatic control system for nuclear ship. 2. Perfect automatic operation after reactor scram events

International Nuclear Information System (INIS)

Yabuuchi, Noriaki; Nakazawa, Toshio; Takahashi, Hiroki; Shimazaki, Junya; Hoshi, Tsutao

1997-11-01

An automatic operation system has been developed for the purpose of realizing a perfect automatic plant operation after reactor scram events. The goal of the automatic operation after a reactor scram event is to bring the reactor hot stand-by condition automatically. The basic functions of this system are as follows; to monitor actions of the equipments of safety actions after a reactor scram, to control necessary control equipments to bring a reactor to a hot stand-by condition automatically, and to energize a decay heat removal system. The performance evaluation on this system was carried out by comparing the results using to Nuclear Ship Engineering Simulation System (NESSY) and the those measured in the scram test of the nuclear ship 'Mutsu'. As the result, it was showed that this system had the sufficient performance to bring a reactor to a hot syand-by condition quickly and safety. (author)

Development of advanced automatic control system for nuclear ship. 2. Perfect automatic operation after reactor scram events

Energy Technology Data Exchange (ETDEWEB)

Yabuuchi, Noriaki; Nakazawa, Toshio; Takahashi, Hiroki; Shimazaki, Junya; Hoshi, Tsutao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-11-01

An automatic operation system has been developed for the purpose of realizing a perfect automatic plant operation after reactor scram events. The goal of the automatic operation after a reactor scram event is to bring the reactor hot stand-by condition automatically. The basic functions of this system are as follows; to monitor actions of the equipments of safety actions after a reactor scram, to control necessary control equipments to bring a reactor to a hot stand-by condition automatically, and to energize a decay heat removal system. The performance evaluation on this system was carried out by comparing the results using to Nuclear Ship Engineering Simulation System (NESSY) and the those measured in the scram test of the nuclear ship `Mutsu`. As the result, it was showed that this system had the sufficient performance to bring a reactor to a hot syand-by condition quickly and safety. (author)

Gait Characteristics Associated with Trip-Induced Falls on Level and Sloped Irregular Surfaces

Directory of Open Access Journals (Sweden)

Andrew Merryweather

2011-11-01

Full Text Available Same level falls continue to contribute to an alarming number of slip/trip/fall injuries in the mining workforce. The objective of this study was to investigate how walking on different surface types and transverse slopes influences gait parameters that may be associated with a trip event. Gait analysis was performed for ten subjects on two orientations (level and sloped on smooth, hard surface (control and irregular (gravel, larger rocks surfaces. Walking on irregular surfaces significantly increased toe clearance compared to walking on the smooth surface. There was a significant (p < 0.05 decrease in cadence (steps/min, stride length (m, and speed (m/s from control to gravel to larger rocks. Significant changes in external rotation and increased knee flexion while walking on irregular surfaces were observed. Toe and heel clearance requirements increased on irregular surfaces, which may provide an explanation for trip-induced falls; however, the gait alterations observed in the experienced workers used as subjects would likely improve stability and recovery from a trip.

Verification of RBMK-1500 reactor main circulation circuit model with Cathare V1.3L

International Nuclear Information System (INIS)

Jasiulevicius, A.

2001-01-01

Among other computer codes, French code CATHARE is also applied for RBMK reactor calculations. In this paper results of such application for Ignalina NPP reactor (RBMK-1500 type) main circulation circuit are presented. Three transients calculations were performed: all main circulation pumps (MCP) trip, trip of one main circulation pump and trip of one main circulation pump without a closure of check valve on the pump line. Calculation results were compared to data from the Ignalina NPP, where all these transients were recorded in the years 1986, 1996 and 1998. The presented studies prove the capability of the CATHARE code to treat thermal-hydraulic transients with a reactor scram in the RBMK, in case of single or multiple pump trips. However, the presented model needs further improvements in order to simulate loss of coolant accidents. For this reason, emergency core cooling system should be included in the model. Additional model improvement is also needed in order to gain more independent pressure behavior in both loops. Also, flow rates through the reactor channels should be modeled by dividing channels into several groups, referring to channel power (in RBMK power produced in a channel, located in different parts of the core is not the same). The point-neutron kinetic model of the CATHARE code is not suitable to predict transients when the reactor is operating at a nominal power level. Such transients would require the use of 3D-neutron kinetics model to describe properly the strong space-time effect on the power distribution in the reactor core

Operation of the main feedwater system turbopump following plant trip with total failure of the auxiliary feedwater system

International Nuclear Information System (INIS)

Lucas Alvaro, A.M. de; Rosa Martinez, B. de la; Alcaide, F.; Toledano Camara, C.

1993-01-01

The Auxiliary Feedwater System (AF) is a safeguard system which has been designed to supply feedwater to the steam generators, cool the primary system and remove decay heat from the reactor when the main feedwater pumps fail due to loss of power or any other reason. Thus, when plant trip occurs, the AF system pumps start up automatically, allowing removal of decay heat from the reactor. However, even though this system (2 motor-driven pumps and 1 turbopump) is highly reliable, injection of water to the steam generators must be ensured when it fails completely. To do this, if plant trip has not been caused by loss of off site power or failure of the Main Feedwater System (FW) turbopumps, one of these turbopumps can be used to achieve removal of decay heat. Since a large amount of steam is consumed by these turbopumps, an analysis has been performed to determine whether one of these pumps can be used and what actions are necessary to inject water into the steam generators. Results show that, for the case in question, a FW turbopump can be used to remove decay heat from the reactor. (author)

Systematic evaluation program review of NRC Safety Topic VI-10.A associated with the electrical, instrumentation and control portions of the testing of reactor trip system and engineered safety features, including response time for the Dresden station, Unit II nuclear power plant

International Nuclear Information System (INIS)

St Leger-Barter, G.

1980-11-01

This report documents the technical evaluation and review of NRC Safety Topic VI-10.A, associated with the electrical, instrumentation, and control portions of the testing of reactor trip systems and engineered safety features including response time for the Dresden II nuclear power plant, using current licensing criteria

Human error probability evaluation as part of reliability analysis of digital protection system of advanced pressurized water reactor - APR 1400

International Nuclear Information System (INIS)

Varde, P. V.; Lee, D. Y.; Han, J. B.

2003-03-01

A case of study on human reliability analysis has been performed as part of reliability analysis of digital protection system of the reactor automatically actuates the shutdown system of the reactor when demanded. However, the safety analysis takes credit for operator action as a diverse mean for tripping the reactor for, though a low probability, ATWS scenario. Based on the available information two cases, viz., human error in tripping the reactor and calibration error for instrumentations in protection system, have been analyzed. Wherever applicable a parametric study has also been performed

analysis and implementation of reactor protection system circuits - case study Egypt's 2 nd research reactor-

International Nuclear Information System (INIS)

Elnokity, O.E.M.

2006-01-01

this work presents a way to design and implement the trip unit of a reactor protection system (RPS) using a field programmable gate arrays (FPGA). instead of the traditional embedded microprocessor based interface design method, a proposed tailor made FPGA based circuit is built to substitute the trip unit (TU), which is used in Egypt's 2 nd research reactor ETRR-2. the existing embedded system is built around the STD32 field computer bus which is used in industrial and process control applications. it is modular, rugged, reliable, and easy-to-use and is able to support a large mix of I/O cards and to easily change its configuration in the future. therefore, the same bus is still used in the proposed design. the state machine of this bus is designed based around its timing diagrams and implemented in VHDL to interface the designed TU circuit

Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors

International Nuclear Information System (INIS)

Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

1990-01-01

A full-scope probabilistic risk assessment (PRA) is being performed for the Savannah River site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident. The SRS PRA has three principal objectives: improved understanding of SRS reactor safety issues through discovery and understanding of the mechanisms involved. Improved risk management capability through tools for assessing the safety impact of both current standard operations and proposed revisions. A quantitative measure of the risks posed by SRS reactor operation to employees and the general public, to allow comparison with declared goals and other societal risks

Feasible reactor power cutback logic development for an integral reactor

International Nuclear Information System (INIS)

Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

2013-01-01

Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

Reactor Shutdown Mechanism by Top-mounted Hydraulic System

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Haun; Cho, Yeong Garp; Choi, Myoung Hwan; Lee, Jin Haeng; Huh, Hyung; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

There are two types of reactor shutdown mechanisms in HANARO. One is the mechanism driven by a hydraulic system, and the other is driven by a stepping motor. In HANARO, there are four Control Rod Drive Mechanisms (CRDMs) with an individual step motor and four Shutoff (SO) Units with an individual hydraulic system located at the top of reactor pool. The absorber rods in SO units are poised at the top of the core by the hydraulic force during normal operation. The rods of SO units drop by gravity as the first reactor showdown mechanism when a trip is commended by the reactor protection system (RPS). The rods in CRDMs also drop by gravity together as a redundant shutdown mechanism. When a trip is commended by the reactor regulating system (RRS), the absorber rods of CRDM only drop; while the absorber rods of SO units stay at the top of the core by the hydraulic system. The reactivity control mechanisms of in JRTR, one of the new research reactor with plate type fuels, consist of four CRDMs driven by an individual step motor and two second shutdown drive mechanisms (SSDMs) driven by an individual hydraulic system as shown in Fig. 1. The CRDMs act as the first reactor shutdown mechanism and reactor regulating as well. The top-mounted SSDM driven by the hydraulic system for the JRTR is under design in KAERI. The SSDM provides an alternate and independent means of reactor shutdown. The second shutdown rods (SSRs) of the SSDM are poised at the top of the core by the hydraulic system during the normal operation and drop by gravity for the reactor trip. Based on the proven technology of the design, operation and maintenance for HANARO, the SSDM for the JRTR has been optimized by the design improvement from the experience and test. This paper aims for the introduction of the SSDM in the process of the basic design. The major differences of the shutdown mechanisms by the hydraulic system are compared between HANARO and JRTR, and the design features, system, structure and

Simulation of a hypothetical liquid relief valve failure (open) at Embalse nuclear power plant when a reactor shutdown is considered; Simulacion de la evolucion de la CNE (central nuclear Embalse) en el caso hipotetico de la apertura espuria de una valvula de alivio liquido con disparo del reactor

Energy Technology Data Exchange (ETDEWEB)

Bedrossian, G; Gersberg, S [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

1997-12-31

The study of the spurious opening of the liquid relief valves is of great interest in CANDU nuclear power plants because this could lead to a loss of coolant through the degasser-condenser relief valves, and implies an undesirable intermittent opening/closure of them. In fact, there is a specific procedure to follow at Embalse nuclear power plant whenever this abnormal situation occurs. This procedure contains a section where a reactor trip is considered. Really, automatic reactor trip is not accepted to occur. No trip parameters set points are through to be reached (neutronic or process). However, the procedure considers the situation where the reactor does trip. We analyzed the plant behavior when a reactor shutdown is triggered. Our objective was to assess if after this trip, the procedure can lead the plant to a safe situation, preventing high pressures in the degasser-condenser and with the inventory recovered in the storage tank. The case was analyzed with Firebird III, Mod. 1.0 code. Two situations were considered: trip at 40 sec. and trip at 180 sec. after the liquid relief valve failed opened (the latter when the degasser-condenser fills up). Procedure analysis and code simulations showed that following the steps recommended, provided the liquid relief valve can be closed manually, the inventory that enters the degasser-condenser from the heat transport primary system through the failed valve could be recovered in the storage tank, leading the plant to shutdown in safe conditions, and preventing the degasser-condenser relief valves setpoint from being reached. (author). 3 refs., 10 figs.

Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2014-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

Guam Commercial Purchases (Trip Ticket)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — DAWR collects Trip Ticket or purchase invoice data from vendors that buy fish directly from the fishermen. Similar to the trip ticket system in Saipan, this is a...

Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

International Nuclear Information System (INIS)

2003-01-01

In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

Automated testing of reactor protection instrumentation made easy

International Nuclear Information System (INIS)

Iborra, A.; De Marcos, F.; Pastor, J.A.; Alvarez, B.; Jimenez, A.; Mesa, E.; Alsonso, L.; Regidor, J.J.

1997-01-01

Maintenance and testing of reactor protection systems is an important cause of unplanned reactor trips. Automated testing is the answer because it minimises test times and reduces human error. The GAMA I system, developed and implemented at Vandellos II in Spain, has the added advantage that it uses visual programming, which means that changing the software does not need specialist programming skills. (author)

Numerical analysis on ingress-of-coolant events in fusion reactors with TRAC-PF1 code

International Nuclear Information System (INIS)

Ose, Yasuo; Takase, Kazuyuki; Akimoto, Hajime

2000-01-01

As for accident events related with thermal-hydraulics, in a fusion experimental reactor an ingress-of-coolant event (ICE) and a loss-of-vacuum-accident event (LOVA) should be considered. An integrated ICE/LOVA test apparatus is under planning in order to estimate quantitatively heat transfer and fluid flow characteristics under ICE and LOVA events. This study was carried out to predict numerically the thermal-hydraulic characteristics in fusion reactors at the ICE events before construction of the integrated ICE/LOVA test apparatus. The TRAC-PF1 code, which was originally developed for the thermal-hydraulic safety analysis in light water reactors, was used. The numerical analyses were performed for two kinds of system configuration with/without a pressure-suppression tank:the former for is investigation of the pressure rise characteristics and two-phase flow behavior; the latter for estimation of an effect of the pressure reduction due to the pressure-suppression tank. From the present analytical results, effects of the ingress water flow rate and vessel temperatures on the pressure rise ware clarified quantitatively. Furthermore, the pressure-rise suppression effect due to the vapor condensation in the pressure-suppression tank was predicted numerically. In addition, the useful information regarding to the design of the integrated ICE/LOVA test apparatus and the knowledge with respect to the effective usage of the TRAC-PF1 code were obtained through the present numerical study. (author)

The development on the methodology of the initiating event frequencies for liquid metal reactor KALIMER

International Nuclear Information System (INIS)

Jeong, K. S.; Yang, Z. A.; Ah, Y. B.; Jang, W. P.; Jeong, H. Y.; Ha, K. S.; Han, D. H.

2002-01-01

In this paper, the PSA methodology of PRISM,Light Water Reactor, Pressurized Heavy Water Reactor are analyzed and the methodology of Initiating Events for KALIMER are suggested. Also,the reliability assessment of assumptions for Pipes Corrosion Frequency is set up. The reliability assessment of Passive Safety System, one of Main Safety System of KALIMER, are discussed and analyzed

Fast reactor fuel pin behavior analyses in a LOF type transient event

International Nuclear Information System (INIS)

Mizuno, Tomoyasu; Koyama, Shin-ichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

2013-06-01

In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, fuel pin behavior calculations were made using the fast reactor fuel pin performance code CEDAR. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross and Stoute type gap conductance model and constant gap conductance model, which is used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross and Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross and Stoute type gap conductance model which is thought to be realistic. (author)

A Bayesian approach to unanticipated events frequency estimation in the decision making context of a nuclear research reactor facility

International Nuclear Information System (INIS)

Chatzidakis, S.; Staras, A.

2013-01-01

Highlights: • The Bayes’ theorem is employed to support the decision making process in a research reactor. • The intention is to calculate parameters related to unanticipated occurrence of events. • Frequency, posterior distribution and confidence limits are calculated. • The approach is demonstrated using two real-world numerical examples. • The approach can be used even if no failures have been observed. - Abstract: Research reactors are considered as multi-tasking environments having the multiple roles of commercial, research and training facilities. Yet, reactor managers have to make decisions, frequently with high economic impact, based on little available knowledge. A systematic approach employing the Bayes’ theorem is proposed to support the decision making process in a research reactor environment. This approach is characterized by low level complexity, appropriate for research reactor facilities. The methodology is demonstrated through the study of two characteristic events that lead to unanticipated system shutdown, namely the de-energization of the control rod magnet and the flapper valve opening. The results obtained demonstrate the suitability of the Bayesian approach in the decision making context when unanticipated events are considered

Development of RPS trip logic based on PLD technology

International Nuclear Information System (INIS)

Choi, Jong Gyun; Lee, Dong Young

2012-01-01

The majority of instrumentation and control (I and C) systems in today's nuclear power plants (NPPs) are based on analog technology. Thus, most existing I and C systems now face obsolescence problems. Existing NPPs have difficulty in repairing and replacing devices and boards during maintenance because manufacturers no longer produce the analog devices and boards used in the implemented I and C systems. Therefore, existing NPPs are replacing the obsolete analog I and C systems with advanced digital systems. New NPPs are also adopting digital I and C systems because the economic efficiencies and usability of the systems are higher than the analog I and C systems. Digital I and C systems are based on two technologies: a microprocessor based system in which software programs manage the required functions and a programmable logic device (PLD) based system in which programmable logic devices, such as field programmable gate arrays, manage the required functions. PLD based systems provide higher levels of performance compared with microprocessor based systems because PLD systems can process the data in parallel while microprocessor based systems process the data sequentially. In this research, a bistable trip logic in a reactor protection system (RPS) was developed using very high speed integrated circuits hardware description language (VHDL), which is a hardware description language used in electronic design to describe the behavior of the digital system. Functional verifications were also performed in order to verify that the bistable trip logic was designed correctly and satisfied the required specifications. For the functional verification, a random testing technique was adopted to generate test inputs for the bistable trip logic.

Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

International Nuclear Information System (INIS)

Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

2012-01-01

700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

Selection of design basis event for modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi

2016-06-01

Japan Atomic Energy Agency (JAEA) has been investigating safety requirements and basic approach of safety guidelines for modular High Temperature Gas-cooled Reactor (HTGR) aiming to increase internarial contribution for nuclear safety by developing an international HTGR safety standard under International Atomic Energy Agency. In this study, we investigate a deterministic approach to select design basis events utilizing information obtained from probabilistic approach. In addition, selections of design basis events are conducted for commercial HTGR designed by JAEA. As a result, an approach for selecting design basis event considering multiple failures of safety systems is established which has not been considered as design basis in the safety guideline for existing nuclear facility. Furthermore, selection of design basis events for commercial HTGR has completed. This report provides an approach and procedure for selecting design basis events of modular HTGR as well as selected events for the commercial HTGR, GTHTR300. (author)

Trend analysis and comparison of operators' human error events occurred at overseas and domestic nuclear power plants

International Nuclear Information System (INIS)

Takagawa, Kenichi

2006-01-01

Human errors by operators at overseas and domestic nuclear power plants during the period from 2002 to 2005 were compared and their trends analyzed. The most frequently cited cause of such errors was 'insufficient team monitoring' (inadequate superiors' and other crews' instructions and supervision) both at overseas and domestic plants, followed by 'insufficient self-checking' (lack of cautions by the operator himself). A comparison of the effects of the errors on the operations of plants in Japan and the United Sates showed that the drop in plant output and plant shutdowns at plants in Japan were approximately one-tenth of those in the United States. The ratio of automatic reactor trips to the total number of human errors reported is about 6% for both Japanese and American plants. Looking at changes in the incidence of human errors by years of occurrence, although a distinctive trend cannot be identified for domestic nuclear power plants due to insufficient reported cases, 'inadequate self-checking' as a factor contributing to human errors at overseas nuclear power plants has decreased significantly over the past four years. Regarding changes in the effects of human errors on the operations of plants during the four-year period, events leading to an automatic reactor trip have tended to increase at American plants. Conceivable factors behind this increasing tendency included lack of operating experience by a team (e.g., plant transients and reactor shutdowns and startups) and excessive dependence on training simulators. (author)

Detailed analysis of the ANO-2 turbine trip test

International Nuclear Information System (INIS)

McDonald, T.A.; Tessier, J.H.; Senda, Y.; Waterman, M.D.

1983-01-01

A RELAP5/MOD1 (Cycle 18) computer code simulation of the ANO-2 turbine trip test from 98% power level was performed for use in vendor code qualification studies. Results focused on potential improvements to simulation capabilities and plant data acquisition systems to provide meaningful comparisons between the calculations and the test data. The turbine trip test was selected because it resulted in an unplanned sequence of events that broadly affected the plant process systems and their controls. The pressurizer spray valve stuck open at an undetermined flow area, and an atmospheric dump valve remained stuck fully open while several atmospheric dump and secondary side safety valves were unavailable throughout. Thus, although the plant remained always in a safe condition, this transient potentially provided an unusual set of data against which the fidelity of a NSSS simulation by RELAP5/MOD1 along with certain vendor analysis codes might be judged

A retrospective look on plant events for prospective affirmation of nuclear safety

International Nuclear Information System (INIS)

Koshy, Thomas; Khamis, Ibrahim

2014-01-01

The nuclear industry continues to rise above the challenges resulting from major plant events around the world. It is important to study the significant events, develop solutions to overcome the vulnerabilities identified, and retain the lessons while technology evolves to the next generation. The historical Station-Black-Out needs to be examined further in a new dimension in the light of 'Fukushima type' events where normal AC power recovery in a reasonable period was not practical. The plants would need to incorporate diversity in emergency core cooling to account for a condition that inhibits electrical energy as a source of motive power. An electrical event in Sweden that propagated from an electrical switchyard resulted in two core cooling divisions disabled and consequently exacerbating the plant condition by opening the relief system for reactor coolant system and that significantly increased the probability for core damage. A minor spark in an electronic control system card in a US plant caused inadvertent emergency core cooling and disabled the Control Room Operators' capability to intervene and prevent the primary loop from getting completely filled. A renewed assessment is needed to address the following areas for advancing reactor safety in the new evolving generation of plants to advance safety from the event lessons of the past. - Evaluate the diversity in core cooling systems following loss of all AC power onsite - Confirm independence in Reactor Trip, Depressurization and Core and Containment cooling systems for sensors, power supplies and actuation systems - Evaluate the suitability of logic/control system failure mode resulting from power supply failures in instrument channels and/or divisions (Conduct Failure Mode and Effects Analysis for system, power supplies and components). (authors)

The digital reactor protection system for the instrumentation and control of reactor TRIGA PUSPATI (RTP)

International Nuclear Information System (INIS)

Nurfarhana Ayuni Joha; Izhar Abu Hussin; Mohd Idris Taib; Zareen Khan Abdul Jalil Khan

2010-01-01

Reactor Protection System (RPS) is important for Reactor Instrumentation and Control System. The RPS comprises all redundant electrical devices and circuitry involved in the generation of those initiating signals associated to the trip protective function. The instrumentation system for the RPS provides automatic protection signals against unsafe and improper reactor operation. The physical separation is provided for all of the redundant instrumentation systems to preserve redundancy. The safety protection systems using circuits composed of analog instruments and relays with relay contacts is difficult to realize from various reasons. Therefore, an application of digital technology can be said a logical conclusion also in the light of its functional superiority. (author)

Safety aspects of unplanned shutdowns and trips

International Nuclear Information System (INIS)

1986-05-01

The issue of unplanned shutdowns and trips is receiving increased attention worldwide in view of its importance to plant safety and availability. There exists significant variation in the number of forced shutdowns for nuclear power plants of the same type operating worldwide. The reduction of the frequency of these events will have safety benefits in terms of reducing the frequency of plant transients and the challenges to the safety systems, and the risks of possible incidents. This report provides an insight into the causes of unplanned shutdowns experienced in operating nuclear power plants worldwide, the good practices that have been found effective in minimizing their occurrence, and the measures that have been taken to reduce these events. Specific information on the experiences, approaches and practices of some countries in dealing with this issue is presented in Appendix A

Analysis of events related to cracks and leaks in the reactor coolant pressure boundary

Energy Technology Data Exchange (ETDEWEB)

Ballesteros, Antonio, E-mail: Antonio.Ballesteros-Avila@ec.europa.eu [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Sanda, Radian; Peinador, Miguel; Zerger, Benoit [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Negri, Patrice [IRSN: Institut de Radioprotection et de Sûreté Nucléaire (France); Wenke, Rainer [GRS: Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH (Germany)

2014-08-15

Highlights: • The important role of Operating Experience Feedback is emphasised. • Events relating to cracks and leaks in the reactor coolant pressure boundary are analysed. • A methodology for event investigation is described. • Some illustrative results of the analysis of events for specific components are presented. - Abstract: The presence of cracks and leaks in the reactor coolant pressure boundary may jeopardise the safe operation of nuclear power plants. Analysis of cracks and leaks related events is an important task for the prevention of their recurrence, which should be performed in the context of activities on Operating Experience Feedback. In response to this concern, the EU Clearinghouse operated by the JRC-IET supports and develops technical and scientific work to disseminate the lessons learned from past operating experience. In particular, concerning cracks and leaks, the studies carried out in collaboration with IRSN and GRS have allowed to identify the most sensitive areas to degradation in the plant primary system and to elaborate recommendations for upgrading the maintenance, ageing management and inspection programmes. An overview of the methodology used in the analysis of cracks and leaks related events is presented in this paper, together with the relevant results obtained in the study.

Event management in research reactors; Gestion de eventos en reactores de investigacion

Energy Technology Data Exchange (ETDEWEB)

Perrin, C D [Coordinador Reactores de Investigacion y Conjuntos Criticos, Autoridad Regulatoria Nuclear (Argentina)

2006-07-01

In the Radiological and Nuclear Safety field, the Nuclear Regulatory Authority of Argentina controls the activities of three investigation reactors and three critical groups, by means of evaluations, audits and inspections, in order to assure the execution of the requirements settled down in the Licenses of the facilities, in the regulatory standards and in the documentation of mandatory character in general. In this work one of the key strategies developed by the ARN to promote an appropriate level of radiological and nuclear safety, based on the control of the administration of the abnormal events that its could happen in the facilities is described. The established specific regulatory requirements in this respect and the activities developed in the entities operators are presented. (Author)

ICARUS trip

CERN Document Server

Caraban Gonzalez, Noemi

2017-01-01

It’s lived in two different countries and is about to make its way to a third. It’s the largest machine of its kind, designed to find extremely elusive particles and tell us more about them. Its pioneering technology is the blueprint for some of the most advanced science experiments in the world. And this summer, it will travel across the Atlantic Ocean to its new home (and its new mission) at the U.S. Department of Energy’s Fermi National Accelerator Laboratory. It’s called ICARUS, and you can follow its journey over land and sea with the help of an interactive map at IcarusTrip.fnal.gov (link is external), or on Facebook (link is external), Twitter (link is external) and Instagram (link is external) using the hashtag #IcarusTrip.

Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code

International Nuclear Information System (INIS)

Youn, Bum soo; Lee, Jong beom; Song, Dong soo; Ha, Sang jun

2014-01-01

The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available

Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code

Energy Technology Data Exchange (ETDEWEB)

Youn, Bum soo; Lee, Jong beom; Song, Dong soo; Ha, Sang jun [KHNP-CRI, Daejeon (Korea, Republic of)

2014-10-15

The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available.

Procedural Aspects of Compulsory Licensing Under TRIPS

DEFF Research Database (Denmark)

Wested, Jakob; Minssen, Timo

2017-01-01

and discussion addressed the framework and context for CL provided by the TRIPS convention. Both the specific requirements enshrined in TRIPS art 31 and the broader objectives and principles enshrined in TRIPS, e.g. transfer and dissemination of technology (art 7), protection of public health (art 8......In 2013, Indian authorities granted a compulsory license to NATCO Pharmaceuticals for a patented pharmaceutical product sold by Bayer. This decision raised several complex issues regarding the grant a CL and their consistency with the principles and objectives of TRIPS. Furthermore, in January 2017...

The return trip is felt shorter only postdictively: A psychophysiological study of the return trip effect [corrected].

Directory of Open Access Journals (Sweden)

Ryosuke Ozawa

Full Text Available The return trip often seems shorter than the outward trip even when the distance and actual time are identical. To date, studies on the return trip effect have failed to confirm its existence in a situation that is ecologically valid in terms of environment and duration. In addition, physiological influences as part of fundamental timing mechanisms in daily activities have not been investigated in the time perception literature. The present study compared round-trip and non-round-trip conditions in an ecological situation. Time estimation in real time and postdictive estimation were used to clarify the situations where the return trip effect occurs. Autonomic nervous system activity was evaluated from the electrocardiogram using the Lorenz plot to demonstrate the relationship between time perception and physiological indices. The results suggest that the return trip effect is caused only postdictively. Electrocardiographic analysis revealed that the two experimental conditions induced different responses in the autonomic nervous system, particularly in sympathetic nervous function, and that parasympathetic function correlated with postdictive timing. To account for the main findings, the discrepancy between the two time estimates is discussed in the light of timing strategies, i.e., prospective and retrospective timing, which reflect different emphasis on attention and memory processes. Also each timing method, i.e., the verbal estimation, production or comparative judgment, has different characteristics such as the quantification of duration in time units or knowledge of the target duration, which may be responsible for the discrepancy. The relationship between postdictive time estimation and the parasympathetic nervous system is also discussed.

Unusual occurrences in fast breeder test reactor

International Nuclear Information System (INIS)

Kapoor, R.P.; Srinivasan, G.; Ellappan, T.R.; Ramalingam, P.V.; Vasudevan, A.T.; Iyer, M.A.K.; Lee, S.M.; Bhoje, S.B.

2000-01-01

parameters initiating reactor trip and has encountered large number of trips since first criticality. The paper also highlights several modifications affected in safety related systems for improved performance and safety reviews to reduce the parameters initiating reactor trip. The lessons learnt from the analysis of these incidents and safety reviews have been significant not only in improving FBTR performance but also as an important input for the design of future fast reactors. (author)

Device for controlling a recirculation flow in a reactor

International Nuclear Information System (INIS)

Shida, Toichi; Tohei, Kazushige; Hirose, Masao; Nakamura, Hideo.

1976-01-01

Object: To provide an emergency cut-off valve in a recirculation system in a reactor to control the recirculation at the time of turbine trip or load cut-off, thereby relieving excessive increase in heat output of fuel. Structure: A recirculation pump is driven through a recirculation pump motor by an AC generator, which is driven by a driving motor through a fluid coupling, so that reactor water passes the emergency cut-off valve and recirculation flow stop valve and then passes a jet pump into the core. At the time of turbine trip or load cut-off, the emergency cut-off valve is closed by a hydraulic circuit, whereby core flow is merely decreased by 20 to 30% in a short period of time to restrain excessive increase in heat output. (Yoshino, Y.)

FIX-II/3025, BWR FIX-II Pump Trip Experiment 3025, Immediate Split Size Break

International Nuclear Information System (INIS)

NILSSON, Lars; GUSTAFSSON, Per-Ake; GUSTAFSON, Lennart; JANCZAK, Rajmund; OESTERLUNDH, Ingrid

1992-01-01

1 - Description of test facility: The FIX-II facility is a volume scaled 1:777 representation of a Swedish BWR with external pumps. The pressure vessel contains a 36 rod full length bundle and a spray condenser at the top to allow steady state operation. The downcomer, bypass channels and guide tube volumes are represented by external piping. The intact loop represents three of the four external reactor loops. The broken loop is constructed such that both guillotine breaks and split breaks may be simulated. The facility is equipped with ADS-simulation, but no ECCS injection are included. The FIX-II loop is also suited to investigate response of pump trips and MSIV closures in internal pump reactors. 2 - Description of test: Test 3025 simulates an intermediate size split break in one of the four main recirculation lines. The break area was 31 per cent of the scaled down pipe area of the reactor. The initial power of the 36-rod bundle was 3.38 MW, corresponding to the hot channel power of the reactor

Analysis of loss of flow events on Brazilian multipurpose reactor by RELAP5 code

International Nuclear Information System (INIS)

Soares, Humberto V.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Aronne, Ivan D.; Rezende, Guilherme P.

2011-01-01

The Brazilian Multipurpose Reactor (BMR) is currently being projected and analyzed. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly, with planar plates. BMR will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of BMR using the RELAP5 model and also three transient cases of loss of flow accident (LOFA), in the primary cooling system. A LOFA may arise through failures associated with the primary cooling system pumps or through events resulting in a decrease in the primary coolant flow with the primary cooling system pumps functioning normally. The cases presented in this paper are: primary cooling system pump shaft seizure, failure of one primary cooling system pump motor and failure of both primary cooling system pump motors. In the shaft seizure case, the flow reduction is sudden, with the blocking of the flow coast down The motor failure cases, deal with the failure of one or two pump motor due to, for example, malfunction or interruption of power and differently of the shaft seizure it can be observed the flow coast down provided by the pump inertia. It is shown that after all initiating events the reactor reaches a safe new steady state keeping the integrity of the fuel elements. (author)

Analysis of loss of flow events on Brazilian multipurpose reactor by RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Soares, Humberto V.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria Auxiliadora F., E-mail: antonella@nuclear.ufmg.br, E-mail: laubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, UFMG, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores, CNPq (Brazil); Aronne, Ivan D.; Rezende, Guilherme P., E-mail: aroneid@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte (Brazil).

2011-07-01

The Brazilian Multipurpose Reactor (BMR) is currently being projected and analyzed. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly, with planar plates. BMR will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of BMR using the RELAP5 model and also three transient cases of loss of flow accident (LOFA), in the primary cooling system. A LOFA may arise through failures associated with the primary cooling system pumps or through events resulting in a decrease in the primary coolant flow with the primary cooling system pumps functioning normally. The cases presented in this paper are: primary cooling system pump shaft seizure, failure of one primary cooling system pump motor and failure of both primary cooling system pump motors. In the shaft seizure case, the flow reduction is sudden, with the blocking of the flow coast down The motor failure cases, deal with the failure of one or two pump motor due to, for example, malfunction or interruption of power and differently of the shaft seizure it can be observed the flow coast down provided by the pump inertia. It is shown that after all initiating events the reactor reaches a safe new steady state keeping the integrity of the fuel elements. (author)

Trip generation and data analysis study.

Science.gov (United States)

2015-09-01

Through the Trip Generation and Data Analysis Study, the District of Columbia Department of : Transportation (DDOT) is undertaking research to better understand multimodal urban trip generation : at mixed-use sites in the District. The study is helpi...

Trip generation characteristics of special generators

Science.gov (United States)

2010-03-01

Special generators are introduced in the sequential four-step modeling procedure to represent certain types of facilities whose trip generation characteristics are not fully captured by the standard trip generation module. They are also used in the t...

Analysis of area events as part of probabilistic safety assessment for Romanian TRIGA SSR 14 MW reactor

International Nuclear Information System (INIS)

Mladin, D.; Stefan, I.

2005-01-01

The international experience has shown that the external events could be an important contributor to plant/ reactor risk. For this reason such events have to be included in the PSA studies. In the context of PSA for nuclear facilities, external events are defined as events originating from outside the plant, but with the potential to create an initiating event at the plant. To support plant safety assessment, PSA can be used to find methods for identification of vulnerable features of the plant and to suggest modifications in order to mitigate the impact of external events or the producing of initiating events. For that purpose, probabilistic assessment of area events concerning fire and flooding risk and impact is necessary. Due to the relatively large power level amongst research reactors, the approach to safety analysis of Romanian 14 MW TRIGA benefits from an ongoing PSA project. In this context, treatment of external events should be considered. The specific tasks proposed for the complete evaluation of area event analysis are: identify the rooms important for facility safety, determine a relative area event risk index for these rooms and a relative area event impact index if the event occurs, evaluate the rooms specific area event frequency, determine the rooms contribution to reactor hazard state frequencies, analyze power supply and room dependencies of safety components (as pumps, motor operated valves). The fire risk analysis methodology is based on Berry's method [1]. This approach provides a systematic procedure to carry out a relative index of different rooms. The factors, which affect the fire probability, are: personal presence in the room, number and type of ignition sources, type and area of combustibles, fuel available in the room, fuel location, and ventilation. The flooding risk analysis is based on the amount of piping in the room. For accuracy of the information regarding piping a facility walk-about is necessary. In case of flooding risk

Field Trips as Valuable Learning Experiences in Geography Courses

Science.gov (United States)

Krakowka, Amy Richmond

2012-01-01

Field trips have been acknowledged as valuable learning experiences in geography. This article uses Kolb's (1984) experiential learning model to discuss how students learn and how field trips can help enhance learning. Using Kolb's experiential learning theory as a guide in the design of field trips helps ensure that field trips contribute to…

Fellows in the Middle: Fabulous Field Trips

Science.gov (United States)

West, Mary Lou

2008-05-01

Montclair State University's NSF GK-12 Program focuses on grades 7 and 8 in five urban public school districts in northern New Jersey. Each year four fieldtrips are taken by the students, middle school teachers, and graduate student Fellows. Many interdisciplinary hands-on lessons are written for use before, during and after each trip with this year's theme of Earth history. The Sterling Hill Mine trip evoked lessons on geology, economics, crystal structure, density, and pH. A virtual trip (webcam link) to scientists in the rainforest of Panama prompted critical thinking, categorizing layers and animals, and construction of model food webs. In the field trip to the NJ School of Conservation the students will build model aquifers, measure tree heights, and measure stream flow to compare to their Hackensack River. Finally the students will travel to MSU for a Math/Science Day with research talks, lab tours, hands-on activities, and a poster session. In January 2008 seventeen teachers, Fellows, and grant personnel took a field trip to China to set up collaborations with researchers and schools in Beijing and Xi'an, including the Beijing Ancient Observatory. All field trips are fabulous! Next year (IYA) our theme will be planetary science and will feature field trips to the Newark Museum's Dreyfuss Planetarium, BCC Buehler Challenger & Science Center, and star parties. We look forward to invigorating middle school science and mathematics with exciting astronomy. Funded by NSF #0638708

Strengthening the First Line of Defence: Delayed Turbine Trip at SCRAM in Westinghouse type NPP's

International Nuclear Information System (INIS)

Van Berlo, Marcel M.A.J.

2015-01-01

The availability of Information, Control and Power (ICP) is not treated as a Critical Safety Function (CSF). After the Forsmark (2006) and Fukushima (2011) incidents there is reason to add ICP as a separate CSF. Adding ICP as a separate CSF would possibly lead to procedural adaptations, or even design changes, for Nuclear Power Plants. As an example, this paper focusses on the transitions immediately after a SCRAM. At a SCRAM in many nuclear power plants the turbine is tripped immediately to prevent the extraction of too much heat from the reactor. However this requires a large and fast transition for the entire secondary system. The rescheduled priorities could lead to the wish NOT to trip the turbine before load has been reduced and alternative power has been secured. This paper discusses a 'soft landing' for the turbine by keeping it running after the SCRAM. Turbine control can follow reactor power by controlling the pressure of the available residual steam from the steam generator. With a proper control design this enables a flexible and precise control of primary temperatures without any fast switching in the secondary system during the first 1/2 to 3 minutes. In this period reactor load and turbine power are smoothly lowered to minimum levels during of which automatic preparatory measures can be triggered. The normal transitions can be initiated in a staged form to provide a soft landing for the entire secondary and electrical system. (author)

Online failed fuel identification using delayed neutron detector signals in pool type reactors

International Nuclear Information System (INIS)

Upadhyay, Chandra Kant; Sivaramakrishna, M.; Nagaraj, C.P.; Madhusoodanan, K.

2011-01-01

In todays world, nuclear reactors are at the forefront of modern day innovation and reactor designs are increasingly incorporating cutting edge technology. It is of utmost importance to detect failure or defects in any part of a nuclear reactor for healthy operation of reactor as well as the safety aspects of the environment. Despite careful fabrication and manufacturing of fuel pins, there is a chance of clad failure. After fuel pin clad rupture takes place, it allows fission products to enter in to sodium pool. There are some potential consequences due to this such as Total Instantaneous Blockage (TIB) of coolant and primary component contamination. At present, the failed fuel detection techniques such as cover gas monitoring (alarming the operator), delayed neutron detection (DND-automatic trip) and standalone failed fuel localization module (FFLM) are exercised in various reactors. The first technique is a quantitative measurement of increase in the cover gas activity background whereas DND system causes automatic trip on detecting certain level of activity during clad wet rupture. FFLM is subsequently used to identify the failed fuel subassembly. The later although accurate, but mainly suffers from downtime and reduction in power during identification process. The proposed scheme, reported in this paper, reduces the operation of FFLM by predicting the faulty sector and therefore reducing reactor down time and thermal shocks. The neutron evolution pattern gets modulated because fission products are the delay neutron precursors. When they travel along with coolant to Intermediate heat Exchangers, experienced three effects i.e. delay; decay and dilution which make the neutron pulse frequency vary depending on the location of failed fuel sub assembly. This paper discusses the method that is followed to study the frequency domain properties, so that it is possible to detect exact fuel subassembly failure online, before the reactor automatically trips. (author)

FIX-II/2032, BWR Pump Trip Experiment 2032, Simulation Mass Flow and Power Transients

International Nuclear Information System (INIS)

1988-01-01

1 - Description of test facility: In the FIX-II pump trip experiments, mass flow and power transients were simulated subsequent to a total loss of power to the recirculation pumps in an internal pump boiling water reactor. The aim was to determine the initial power limit to give dryout in the fuel bundle for the specified transient. In addition, the peak cladding temperature was measured and the rewetting was studied. 2 - Description of test: Pump trip experiment 2032 was a part of test group 2, i.e. the mass flow transient was to simulate the pump coast down with a pump inertia of 11.3 kg.m -2 . The initial power in the 36-rod bundle was 4.44 MW which gave dryout after 1.4 s from the start of the flow transient. A maximum rod cladding temperature of 457 degrees C was measured. Rewetting was obtained after 7.6 s. 3 - Experimental limitations or shortcomings: No ECCS injection systems

How combined trip purposes are associated with transport choice for short distance trips. Results from a cross-sectional study in the Netherlands.

Directory of Open Access Journals (Sweden)

Eline Scheepers

Full Text Available One way to increase physical activity is to stimulate a shift from car use to walking or cycling. In single-purpose trips, purpose was found to be an important predictor of transport choice. However, as far as known, no studies have been conducted to see how trips with combined purposes affect this decision. This study was designed to provide insight into associations between combined purposes and transport choice.An online questionnaire (N = 3,663 was used to collect data concerning transport choice for four primary purposes: shopping, going to public natural spaces, sports, and commuting. Per combination of primary trip purpose and transport choice, participants were asked to give examples of secondary purposes that they combine with the primary purpose. Logistic regression analyses were used to model the odds of both cycling and walking versus car use.Primary trip purposes combined with commuting, shopping, visiting private contacts or medical care were more likely to be made by car than by cycling or walking. Combinations with visiting catering facilities, trips to social infrastructure facilities, recreational outings, trips to facilities for the provision of daily requirements or private contacts during the trip were more likely to be made by walking and/or cycling than by car.Combined trip purposes were found to be associated with transport choice. When stimulating active transport focus should be on the combined-trip purposes which were more likely to be made by car, namely trips combined with commuting, other shopping, visiting private contacts or medical care.

Indian Point 2 steam generator tube rupture analyses

International Nuclear Information System (INIS)

Dayan, A.

1985-01-01

Analyses were conducted with RETRAN-02 to study consequences of steam generator tube rupture (SGTR) events. The Indian Point, Unit 2, power plant (IP2, PWR) was modeled as a two asymmetric loops, consisting of 27 volumes and 37 junctions. The break section was modeled once, conservatively, as a 150% flow area opening at the wall of the steam generator cold leg plenum, and once as a 200% double-ended tube break. Results revealed 60% overprediction of breakflow rates by the traditional conservative model. Two SGTR transients were studied, one with low-pressure reactor trip and one with an earlier reactor trip via over temperature ΔT. The former is more typical to a plant with low reactor average temperature such as IP2. Transient analyses for a single tube break event over 500 seconds indicated continued primary subcooling and no need for steam line pressure relief. In addition, SGTR transients with reactor trip while the pressurizer still contains water were found to favorably reduce depressurization rates. Comparison of the conservative results with independent LOFTRAN predictions showed good agreement

Using GIS for planning field trips: In-situ assessment of Geopoints for field trips with mobile devices

Science.gov (United States)

Böhm, Sarah; Kisser, Thomas; Ditter, Raimund

2016-04-01

Up to now no application is existing for collecting data via mobile devices using a geographical information system referring to the evaluation of Geopoints. Classified in different geographical topics a Geopark can be rated for suitability of Geopoints for field trips. The systematically acquisition of the suitability of Geopoints is necessary, especially when doing field trips with lower grade students who see a physical-geographic phenomenon for the first time. For this reason, the development of such an application is an invention for easy handling evaluations of Geopoints on the basis of commonly valid criteria like esthetic attraction, interestingness, and pithiness (Streifinger 2010). Collecting data provides the opportunity of receiving information of particularly suitable Geopoints out of the sight from students, tourists and others. One solution for collecting data in a simple and intuitive form is Survey123 for ArcGIS (http://survey123.esri.com/#/). You can create surveys using an ArcGIS Online organizational account and download your own survey or surveys "that may have been shared with you" (https://itunes.apple.com/us/app/survey-123-for-arcgis/id993015031?mt=8) on your mobile device. "Once a form is downloaded, you will be able to start collecting data."(https://itunes.apple.com/us/app/survey-123-for-arcgis/id993015031?mt=8) Free of cost and use while disconnected the application can easily be used via mobile device on field trips. On a 3-day field trip which is held three times per year in the Geopark Bergstraße-Odenwald Survey123 is being used to evaluate the suitability of different Geopoints for different topics (geology, soils, vegetation, climate). With every field trip about 25 students take part in the survey and evaluate each Geopoint at the route. So, over the time, the docents know exactly which Geopoints suites perfect for teaching geology for example, and why it suites that good. The field trip is organized in an innovative way. Before

Slip, trip and fall accidents occurring during the delivery of mail.

Science.gov (United States)

Bentley, T A; Haslam, R A

1998-12-01

This study sought to identify causal factors for slip, trip and fall accidents occurring during the delivery of mail. Analysis of in-house data produced information about accident circumstances for 1734 fall cases. The most common initiating events in delivery falls were slips and trips. Slips most often occurred on snow, ice or grass, while trips tended to involve uneven pavements, obstacles and kerbs. Nearly one-fifth of falls occurred on steps, with step falls requiring longer absence from work than falls on the level. Half of all falls occurred during November-February and three-quarters of falls occurred between 7 and 9 a.m. Incidence rates for female employees were 50% higher than for their male colleagues. Accident-independent methods included interviews with safety personnel and managers, discussion groups with delivery employees, and a questionnaire survey of employees and managers. These techniques provided data on risk factors related to the task, behaviour, footwear and equipment. Arising from these accident-independent investigations, it is suggested that unsafe working practices, such as reading addresses while walking and taking shortcuts, increase the risk of falls. Organizational issues include management safety activities, training and equipment provision. Measures are discussed that might lead to a reduction in the incidence of delivery fall accidents.

Development of the digitalized automatic seismic trip system for nuclear AR power plants using the systems engineering approach

International Nuclear Information System (INIS)

Jung, Jae Cheon

2014-01-01

The automatic seismic trip system (ASTS) continuously monitors PGA (peak ground acceleration) from the seismic wave, and automatically generates a trip signal. This work presents how the system can be designed by using a systems engineering approach under the given regulatory criteria. Overall design stages, from the needs analysis to design verification, have been executed under the defined processes and activities. Moreover, this work contributes two significant design areas for digitalized ASTS. These are firstly, how to categorize the ASTS if the ASTS has a backed up function of the manual reactor trip, and secondly, how to set the requirements using the given design practices either in overseas ASTS design or similar design. In addition, the methodology for determining the setpoint can be applied to the I and C design and development project which needs to justify the error sources correctly. The systematic approach that has been developed and realized in this work can be utilized in designing new I and C (instrument and control system) as well.

28 CFR 570.45 - Violation of escorted trip.

Science.gov (United States)

2010-07-01

... 28 Judicial Administration 2 2010-07-01 2010-07-01 false Violation of escorted trip. 570.45 Section 570.45 Judicial Administration BUREAU OF PRISONS, DEPARTMENT OF JUSTICE COMMUNITY PROGRAMS AND RELEASE COMMUNITY PROGRAMS Escorted Trips § 570.45 Violation of escorted trip. (a) Staff shall process as...

Aging assessment of surge protective devices in nuclear power plants

International Nuclear Information System (INIS)

Davis, J.F.; Subudhi, M.; Carroll, D.P.

1996-01-01

An assessment was performed to determine the effects of aging on the performance and availability of surge protective devices (SPDs), used in electrical power and control systems in nuclear power plants. Although SPDs have not been classified as safety-related, they are risk-important because they can minimize the initiating event frequencies associated with loss of offsite power and reactor trips. Conversely, their failure due to age might cause some of those initiating events, e.g., through short circuit failure modes, or by allowing deterioration of the safety-related component(s) they are protecting from overvoltages, perhaps preventing a reactor trip, from an open circuit failure mode. From the data evaluated during 1980--1994, it was found that failures of surge arresters and suppressers by short circuits were neither a significant risk nor safety concern, and there were no failures of surge suppressers preventing a reactor trip. Simulations, using the ElectroMagnetic Transients Program (EMTP) were performed to determine the adequacy of high voltage surge arresters

TREAT Reactor Control and Protection System

International Nuclear Information System (INIS)

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

Logic elements for reactor period meter

Science.gov (United States)

McDowell, William P.; Bobis, James P.

1976-01-01

Logic elements are provided for a reactor period meter trip circuit. For one element, first and second inputs are applied to first and second chopper comparators, respectively. The output of each comparator is O if the input applied to it is greater than or equal to a trip level associated with each input and each output is a square wave of frequency f if the input applied to it is less than the associated trip level. The outputs of the comparators are algebraically summed and applied to a bandpass filter tuned to f. For another element, the output of each comparator is applied to a bandpass filter which is tuned to f to give a sine wave of frequency f. The outputs of the filters are multiplied by an analog multiplier whose output is 0 if either input is 0 and a sine wave of frequency 2f if both inputs are a frequency f.

Observing Trip Chain Characteristics of Round-Trip Carsharing Users in China: A Case Study Based on GPS Data in Hangzhou City

Directory of Open Access Journals (Sweden)

Ying Hui

2017-06-01

Full Text Available Carsharing as a means to provide individuals with access to automobiles to complete a personal trip has grown significantly in recent years in China. However, there are few case studies based on operational data to show the role carsharing systems play in citizens’ daily trips. In this study, vehicle GPS data of a round-trip carsharing system in Hangzhou, China was used to describe the trip chain characteristics of users. For clearer delineation of carshare usage, the car use time length of all observations chosen in the study was within 24 h or less. Through data preprocessing, a large pool (26,085 of valid behavior samples was obtained, and several trip chaining attributes were selected to describe the characteristics. The pool of observations was then classified into five clusters, with each cluster having significant differences in one or two trip chain characteristics. The cluster results reflected that different use patterns exist. By a comparative analysis with trip survey data in Hangzhou, differences in trip chain characteristics exist between carsharing and private cars, but in some cases, shared vehicles can be a substitute for private cars to satisfy motorized travel. The proposed method could facilitate companies in formulating a flexible pricing strategy and determining target customers. In addition, traffic administration agencies could have a deeper understanding of the position and function of various carsharing modes in an urban transportation system.

Operational safety evaluation for minor reactor accidents

International Nuclear Information System (INIS)

Wang, O.S.

1981-01-01

The purpose of this paper is to address a concern of applying conservatism in analysing minor reactor incidents. A so-called ''conservative'' safety analysis may exaggerate the system responses and result in a reactor scram tripped by the reactor protective system (RPS). In reality, a minor incident may lead the reactor to a new thermal hydraulic steady-state without scram, and the mitigation or termination of the incident may entirely depend on operator actions. An example on a small steamline break evaluation for a pressurized water reactor recently investigated by the staff at the Washington Public Power Supply System is presented to illustrate this point. A safety evaluation using mainly the safety-related systems to be consistent with the conservative assumptions used in the Safety Analysis Report was conducted. For comparison, a realistic analysis was also performed using both the safety- and control-related systems. The analyses were performed using the RETRAN plant simulation computer code. The ''conservative'' safety analysis predicts that the incident can be turned over by the RPS scram trips without operator intervention. However, the realistic analysis concludes that the reactor will reach a new steady-state at a different plant thermal hydraulic condition. As a result, the termination of the incident at this stage depends entirely on proper operator action. On the basis of this investigation it is concluded that, for minor incidents, ''conservative'' assumptions are not necessary, sometimes not justifiable. A realistic investigation from the operational safety point of view is more appropriate. It is essential to highlight the key transient indications for specific incident recognition in the operator training program

Evaluation of Steam Generator Level behavior for Determination of Turbine Runback rate on COPs trip for Yonggwang 1 and 2 Power Uprating Units

International Nuclear Information System (INIS)

Lee, Kyung Jin; Hwang, Su Hyun; Yoo, Tae Geun; Chung, Soon Il; An, Byung Chang; Park, Jung Gu

2010-01-01

4.5% power uprate project has been progressing for the first time in Yonggwang 1 and 2(YGN1 and 2). Reviews for design change due to the power uprate were accomplished. Steam generator level behavior was one of the most important parameters because it could be cause of reactor trip or turbine trip. As the results of the reviews, YGN1 and 2 had to reassess it for change of turbine runback rate when turbine runback occurs due to the condensate operating pumps (COP) trip. This study has been carried out for evaluating the steam generator level behavior for determination of turbine runback rate on COPs trip for Yonggwang 1 and 2 Power Uprating Units. The steam generator water level evaluation program for YGN1 and 2 (SLEP-Y1) has been developed for it. The program includes models for the steam generator water level response. SLEP-Y1 is programmed with advanced continuous system simulation language (ACSL). The language has been used to simulate physical systems as a commercial tool used to evaluate system designs

Analysis of core damage frequency due to external events at the DOE [Department of Energy] N-Reactor

International Nuclear Information System (INIS)

Lambright, J.A.; Bohn, M.P.; Daniel, S.L.; Baxter, J.T.; Johnson, J.J.; Ravindra, M.K.; Hashimoto, P.O.; Mraz, M.J.; Tong, W.H.; Conoscente, J.P.; Brosseau, D.A.

1990-11-01

A complete external events probabilistic risk assessment has been performed for the N-Reactor power plant, making full use of all insights gained during the past ten years' developments in risk assessment methodologies. A detailed screening analysis was performed which showed that all external events had negligible contribution to core damage frequency except fires, seismic events, and external flooding. A limited scope analysis of the external flooding risk indicated that it is not a major risk contributor. Detailed analyses of the fire and seismic risks resulted in total (mean) core damage frequencies of 1.96E-5 and 4.60E-05 per reactor year, respectively. Detailed uncertainty analyses were performed for both fire and seismic risks. These results show that the core damage frequency profile for these events is comparable to that found for existing commercial power plants if proposed fixes are completed as part of the restart program. 108 refs., 85 figs., 80 tabs

Analysis of core damage frequency due to external events at the DOE (Department of Energy) N-Reactor

Energy Technology Data Exchange (ETDEWEB)

Lambright, J.A.; Bohn, M.P.; Daniel, S.L. (Sandia National Labs., Albuquerque, NM (USA)); Baxter, J.T. (Westinghouse Hanford Co., Richland, WA (USA)); Johnson, J.J.; Ravindra, M.K.; Hashimoto, P.O.; Mraz, M.J.; Tong, W.H.; Conoscente, J.P. (EQE, Inc., San Francisco, CA (USA)); Brosseau, D.A. (ERCE, Inc., Albuquerque, NM (USA))

1990-11-01

A complete external events probabilistic risk assessment has been performed for the N-Reactor power plant, making full use of all insights gained during the past ten years' developments in risk assessment methodologies. A detailed screening analysis was performed which showed that all external events had negligible contribution to core damage frequency except fires, seismic events, and external flooding. A limited scope analysis of the external flooding risk indicated that it is not a major risk contributor. Detailed analyses of the fire and seismic risks resulted in total (mean) core damage frequencies of 1.96E-5 and 4.60E-05 per reactor year, respectively. Detailed uncertainty analyses were performed for both fire and seismic risks. These results show that the core damage frequency profile for these events is comparable to that found for existing commercial power plants if proposed fixes are completed as part of the restart program. 108 refs., 85 figs., 80 tabs.

Field Trip - Conservation of Carnivores in Namibia

Science.gov (United States)

Gibson, Amanda

2017-04-01

Field trips are a key component of our curriculum at ISWB. Classroom teaching is invaluable but field trips provide pupils with a tangible connection to pertinent issues of conservation. ISWB realises the importance of out of the classroom learning in field trips and to this end our students have an opportunity to partake in a number of 3-5 day field trips per academic year. In 2016, several Year 8, 9, 10, 11 and 12 students visited the AfriCat Foundation on Okonjima in central Namibia for 4 days to learn about the conservation of the predator population in Namibia. The trips were very successful and another trip this year to AfriCat North close to Etosha National Park, where the students will work closely with the local farming communities, is planned. AfriCat provides Environmental Education programmes for the youth of Namibia giving them a greater understanding of the importance of wildlife conservation. Their main objective is promoting predator and environmental awareness amongst the youth of Namibia. AfriCat Environmental Education Programme is based on 1997 UNESCO-UNEP Environmental Education objectives. "Attitudes: To raise concern about problems, values, personal responsibility and willingness to participate/act. In the end, we conserve only what we love. We will love only what we understand. We will understand only what we are taught."

Containment Performance Evaluation of a Sodium Fire Event Due to Air Ingress into the Cover Gas Region of the Reactor Vessel in the PGSFR

International Nuclear Information System (INIS)

Ahn, Sang June; Chang, Won-Pyo; Kang, Seok Hun; Choi, Chi-Woong; Yoo, Jin; Lee, Kwi Lim; Jeong, Jae-Ho; Lee, Seung Won; Jeong, Taekyeong; Ha, Kwi-Seok

2015-01-01

Comparing with the light water reactor, sodium as a reactor coolant violently reacts with oxygen in the containment atmosphere. Due to this chemical reaction, heat generated from the combustion heat increases the temperature and pressure in the containment atmosphere. The structural integrity of the containment building which is a final radiological defense barrier is threaten. A sodium fire event in the containment due to air ingress into the cover gas region in the reactor vessel is classified as one of the design basis events in the PGSFR. This event comes from a leak or crack on the reactor upper closure header surface. It accompanys an event of the radiological fission products release to the inside the containment. In this paper, evaluation for the sodium fire and radiological influence due to air ingress into the cover gas region of the reactor vessel is described. To evaluate this event, the CONTAIN-LMR, MACCS-II and OR-IGEN-II codes are used. For the sodium pool fire event in the containment, the performance evaluation and radiological influence are carried out. In the thermal hydraulic aspects, the 1 cell containment yields the most conservative result. In this event, the maximum temperature and pressure in the containment are calculated 0.185 MPa, 280.0 .deg. C, respectively. The radiological dose at the EAB and LPZ are below the acceptance criteria specified in the 10CFR100

Containment Performance Evaluation of a Sodium Fire Event Due to Air Ingress into the Cover Gas Region of the Reactor Vessel in the PGSFR

Energy Technology Data Exchange (ETDEWEB)

Ahn, Sang June; Chang, Won-Pyo; Kang, Seok Hun; Choi, Chi-Woong; Yoo, Jin; Lee, Kwi Lim; Jeong, Jae-Ho; Lee, Seung Won; Jeong, Taekyeong; Ha, Kwi-Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Comparing with the light water reactor, sodium as a reactor coolant violently reacts with oxygen in the containment atmosphere. Due to this chemical reaction, heat generated from the combustion heat increases the temperature and pressure in the containment atmosphere. The structural integrity of the containment building which is a final radiological defense barrier is threaten. A sodium fire event in the containment due to air ingress into the cover gas region in the reactor vessel is classified as one of the design basis events in the PGSFR. This event comes from a leak or crack on the reactor upper closure header surface. It accompanys an event of the radiological fission products release to the inside the containment. In this paper, evaluation for the sodium fire and radiological influence due to air ingress into the cover gas region of the reactor vessel is described. To evaluate this event, the CONTAIN-LMR, MACCS-II and OR-IGEN-II codes are used. For the sodium pool fire event in the containment, the performance evaluation and radiological influence are carried out. In the thermal hydraulic aspects, the 1 cell containment yields the most conservative result. In this event, the maximum temperature and pressure in the containment are calculated 0.185 MPa, 280.0 .deg. C, respectively. The radiological dose at the EAB and LPZ are below the acceptance criteria specified in the 10CFR100.

Requirements of a proton beam accelerator for an accelerator-driven reactor

International Nuclear Information System (INIS)

Takahashi, H.; Zhao, Y.; Tsoupas, N.; An, Y.; Yamazaki, Y.

1997-01-01

When the authors first proposed an accelerator-driven reactor, the concept was opposed by physicists who had earlier used the accelerator for their physics experiments. This opposition arose because they had nuisance experiences in that the accelerator was not reliable, and very often disrupted their work as the accelerator shut down due to electric tripping. This paper discusses the requirements for the proton beam accelerator. It addresses how to solve the tripping problem and how to shape the proton beam

Effect of fuel string relocation on the consequences of postulated inlet header LBLOCA in KANUPP reactor

International Nuclear Information System (INIS)

Ahmed, I.; Chow, H.C.; Younis, M.H.

1996-01-01

An investigation aimed at determining the effect of fuel string relocation on reactivity excursion and power pulse following a hypothetical Large Break Loss of Coolant Accident in KANUPP reactor is reported. The assessment of reactivity insertion was performed making use of global (reactor) core analysis computer code RFSP. The reactor kinetics module CERBERUS of the RFSP code and the SOPHT (thermal-hydraulics code) were subsequently employed for the neutronic transient analysis. The effect was evaluated in context of determining the adequacy of moderator dump shutdown system. Because of the presence of the gap between the inlet shield plug and the fuel string, the fuel bundles may shift in such a manner that low-irradiated fuel is moved towards the core centre. This represents an additional reactivity increase to be accounted for in the analysis. The reactivity excursion, however, is alleviated by an earlier reactor trip. The net impact is that the energy deposited in the maximum rated fuel pencil is increased from 56% of the 960 kJ/kg fuel-centre-line melting limit to 63%. The result demonstrated the adequacy of the shutdown system against the maximum credible accident event. (author)

Evidence, explanations, and recommendations for teachers' field trip strategies

Science.gov (United States)

Rebar, Bryan

Field trips are well recognized by researchers as an educational approach with the potential to complement and enhance classroom science teaching by exposing students to unique activities, resources, and content in informal settings. The following investigation addresses teachers' field trip practices in three related manuscripts: (1) A study examining the details of teachers' pedagogical strategies intended to facilitate connections between students' experiences and the school curricula while visiting an aquarium; (2) A study documenting and describing sources of knowledge that teachers draw from when leading field trips to an aquarium; (3) A position paper that reviews and summarizes research on effective pedagogical strategies for field trips. Together these three pieces address key questions regarding teachers' practices on field trips: (1) What strategies are teachers employing (and not employing) during self-guided field trips to facilitate learning tied to the class curriculum? (2) What sources of knowledge do teachers utilize when leading field trips? (3) How can teachers be better prepared to lead trips that promote learning? The Oregon Coast Aquarium served as the field trip site for teachers included in this study. The setting suited these questions because the aquarium serves tens of thousands of students on field trips each year but provides no targeted programming for these students as they explore the exhibits. In other words, the teachers who lead field trips assume much of the responsibility for facilitating students' experience. In order to describe and characterize teachers' strategies to link students' experiences to the curriculum, a number of teachers (26) were observed as they led their students' visit to the public spaces of the aquarium. Artifacts, such as worksheets, used during the visit were collected for analysis as well. Subsequently, all teachers were surveyed regarding their use of the field trip and their sources of knowledge for

Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix I. Accident definition and use of event trees

International Nuclear Information System (INIS)

1975-10-01

Information is presented concerning accident definition and use of event trees, event tree methodology, potential accidents covered by the reactor safety study, analysis of potential accidents involving the reactor core, and analysis of potential accidents not involving the core

NRC Fact-Finding Task Force report on the ATWS event at Salem Nuclear Generating Station, Unit 1, on February 25, 1983

International Nuclear Information System (INIS)

1983-03-01

An NRC Region I Task Force was established on March 1, 1983 to conduct fact finding and data collection with regard to the circumstances which led to an anticipated transient without scram (ATWS) event at the Public Service Electric and Gas Company's Salem Generating Station, Unit 1 on February 25, 1983. The charter of the Task Force was to determine the factual information pertinent to management and administrative controls which should have ensured proper operation of the reactor trip breakers in the solid state protection system. This report documents the findings of the Task Force along with its conclusions

Elementary school children's science learning from school field trips

Science.gov (United States)

Glick, Marilyn Petty

This research examines the impact of classroom anchoring activities on elementary school students' science learning from a school field trip. Although there is prior research demonstrating that students can learn science from school field trips, most of this research is descriptive in nature and does not examine the conditions that enhance or facilitate such learning. The current study draws upon research in psychology and education to create an intervention that is designed to enhance what students learn from school science field trips. The intervention comprises of a set of "anchoring" activities that include: (1) Orientation to context, (2) Discussion to activate prior knowledge and generate questions, (3) Use of field notebooks during the field trip to record observations and answer questions generated prior to field trip, (4) Post-visit discussion of what was learned. The effects of the intervention are examined by comparing two groups of students: an intervention group which receives anchoring classroom activities related to their field trip and an equivalent control group which visits the same field trip site for the same duration but does not receive any anchoring classroom activities. Learning of target concepts in both groups was compared using objective pre and posttests. Additionally, a subset of students in each group were interviewed to obtain more detailed descriptive data on what children learned through their field trip.

Boiling water reactor turbine trip (TT) benchmark. Volume II: Summary Results of Exercise 1

International Nuclear Information System (INIS)

Akdeniz, Bedirhan; Ivanov, Kostadin N.; Olson, Andy M.

2005-06-01

The OECD Nuclear Energy Agency (NEA) completed under US Nuclear Regulatory Commission (NRC) sponsorship a PWR main steam line break (MSLB) benchmark against coupled system three-dimensional (3-D) neutron kinetics and thermal-hydraulic codes. Another OECD/NRC coupled-code benchmark was recently completed for a BWR turbine trip (TT) transient and is the object of the present report. Turbine trip transients in a BWR are pressurisation events in which the coupling between core space-dependent neutronic phenomena and system dynamics plays an important role. The data made available from actual experiments carried out at the Peach Bottom 2 plant make the present benchmark particularly valuable. While defining and coordinating the BWR TT benchmark, a systematic approach and level methodology not only allowed for a consistent and comprehensive validation process, but also contributed to the study of key parameters of pressurisation transients. The benchmark consists of three separate exercises, two initial states and five transient scenarios. The BWR TT Benchmark will be published in four volumes as NEA reports. CD-ROMs will also be prepared and will include the four reports and the transient boundary conditions, decay heat values as a function of time, cross-section libraries and supplementary tables and graphs not published in the paper version. BWR TT Benchmark - Volume I: Final Specifications was issued in 2001 [NEA/NSC/DOC(2001)]. The benchmark team [Pennsylvania State University (PSU) in co-operation with Exelon Nuclear and the NEA] has been responsible for coordinating benchmark activities, answering participant questions and assisting them in developing their models, as well as analysing submitted solutions and providing reports summarising the results for each phase. The benchmark team has also been involved in the technical aspects of the benchmark, including sensitivity studies for the different exercises. Volume II summarises the results for Exercise 1 of the

Development of a Seismic Setpoint Calculation Methodology Using a Safety System Approach

International Nuclear Information System (INIS)

Lee, Chang Jae; Baik, Kwang Il; Lee, Sang Jeong

2013-01-01

The Automatic Seismic Trip System (ASTS) automatically actuates reactor trip when it detects seismic activities whose magnitudes are comparable to a Safe Shutdown Earthquake (SSE), which is the maximum hypothetical earthquake at the nuclear power plant site. To ensure that the reactor is tripped before the magnitude of earthquake exceeds the SSE, it is crucial to reasonably determine the seismic setpoint. The trip setpoint and allowable value for the ASTS for Advanced Power Reactor (APR) 1400 Nuclear Power Plants (NPPs) were determined by the methodology presented in this paper. The ASTS that trips the reactor when a large earthquake occurs is categorized as a non safety system because the system is not required by design basis event criteria. This means ASTS has neither specific analytical limit nor dedicated setpoint calculation methodology. Therefore, we developed the ASTS setpoint calculation methodology by conservatively considering that of PPS. By incorporating the developed methodology into the ASTS for APR1400, the more conservative trip setpoint and allowable value were determined. In addition, the ZPA from the Operating Basis Earthquake (OBE) FRS of the floor where the sensor module is located is 0.1g. Thus, the allowance of 0.17g between OBE of 0.1 g and ASTS trip setpoint of 0.27 g is sufficient to prevent the reactor trip before the magnitude of the earthquake exceeds the OBE. In result, the developed ASTS setpoint calculation methodology is evaluated as reasonable in both aspects of the safety and performance of the NPPs. This will be used to determine the ASTS trip setpoint and allowable for newly constructed plants

Advance Liquid Metal Reactor Discrete Dynamic Event Tree/Bayesian Network Analysis and Incident Management Guidelines (Risk Management for Sodium Fast Reactors)

Energy Technology Data Exchange (ETDEWEB)

Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-04-01

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self-correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the system's design to manage the accident. Inherently and passively safe designs are laudable, but nonetheless extreme boundary conditions can interfere with the design attributes which facilitate inherent safety, thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayesian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The authors would like to acknowledge the U.S. Department of Energy's Office of Nuclear Energy for funding this research through Work Package SR-14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at Argonne National Laboratory, Oak Ridge National Laboratory, and Idaho National Laboratory for their continue d contributions to the advanced reactor PRA mission area.

Microstructure characterization of Friction Stir Spot Welded TRIP steel

DEFF Research Database (Denmark)

Lomholt, Trine Colding; Adachi, Yoshitaka; Peterson, Jeremy

2012-01-01

Transformation Induced Plasticity (TRIP) steels have not yet been successfully joined by any welding technique. It is desirable to search for a suitable welding technique that opens up for full usability of TRIP steels. In this study, the potential of joining TRIP steel with Friction Stir Spot...

Austenite stability in TRIP steels studied by synchrotron radiation

NARCIS (Netherlands)

Blondé, R.

2014-01-01

TRIP steel is a material providing great mechanical properties. Such steels show a good balance between high-strength and ductility, not only as a result of the fine microstructure, but also because of the well-known TRIP effect. The Transformation Induced-Plasticity (TRIP) phenomenon is the

Compatibility analysis of DUPIC fuel (Part II) - Reactor physics design and analysis

Energy Technology Data Exchange (ETDEWEB)

Jeong, Chang Joon; Choi, Hang Bok; Rhee, Bo Wook; Roh, Gyu Hong; Kim, Do Hun [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-03-01

The compatibility analysis of the DUPIC fuel in a CANDU reactor has been assessed. This study includes the fuel composition adjustment, comparison of lattice properties, performance analysis of reactivity devices, determination of regional over-power (ROP) trip setpoint, and uncertainty estimation of core performance parameters. For the DUPIC fuel composition adjustment, three options have been proposed, which can produce uniform neutronic characteristics of the DUPIC fuel. The lattice analysis has shown that the characteristics of the DUPIC fuel is compatible with those of natural uranium fuel. The reactivity devices of the CANDU-6 reactor maintain their functional requirements even for the DUPIC fuel system. The ROP analysis has shown that the trip setpoint is not sacrificed for the DUPIC fuel system owing to the power shape that enhances more thermal margin. The uncertainty analysis of the core performance parameter has shown that the uncertainty associated with the fuel composition variation is reduced appreciably, which is primarily due to the fuel composition adjustment and secondly the on-power refueling feature and spatial control function of the CANDU reactor. The reactor physics calculation has also shown that it is feasible to use spent PWR fuel directly in CANDU reactors without deteriorating the CANDU-6 core physics design requirements. 29 refs., 67 figs., 60 tabs. (Author)

Multi-Destination and Multi-Purpose Trip Effects in the Analysis of the Demand for Trips to a Remote Recreational Site

Science.gov (United States)

Martínez-Espiñeira, Roberto; Amoako-Tuffour, Joe

2009-06-01

One of the basic assumptions of the travel cost method for recreational demand analysis is that the travel cost is always incurred for a single purpose recreational trip. Several studies have skirted around the issue with simplifying assumptions and dropping observations considered as nonconventional holiday-makers or as nontraditional visitors from the sample. The effect of such simplifications on the benefit estimates remains conjectural. Given the remoteness of notable recreational parks, multi-destination or multi-purpose trips are not uncommon. This article examines the consequences of allocating travel costs to a recreational site when some trips were taken for purposes other than recreation and/or included visits to other recreational sites. Using a multi-purpose weighting approach on data from Gros Morne National Park, Canada, we conclude that a proper correction for multi-destination or multi-purpose trip is more of what is needed to avoid potential biases in the estimated effects of the price (travel-cost) variable and of the income variable in the trip generation equation.

Field-trip guide to the geologic highlights of Newberry Volcano, Oregon

Science.gov (United States)

Jensen, Robert A.; Donnelly-Nolan, Julie M.

2017-08-09

Newberry Volcano and its surrounding lavas cover about 3,000 square kilometers (km2) in central Oregon. This massive, shield-shaped, composite volcano is located in the rear of the Cascades Volcanic Arc, ~60 km east of the Cascade Range crest. The volcano overlaps the northwestern corner of the Basin and Range tectonic province, known locally as the High Lava Plains, and is strongly influenced by the east-west extensional environment. Lava compositions range from basalt to rhyolite. Eruptions began about half a million years ago and built a broad composite edifice that has generated more than one caldera collapse event. At the center of the volcano is the 6- by 8-km caldera, created ~75,000 years ago when a major explosive eruption of compositionally zoned tephra led to caldera collapse, leaving the massive shield shape visible today. The volcano hosts Newberry National Volcanic Monument, which encompasses the caldera and much of the northwest rift zone where mafic eruptions occurred about 7,000 years ago. These young lava flows erupted after the volcano was mantled by the informally named Mazama ash, a blanket of volcanic ash generated by the eruption that created Crater Lake about 7,700 years ago. This field trip guide takes the visitor to a variety of easily accessible geologic sites in Newberry National Volcanic Monument, including the youngest and most spectacular lava flows. The selected sites offer an overview of the geologic story of Newberry Volcano and feature a broad range of lava compositions. Newberry’s most recent eruption took place about 1,300 years ago in the center of the caldera and produced tephra and lava of rhyolitic composition. A significant mafic eruptive event occurred about 7,000 years ago along the northwest rift zone. This event produced lavas ranging in composition from basalt to andesite, which erupted over a distance of 35 km from south of the caldera to Lava Butte where erupted lava flowed west to temporarily block the Deschutes

Some post operational adjustments to the prototype fast reactor at Dounreay

International Nuclear Information System (INIS)

Lunt, A.R.W.

1979-01-01

Prior to and during the initial operation of the Prototype Fast Reactor at Dounreay certain features have been considered to be in need of adjustment to provide better operating characteristics. This article describes the work done to support the consequential changes of operational techniques and plant design in the following areas: maintenance of dry conditions at the superheater steam inlets, the temperature control of the reactor roof, and the introduction of a system enabling the reactor to continue running after a turbine trip. (author)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Development of Integrated PSA Database and Application Technology

Energy Technology Data Exchange (ETDEWEB)

Han, Sang Hoon; Park, Jin Hee; Kim, Seung Hwan; Choi, Sun Yeong; Jung, Woo Sik; Jeong, Kwang Sub; Ha Jae Joo; Yang, Joon Eon; Min Kyung Ran; Kim, Tae Woon

2005-04-15

The purpose of this project is to develop 1) the reliability database framework, 2) the methodology for the reactor trip and abnormal event analysis, and 3) the prototype PSA information DB system. We already have a part of the reactor trip and component reliability data. In this study, we extend the collection of data up to 2002. We construct the pilot reliability database for common cause failure and piping failure data. A reactor trip or a component failure may have an impact on the safety of a nuclear power plant. We perform the precursor analysis for such events that occurred in the KSNP, and to develop a procedure for the precursor analysis. A risk monitor provides a mean to trace the changes in the risk following the changes in the plant configurations. We develop a methodology incorporating the model of secondary system related to the reactor trip into the risk monitor model. We develop a prototype PSA information system for the UCN 3 and 4 PSA models where information for the PSA is inputted into the system such as PSA reports, analysis reports, thermal-hydraulic analysis results, system notebooks, and so on. We develop a unique coherent BDD method to quantify a fault tree and the fastest fault tree quantification engine FTREX. We develop quantification software for a full PSA model and a one top model.

Analysis of Loss-of-Offsite-Power Events 1997-2015

Energy Technology Data Exchange (ETDEWEB)

Johnson, Nancy Ellen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-07-01

Loss of offsite power (LOOP) can have a major negative impact on a power plant’s ability to achieve and maintain safe shutdown conditions. LOOP event frequencies and times required for subsequent restoration of offsite power are important inputs to plant probabilistic risk assessments. This report presents a statistical and engineering analysis of LOOP frequencies and durations at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience during calendar years 1997 through 2015. LOOP events during critical operation that do not result in a reactor trip, are not included. Frequencies and durations were determined for four event categories: plant-centered, switchyard-centered, grid-related, and weather-related. Emergency diesel generator reliability is also considered (failure to start, failure to load and run, and failure to run more than 1 hour). There is an adverse trend in LOOP durations. The previously reported adverse trend in LOOP frequency was not statistically significant for 2006-2015. Grid-related LOOPs happen predominantly in the summer. Switchyard-centered LOOPs happen predominantly in winter and spring. Plant-centered and weather-related LOOPs do not show statistically significant seasonality. The engineering analysis of LOOP data shows that human errors have been much less frequent since 1997 than in the 1986 -1996 time period.

Model Based Cyber Security Analysis for Research Reactor Protection System

Energy Technology Data Exchange (ETDEWEB)

Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of)

2013-07-01

The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN.

Model Based Cyber Security Analysis for Research Reactor Protection System

International Nuclear Information System (INIS)

Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung; Son, Hanseong

2013-01-01

The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN

POLCA-T simulation of OECD/NRC BWR turbine trip benchmark exercise 3 best estimate scenario TT2 test and four extreme scenarios

International Nuclear Information System (INIS)

Panayotov, D.

2004-01-01

Westinghouse transient code POLCA-T brings together the system thermal-hydraulics plant models and the 3D neutron kinetics core model. Code validation plan includes the calculations of Peach Bottom end of cycle 2 turbine trip transients and low-flow stability tests. The paper describes the objectives, method, and results of analyses performed in the final phase of OECD/NRC Peach Bottom 2 Boiling Water Reactor Turbine Trip Benchmark. Brief overview of the code features, the method of simulation, the developed 3D core model and system input deck for Peach Bottom 2 are given. The paper presents the results of benchmark exercise 3 best estimate scenario: coupled 3D core neutron kinetics with system thermal-hydraulics analyses. Performed sensitivity studies cover the SCRAM initiation, carry-under, and decay power. Obtained results including total power, steam dome, core exit, lower and upper plenum, main steam line and turbine inlet pressures showed good agreement with measured plant data Thus the POLCA-T code capabilities for correct simulation of turbine trip transients were proved The performed calculations and obtained results for extreme cases demonstrate the POLCA-T code wide range capabilities to simulate transients when scram, steam bypass, and safety and relief valves are not activated. The code is able to handle such transients even when the reactor power and pressure reach values higher than 600 % of rated power, and 10.8 MPa. (authors)

A model for TRIP steel constitutive behaviour

NARCIS (Netherlands)

Perdahcioglu, Emin Semih; Geijselaers, Hubertus J.M.; Menari, G

2011-01-01

A constitutive model is developed for TRIP steel. This is a steel which contains three or four different phases in its microstructure. One of the phases in TRIP steels is metastable austenite (Retained Austenite) which transforms to martensite upon deformation. The accompanying transformation strain

Ageing study of Cirus reactor vessel expansion bellow

International Nuclear Information System (INIS)

Ramana, W.V.; Dutta, B.K.; Kushwaha, H.S.; Sahu, A.K.; Bhatnagar, A.; Pant, R.C.

1994-01-01

Expansion bellow of Cirus reactor vessel is a comparatively weak component which is joined to top tube sheet and shell by helium tight lap weld. This has been subjected to thermal stress caused by high temperature during reactor operation and thermal shock due to trip or shutdown. Therefore a finite element analysis was carried out to assess thermal stresses and fatigue life of the component. It was found that the fluctuating stress in the bellow is far less than its endurance limit. (author). 2 tabs., 3 figs

Transforming an Exposure trip to Botanical Expedition: Introducing Ecological Research thru Exposure Trip in an Eco-tourism Site

Directory of Open Access Journals (Sweden)

Bernardo C. Lunar

2014-10-01

Full Text Available – Fieldtrips can be considered as one of the three avenues through which science can be taught - through formal classroom teaching, practical work and field trips. An exposure trip at Bangkong Kahoy Valley Field Study Center was arranged for a class of BS Biology and BS Education students enrolled in Ecology Course. This approach purposefully transformed the usual exposure trip from being a casual site visit into a focused and productive learning experience. This transformation from exposure trip to a botanical expedition has exceeded the initial activity goals. Rather than a day off from learning, the time spent at the study center has been a meaningful opportunity to engage students in an active ecological research project while delivering valuable science content. Employing the descriptive survey design, the learning gains of the students were assessed and students were directed to do a guided reflection writing using the ORID Model of Focused Conversation. The learning gains and reflections of the students confirmed that students can collaboratively develop focused research questions, make meaning from a variety of sources, carry out a vegetation analysis and conduct surveys on socio-economic status, plant resource utilization and ecotourism assessment of the host community. As students prepared for their trip and synthesized their learning afterward, they were able to come up with very impressive and scientifically sound research outputs.

Trip electrical circuit of the gyrotion

International Nuclear Information System (INIS)

Rossi, J.O.

1987-09-01

The electron cyclotron resonance heating system of INPE/LAP is shown and the trip electrical circuit of the gyrotron is described, together with its fundamental aspects. The trip electrical circuit consists basically of a series regulator circuit which regulates the output voltage level and controls the pulse width time. Besides that, a protection circuit for both tubes, regulator and gyrotron, against faults in the system. (author) [pt

Complementary, substitution, and independence among tourist trips

NARCIS (Netherlands)

Middelkoop, van M.; Borgers, A.W.J.; Timmermans, H.J.P.

1999-01-01

The relationship between day trips, short breaks (2-4 days), and holidays (5+ days) has never been examined at the level of the individual consumer because surveys on day and overnight trips are typically conducted independently. In this article, both the stated and the inferred relationship between

Lesson Plan Prototype for International Space Station's Interactive Video Education Events

Science.gov (United States)

Zigon, Thomas

1999-01-01

The outreach and education components of the International Space Station Program are creating a number of materials, programs, and activities that educate and inform various groups as to the implementation and purposes of the International Space Station. One of the strategies for disseminating this information to K-12 students involves an electronic class room using state of the art video conferencing technology. K-12 classrooms are able to visit the JSC, via an electronic field trip. Students interact with outreach personnel as they are taken on a tour of ISS mockups. Currently these events can be generally characterized as: Being limited to a one shot events, providing only one opportunity for students to view the ISS mockups; Using a "one to many" mode of communications; Using a transmissive, lecture based method of presenting information; Having student interactions limited to Q&A during the live event; Making limited use of media; and Lacking any formal, performance based, demonstration of learning on the part of students. My project involved developing interactive lessons for K-12 students (specifically 7th grade) that will reflect a 2nd generation design for electronic field trips. The goal of this design will be to create electronic field trips that will: Conform to national education standards; More fully utilize existing information resources; Integrate media into field trip presentations; Make support media accessible to both presenters and students; Challenge students to actively participate in field trip related activities; and Provide students with opportunities to demonstrate learning

Dynamic analysis of the condensate feedwater system in boiling water reactor plants

International Nuclear Information System (INIS)

Tanji, J.; Omori, T.

1982-01-01

The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents

Propagation of the trip behavior in the VENUS vertex chamber

International Nuclear Information System (INIS)

Ohama, Taro; Yamada, Yoshikazu.

1995-03-01

The high voltage system of the VENUS vertex chamber occasionally trips by a discharge somewhere among cathode electrodes during data taking. This trip behavior induces often additional trips at other electrodes such as the skin and the grid electrodes in the vertex chamber. This propagation mechanism of trips is so complicated in this system related with multi-electrodes. Although the vertex chamber is already installed inside the VENUS detector and consequently the discharge is not able to observe directly, a trial to estimate the propagation has been done using only the information which appears around the trip circuits and the power supply of the vertex chamber. (author)

Advances in global development and deployment of small modular reactors and incorporating lessons learned from the Fukushima Daiichi accident into the designs of engineered safety features of advanced reactors

International Nuclear Information System (INIS)

Hadid Subki, M.; )

2014-01-01

The IAEA has been facilitating the Member States in incorporating the lessons-learned from the Fukushima Dai-ichi Accident into the designs of engineered safety features of advanced reactors, including small modular reactors. An extended assessment is required to address challenges for advancing reactor safety in the new evolving generation of SMR plants to preserve the historic lessons in safety, through: assuring the diversity in emergency core cooling systems following loss of onsite AC power; ensuring diversity in reactor depressurization following a transient or accident; confirming independence in reactor trip and safety systems for sensors, power supplies and actuation systems, and finally diversity in maintaining containment integrity following a severe accident

TRACE/PARCS modelling of rips trip transients for Lungmen ABWR

Energy Technology Data Exchange (ETDEWEB)

Chang, C. Y. [Inst. of Nuclear Engineering and Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Lin, H. T.; Wang, J. R. [Inst. of Nuclear Energy Research, No. 1000, Wenhua Rd., Longtan Township, Taoyuan County 32546, Taiwan (China); Shih, C. [Inst. of Nuclear Engineering and Science, Dept. of Engineering and System Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China)

2012-07-01

The objectives of this study are to examine the performances of the steady-state results calculated by the Lungmen TRACE/PARCS model compared to SIMULATE-3 code, as well as to use the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen TRACE/PARCS model. In this study, three power generation methods in TRACE were utilized to analyze the three reactor internal pumps (RIPs) trip transient for the purpose of validating the TRACE/PARCS model. In general, the comparisons show that the transient responses of key system parameters agree well with the FSAR results, including core power, core inlet flow, reactivity, etc. Further studies will be performed in the future using Lungmen TRACE/PARCS model. After the commercial operation of Lungmen nuclear power plant, TRACE/PARCS model will be verified. (authors)

Study on Operator Actions during the Occurrences of Undesirable Events in PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Tom, P.P.; Nurul Husna Zainal Abidin; Lanyau, T.A.; Zaredah Hashim

2016-01-01

Due to the recent Fukushima accident, the potential risks at one and only nuclear research reactor in the country, which is the PUSPATI TRIGA Reactor (RTP), has increasingly gain concerns and an attempt on the development of Level 1 Probabilistic Safety Assessment (PSA) for this reactor has been commenced. The preliminary scope of the PSA is to analyse the risk of core degradation during normal daily operation due to the random component failure and human error. SPAR-H and THERP method is used for quantifying human error probability (HEP). However, the scopes of this study only cover the qualitative parts that use interview/questionnaire method. The objectives of the questionnaire are to identify the main action for RTP operators when any undesired incident occurs during full power operation that might be caused by random component failures. From the questionnaires that have been conducted, the respondents consisted of 4 licensed operators and 9 trainee operators. All licensed operators have experience of operating reactor for more than 15 years while the trainee operator have been operate the reactor with experience of less than 10 years. Generally, in the event of an abnormal condition involving the reactor, an operator whether a licensed operator or the trainee does not have to ask permission in advance from the top individuals to carry out scram. This is to prevent the situation becoming increasingly severe if the reactor is still operating. With complete training and knowledge derived from the management, an operator can act efficiently in any emergency case. (author)

Experimentally Evoking Nonbelieved Memories for Childhood Events

Science.gov (United States)

Otgaar, Henry; Scoboria, Alan; Smeets, Tom

2013-01-01

We report on the 1st experimental elicitation of nonbelieved memories for childhood events in adults (Study 1) and children (Study 2) using a modified false memory implantation paradigm. Participants received true (trip to a theme park) and false (hot air balloon ride) narratives and recalled these events during 2 interviews. After debriefing, 13%…

UK experience of safety requirements for thermal reactor stations

International Nuclear Information System (INIS)

Matthews, R.R.; Dale, G.C.; Tweedy, J.N.

1977-01-01

The paper summarises the development of safety requirements since the first of the Generating Boards' Magnox reactors commenced operation in 1962 and includes A.G.R. safety together with the preparation of S.G.H.W.R. design safety criteria. It outlines the basic principles originally adopted and shows how safety assessment is a continuing process throughout the life of a reactor. Some description is given of the continuous effort over the years to obtain increased safety margins for existing and new reactors, taking into account the construction and operating experience, experimental information, and more sophisticated computer-aided design techniques which have become available. The main safeguards against risks arising from the Generating Boards' reactors are the achievement of high standards of design, construction and operation, in conjunction with comprehensive fault analyses to ensure that adequate protective equipment is provided. The most important analyses refer to faults which can lead to excessive fuel element temperatures arising from an increase in power or a reduction in cooling capacity. They include the possibility of unintended control rod withdrawal at power or at start-up, coolant flow failure, pressure circuit failure, loss of boiler feed water, and failure of electric power. The paper reviews the protective equipment, and the policy for reactor safety assessments which include application of maximum credible accident philosophy and later the limited use of reliability and probability methods. Some of the Generating Boards' reactors are now more than half way through their planned working lives and during this time safety protective equipment has occasionally been brought into operation, often for spurious reasons. The general performance, of safety equipment is reviewed particularly for incidents such as main turbo-alternator trip, circulator failure, fuel element failures and other similar events, and some problems which have given rise to

Language Travel or Language Tourism: Have Educational Trips Changed So Much?

Science.gov (United States)

Laborda, Jesus Garcia

2007-01-01

This article points out the changes in organization, students and language learning that language trips, as contrasted with educational trips (of which language trips are a subgroup) have gone through in the last years. The article emphasizes the need to differentiate between language trips and language tourism based on issues of additional…

Simplified method for measuring the response time of scram release electromagnet in a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Patri, Sudheer, E-mail: patri@igcar.gov.in; Mohana, M.; Kameswari, K.; Kumar, S. Suresh; Narmadha, S.; Vijayshree, R.; Meikandamurthy, C.; Venkatesan, A.; Palanisami, K.; Murthy, D. Thirugnana; Babu, B.; Prakash, V.; Rajan, K.K.

2015-04-15

Highlights: • An alternative method for estimating the electromagnet clutch release time. • A systematic approach to develop a computer based measuring system. • Prototype tests on the measurement system. • Accuracy of the method is ±6% and repeatability error is within 2%. - Abstract: The delay time in electromagnet clutch release during a reactor trip (scram action) is an important safety parameter, having a bearing on the plant safety during various design basis events. Generally, it is measured using current decay characteristics of electromagnet coil and its energising circuit. A simplified method of measuring the same in a Sodium cooled fast reactors (SFR) is proposed in this paper. The method utilises the position data of control rod to estimate the delay time in electromagnet clutch release. A computer based real time measurement system for measuring the electromagnet clutch delay time is developed and qualified for retrofitting in prototype fast breeder reactor. Various stages involved in the development of the system are principle demonstration, experimental verification of hardware capabilities and prototype system testing. Tests on prototype system have demonstrated the satisfactory performance of the system with intended accuracy and repeatability.

Reactor shutdown system of prototype fast breeder reactor

International Nuclear Information System (INIS)

Govindarajan, S.; Singh, Om Pal; Kasinathan, N.; Paramasivan Pillai, C.; Arul, A.J.; Chetal, S.C.

2002-01-01

Full text: The shutdown system of PFBR is designed to assure a very high reliability by employing well known principles of redundancy, diversity and independence. The failure probability of the shutdown system limited to -6 / ry. Salient features of the shutdown system are: Two independent shutdown systems, each of them able to accommodate an additional single failure and made up of a trip system and an associated absorber rod group. Diversity between trip systems, rods and mechanisms. Initiation of SCRAM by two diverse physical parameters of the two shutdown systems for design events leading potentially to unacceptable conditions is the core. The first group of nine rods called control and safety rods (CSR) is used for both shutdown as well as power regulation. The second group consisting of three rods known as diverse safety rods (DSR) is used only for shutdown. Diversity between the two groups is ensured by varying the operating conditions of the electromagnets and the configurations of the mobile parts. The reactivity worth of the absorber rods have been chosen such that each group of rods would ensure cold shutdown on SCRAM even when the most reactive rod of the group fails to drop. Together the two groups ensure a shutdown margin of 5000 pcm. The speed and individual rod worth of the CSR is chosen from operational and safety considerations during reactor start up and raising of power. Required drop time of rods during SCRAM depends on the incident considered. For a severe reactivity incident of 3 $/s this has to be limited to 1s and is ensured by limiting electromagnet response time and facilitating drop by gravity. Design safety limits for core components have been determined and SCRAM parameters have been identified by plant dynamic analysis to restrict the temperatures of core components within the limits. The SCRAM parameters are distributed between the two systems appropriately. Fault tree analysis of the system has been carried out to determine the

Development of time dependent safety analysis code for plasma anomaly events in fusion reactors

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

1997-01-01

A safety analysis code SAFALY has been developed to analyze plasma anomaly events in fusion reactors, e.g., a loss of plasma control. The code is a hybrid code comprising a zero-dimensional plasma dynamics and a one-dimensional thermal analysis of in-vessel components. The code evaluates the time evolution of plasma parameters and temperature distributions of in-vessel components. As the plasma-safety interface model, we proposed a robust plasma physics model taking into account updated data for safety assessment. For example, physics safety guidelines for beta limit, density limit and H-L mode confinement transition threshold power, etc. are provided in the model. The model of the in-vessel components are divided into twenty temperature regions in the poloidal direction taking account of radiative heat transfer between each surface of each region. This code can also describe the coolant behavior under hydraulic accidents with the results by hydraulics code and treat vaporization (sublimation) from plasma facing components (PFCs). Furthermore, the code includes the model of impurity transport form PFCs by using a transport probability and a time delay. Quantitative analysis based on the model is possible for a scenario of plasma passive shutdown. We examined the possibility of the code as a safety analysis code for plasma anomaly events in fusion reactors and had a prospect that it would contribute to the safety analysis of the International Thermonuclear Experimental Reactor (ITER). (author)

A Bayesian Additive Model for Understanding Public Transport Usage in Special Events

DEFF Research Database (Denmark)

Rodrigues, Filipe; Borysov, Stanislav S.; Ribeiro, Bernardete

2017-01-01

Public special events, like sports games, concerts and festivals are well known to create disruptions in transportation systems, often catching the operators by surprise. Although these are usually planned well in advance, their impact is difficult to predict, even when organisers...... additive model with Gaussian process components that combines smart card records from public transport with context information about events that is continuously mined from the Web. We develop an efficient approximate inference algorithm using expectation propagation, which allows us to predict the total...... number of public transportation trips to the special event areas, thereby contributing to a more adaptive transportation system. Furthermore, for multiple concurrent event scenarios, the proposed algorithm is able to disaggregate gross trip counts into their most likely components related to specific...

A Trip to the Zoo: Children's Words and Photographs.

Science.gov (United States)

DeMarie, Darlene

Field trips are a regular part of many programs for young children. Field trips can serve a variety of purposes, such as exposing children to new things or helping children to see familiar things in new ways. The purpose of this study was to learn the meaning children gave to a field trip. Cameras were made available to each of the children in a…

Flow in Rotating Serpentine Coolant Passages With Skewed Trip Strips

Science.gov (United States)

Tse, David G.N.; Steuber, Gary

1996-01-01

Laser velocimetry was utilized to map the velocity field in serpentine turbine blade cooling passages with skewed trip strips. The measurements were obtained at Reynolds and Rotation numbers of 25,000 and 0.24 to assess the influence of trips, passage curvature and Coriolis force on the flow field. The interaction of the secondary flows induced by skewed trips with the passage rotation produces a swirling vortex and a corner recirculation zone. With trips skewed at +45 deg, the secondary flows remain unaltered as the cross-flow proceeds from the passage to the turn. However, the flow characteristics at these locations differ when trips are skewed at -45 deg. Changes in the flow structure are expected to augment heat transfer, in agreement with the heat transfer measurements of Johnson, et al. The present results show that trips are skewed at -45 deg in the outward flow passage and trips are skewed at +45 deg in the inward flow passage maximize heat transfer. Details of the present measurements were related to the heat transfer measurements of Johnson, et al. to relate fluid flow and heat transfer measurements.

Reactor power cutback system test experience at YGN 4

International Nuclear Information System (INIS)

Chi, Sung Goo; Kim, Se Chang; Seo, Jong Tae; Eom, Young Meen; Wook, Jeong Dae; Choi, Young Boo

1995-01-01

YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

Actual and Virtual Reality: Making the Most of Field Trips.

Science.gov (United States)

Bellan, Jennifer Marie; Scheurman, Geoffrey

1998-01-01

Argues that a virtual field trip can complement and enhance a real one. Discusses the benefits and pitfalls of both types of field trips. Outlines a series of student and teacher activities combining an actual field trip and a virtual one to Fort Snelling in St. Paul, Minnesota. (MJP)

CANDU safety analysis system establishment; development of trip coverage and multi-dimensional hydrogen analysis methodology

Energy Technology Data Exchange (ETDEWEB)

Choi, Jong Ho; Ohn, M. Y.; Cho, C. H. [KOPEC, Taejon (Korea)

2002-03-01

The trip coverage analysis model requires the geometry network for primary and secondary circuit as well as the plant control system to simulate all the possible plant operating conditions throughout the plant life. The model was validated for the power maneuvering and the Wolsong 4 commissioning test. The trip coverage map was produced for the large break loss of coolant accident and the complete loss of class IV power event. The reliable multi-dimensional hydrogen analysis requires the high capability for thermal hydraulic modelling. To acquire such a basic capability and verify the applicability of GOTHIC code, the assessment of heat transfer model, hydrogen mixing and combustion model was performed. Also, the assessment methodology for flame acceleration and deflagration-to-detonation transition is established. 22 refs., 120 figs., 31 tabs. (Author)

Vanpool trip planning based on evolutionary multiple objective optimization

Science.gov (United States)

Zhao, Ming; Yang, Disheng; Feng, Shibing; Liu, Hengchang

2017-08-01

Carpool and vanpool draw a lot of researchers’ attention, which is the emphasis of this paper. A concrete vanpool operation definition is given, based on the given definition, this paper tackles vanpool operation optimization using user experience decline index(UEDI). This paper is focused on making each user having identical UEDI and the system having minimum sum of all users’ UEDI. Three contributions are made, the first contribution is a vanpool operation scheme diagram, each component of the scheme is explained in detail. The second contribution is getting all customer’s UEDI as a set, standard deviation and sum of all users’ UEDI set are used as objectives in multiple objective optimization to decide trip start address, trip start time and trip destination address. The third contribution is a trip planning algorithm, which tries to minimize the sum of all users’ UEDI. Geographical distribution of the charging stations and utilization rate of the charging stations are considered in the trip planning process.

HOW DO YOUNG PEOPLE SELECT INFORMATION TO PLAN A TRIP

Directory of Open Access Journals (Sweden)

Oana ŢUGULEA

2013-12-01

Full Text Available The purpose of the research is to reveal the young tourists preferences in the process of planning a trip. Sources of information used, the utility of Internet/travel agencies in planning travel trip activities, preferred means of transportation and types of accommodation are investigated. As research methods, there used both qualitative and quantitative methods: focus group and survey. Internet is more used by young tourists in planning trips than travel agencies are. Internet is considered more useful in the documentation stage and when buying airline tickets. Young tourists are more influenced by friends when planning a trip. Young tourists prefer cars and planes as means of transportation for a trip and hotels and guesthouses as accommodation when traveling.

Applying Bayesian neural networks to event reconstruction in reactor neutrino experiments

International Nuclear Information System (INIS)

Xu Ye; Xu Weiwei; Meng Yixiong; Zhu Kaien; Xu Wei

2008-01-01

A toy detector has been designed to simulate central detectors in reactor neutrino experiments in the paper. The electron samples from the Monte-Carlo simulation of the toy detector have been reconstructed by the method of Bayesian neural networks (BNNs) and the standard algorithm, a maximum likelihood method (MLD), respectively. The result of the event reconstruction using BNN has been compared with the one using MLD. Compared to MLD, the uncertainties of the electron vertex are not improved, but the energy resolutions are significantly improved using BNN. And the improvement is more obvious for the high energy electrons than the low energy ones

Hybrid intelligent monironing systems for thermal power plant trips

Science.gov (United States)

Barsoum, Nader; Ismail, Firas Basim

2012-11-01

Steam boiler is one of the main equipment in thermal power plants. If the steam boiler trips it may lead to entire shutdown of the plant, which is economically burdensome. Early boiler trips monitoring is crucial to maintain normal and safe operational conditions. In the present work two artificial intelligent monitoring systems specialized in boiler trips have been proposed and coded within the MATLAB environment. The training and validation of the two systems has been performed using real operational data captured from the plant control system of selected power plant. An integrated plant data preparation framework for seven boiler trips with related operational variables has been proposed for IMSs data analysis. The first IMS represents the use of pure Artificial Neural Network system for boiler trip detection. All seven boiler trips under consideration have been detected by IMSs before or at the same time of the plant control system. The second IMS represents the use of Genetic Algorithms and Artificial Neural Networks as a hybrid intelligent system. A slightly lower root mean square error was observed in the second system which reveals that the hybrid intelligent system performed better than the pure neural network system. Also, the optimal selection of the most influencing variables performed successfully by the hybrid intelligent system.

TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

Energy Technology Data Exchange (ETDEWEB)

Sekhri, A.; Baumann, P. [KernkraftwerkLeibstadt AG, 5325 Leibstadt (Switzerland); Wicaksono, D. [Swiss Federal Inst. of Technology Zurich ETH, 8092 Zurich (Switzerland); Miro, R.; Barrachina, T.; Verdu, G. [Inst. for Industrial, Radiophysical and Environmental Safety ISIRYM, Universitat Politecnica de Valencia UPV, Cami de Vera s/n, 46021 Valencia (Spain)

2012-07-01

In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

International Nuclear Information System (INIS)

Sekhri, A.; Baumann, P.; Wicaksono, D.; Miro, R.; Barrachina, T.; Verdu, G.

2012-01-01

In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

Field trip guidebook for the post-meeting field trip: The Central Appalachians

Science.gov (United States)

Taylor, John F.; Loch, James D.; Ganis, G. Robert; Repetski, John E.; Mitchell, Charles E.; Blackmer, Gale C.; Brezinski, David K.; Goldman, Daniel; Orndorff, Randall C.; Sell, Bryan K.

2015-01-01

shale and sandstone turbidites accumulated. The foreland basin thus created would fill with progressively coarser and more shallow/proximal clastic facies through the Upper Ordovician, culminating in deposition of fluvial redbeds that cap the Taconic clastic wedge. Arguably the most controversial rocks within the Tippecanoe Sequence in this area are unusual, Lower Ordovician deep marine facies that are associated with the much younger flysch of the Martinsburg Formation in the Great Valley of eastern Pennsylvania. Long considered the erosional remnants of a Taconic-style thrust sheet, and referred to as the Hamburg Klippe, these deep marine deposits have recently been reinterpreted as olistostromal deposits that were introduced by gravity sliding into the flysch basin contemporaneous with Martinsburg deposition.Besides their constituent lithofacies, rocks of the Sauk and Tippecanoe megasequences also present a stark contrast in faunas. Cambrian and Lower Ordovician faunas predate the Great Ordovician Biodiversification Event (GOBE), a global event that saw unprecedented diversification within many major invertebrate groups (mollusks, corals, and bryozoans to name a few) that previously were only minor components of the marine fauna. Unfortunately, the much higher diversity of Middle and Upper Ordovician faunas wrought by the GOBE is somewhat muted in this region by the stresses introduced by conversion of the Appalachian shelf into a flysch basin. Another noteworthy difference between the Cambrian and Ordovician biota related to the paleogeographic setting of the rocks to be examined on this trip derives from their evolution in the shallow marine environments of Laurentia. Several shelf-wide extinctions decimated the shallow marine faunas of the Laurentian shelf through the late Cambrian producing stage-level biostratigraphic units known as biomeres. The biomere phenomenon is discussed in this guidebook and a few stops to examine Cambrian faunas and one biomere boundary

Process modeling of a reversible solid oxide cell (r-SOC) energy storage system utilizing commercially available SOC reactor

International Nuclear Information System (INIS)

Mottaghizadeh, Pegah; Santhanam, Srikanth; Heddrich, Marc P.; Friedrich, K. Andreas; Rinaldi, Fabio

2017-01-01

Highlights: • An electric energy storage system was developed based on a commercially available SOC reactor. • Heat generated in SOFC mode of r-SOC is utilized in SOEC operation of r-SOC using latent heat storage. • A round trip efficiency of 54.3% was reached for the reference system at atmospheric pressure. • An improved process system design achieved a round-trip efficiency of 60.4% at 25 bar. - Abstract: The increase of intermittent renewable energy contribution in power grids has urged us to seek means for temporal decoupling of electricity production and consumption. A reversible solid oxide cell (r-SOC) enables storage of surplus electricity through electrochemical reactions when it is in electrolysis mode. The reserved energy in form of chemical compounds is then converted to electricity when the cell operates as a fuel cell. A process system model was implemented using Aspen Plus® V8.8 based on a commercially available r-SOC reactor experimentally characterized at DLR. In this study a complete self-sustaining system configuration is designed by optimal thermal integration and balance of plant. Under reference conditions a round trip efficiency of 54.3% was achieved. Generated heat in fuel cell mode is exploited by latent heat storage tanks to enable endothermic operation of reactor in its electrolysis mode. In total, out of 100 units of thermal energy stored in heat storage tanks during fuel cell mode, 90% was utilized to offset heat demand of system in its electrolysis mode. Parametric analysis revealed the significance of heat storage tanks in thermal management even when reactor entered its exothermic mode of electrolysis. An improved process system design demonstrates a system round-trip efficiency of 60.4% at 25 bar.

Loss of Power and Water Hammer Event at San Onofre, Unit 1, on November 21, 1985

International Nuclear Information System (INIS)

1986-01-01

On November 21, 1985, Southern California Edison's Onofre Nuclear Generating Station, Unit 1, located south of San Clemente, California, experienced a partial loss of inplant ac electrical power while the plant was operating at 60% power. Following a manual reactor trip, the plant lost all inplant ac power for 4 minutes and experienced a severe incidence of water hammer in the feedwater system which caused a leak, damaged plant equipment, and challenged the integrity of the plant's heat sink. The most significant aspect of the event involved the failure of five safety-related check valves in the feed-water system whose failure occurred in less than year, without detection, and jeopardized the integrity of safety systems. The event involved a number of equipment malfunctions, operator errors, and procedural deficiencies. This report documents the findings and conclusions of an NRC Incident Investigation Team sent to San Onofre by the NRC Executive Director for Operations in conformance with NRC's recently established Incident Investigation Program

A study for the sequence of events (SOE) system on the nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Lee, Byung Chae; Jeon, Jong Sun; Lee, Sun Sung; Lee, Kyung Ho; Lee, Byung Ju; Sohn, Kwang Young [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

It is important to identify where and why an event or a trip is occurred in the Nuclear Power Plant(NPP) and to provide proper resolution against above situation. In order to analyze the prime cause or conspicuous reason of trouble occurred after events or trips occur, the Sequence of Events(SOE) system has been adopted in Korean NPP to acquire the sequential information along where and when an event or a trip take place. The SOE system of UCN 3 and 4 plant which is included in the Plant Data Acquisition System (PDAS), shares the 3205 computer and system software with PDAS. Sharing of the computer H/w and S/W, however, requires more complicated process to provide the events or trip signals due to the inherent characteristics of the shared system. Moreover there are high potentiality of collision between synchronization signals and data transmitted to the Plant Computer System (PCS), when the synchronization signals are sent from PCS to the three SOE processors. When this collision happens the SOE system will break down, thus it is not possible to analyze the trend of events or trips. An independent SOE system composed with single processor is proposed in this paper. To begin with, the analyses on the hardware and software of SOE and PDAS system of UCN 3 and 4 were performed to justify the problems and the resolution if it exists. In order to test the new SOE system, VMEbus, VM30 CPU, change of status I/O card and OS-9 for the operating system were adopted and the analysis for this test system was done as follows; the verification should be achieved through the simulation; the simulated signals for events are given the test system as inputs and the outputs are monitored to verify whether the sequential events logging function works well or not on PC. In conclusion, this report is expected to provide the technical background for the improvement and changing of the NPP PDAS and SOE system in the future. 18 tabs., 33 figs., 26 refs. (Author) .new.

A study for the sequence of events (SOE) system on the nuclear power plant

International Nuclear Information System (INIS)

Lee, Byung Chae; Jeon, Jong Sun; Lee, Sun Sung; Lee, Kyung Ho; Lee, Byung Ju; Sohn, Kwang Young

1996-06-01

It is important to identify where and why an event or a trip is occurred in the Nuclear Power Plant(NPP) and to provide proper resolution against above situation. In order to analyze the prime cause or conspicuous reason of trouble occurred after events or trips occur, the Sequence of Events(SOE) system has been adopted in Korean NPP to acquire the sequential information along where and when an event or a trip take place. The SOE system of UCN 3 and 4 plant which is included in the Plant Data Acquisition System (PDAS), shares the 3205 computer and system software with PDAS. Sharing of the computer H/w and S/W, however, requires more complicated process to provide the events or trip signals due to the inherent characteristics of the shared system. Moreover there are high potentiality of collision between synchronization signals and data transmitted to the Plant Computer System (PCS), when the synchronization signals are sent from PCS to the three SOE processors. When this collision happens the SOE system will break down, thus it is not possible to analyze the trend of events or trips. An independent SOE system composed with single processor is proposed in this paper. To begin with, the analyses on the hardware and software of SOE and PDAS system of UCN 3 and 4 were performed to justify the problems and the resolution if it exists. In order to test the new SOE system, VMEbus, VM30 CPU, change of status I/O card and OS-9 for the operating system were adopted and the analysis for this test system was done as follows; the verification should be achieved through the simulation; the simulated signals for events are given the test system as inputs and the outputs are monitored to verify whether the sequential events logging function works well or not on PC. In conclusion, this report is expected to provide the technical background for the improvement and changing of the NPP PDAS and SOE system in the future. 18 tabs., 33 figs., 26 refs. (Author) .new

Improving the peak power density estimation for the DNBR trip signal

International Nuclear Information System (INIS)

Moreira, Joao M. L.; Souza, Rose Mary G.P.

2002-01-01

The departure from nucleate boiling (DNB) core protection in PWR reactors is usually carried out through the over temperature trip or the instantaneous minimum DNB ratio (DNBR) trip. The protection is obtained through specialized correlations or fast digital computer simulators that infer the core power level, and local coolant thermal and flow conditions out of process variables furnished by the instrumentation. The power density distribution information is usually expressed in terms of F q , the power peak factor, and its location. F q , in its turn, can be determined through the control rod position or, more often, through the power axial offset (AO) F q =f (AO, control rod positions). The AO, defined as the difference between upper and lower long ion chambers signals, is supplied for each channel by separate sets of out-of-core detectors positioned 90 or 120 degrees apart in plan. The AO is given by AO=(S t -S b )/(S t +S b ) where S t and S b are the out-of-core signals from the top and the bottom sections, respectively. In current PWRs a large penalty is imposed to the result of the first equation, because of the difficult of inferring with good accuracy the peak factor from the AO obtained from the out-of-core instrumentation. This ends up reducing the plant capacity factor. In this work, the f function in the first equation, which correlates the power peak factor with the axial offset yielded by out-of-core detectors and control rod positions, is obtained through a combination of specific experiments in the IPEN/MB-01 zero-power reactor and calculation results. For improving the peak factor estimation, it is necessary to consider accurately the response of the out-of-core detectors to different power density distribution in the core. This task is not easily accomplished through calculation due to the difficulties involved in the necessary neutron transport treatment for the out-of-core detector responses

ROP design for Enhanced CANDU 6 reactor

Energy Technology Data Exchange (ETDEWEB)

Hu, J.; Scherbakova, D; Kastanya, D.; Ovanes, M. [Candu Energy Inc., Mississauga, Ontario (Canada)

2011-07-01

The Enhanced CANDU 6 (EC6) nuclear power plant is a mid-sized pressurized heavy water reactor design, based on the highly successful CANDU 6 (C6) family of power plants, upgraded to meet today's Canadian and international safety requirements and to satisfy Generation III expectations. The EC6 reactor is equipped with two independent Regional Overpower Protection (ROP) systems to prevent overpowers in the reactor fuel. The ROP system design, retaining the traditional C6 methodology, is determined to cover the End-of-Life (EOL) reactor core condition since the reactor operating/thermal margin gradually decreases as plant equipment ages. Several design changes have been incorporated into the reference C6 plant to mitigate the ageing effect on the ROP trip margin. This paper outlines the basis for the EC6 ROP physics design and presents the ROP related improvements made in the EC6 design to ensure that full power operation is not limited by the ROP throughout the entire life of the reactor. (author)

Appraisal of boundary layer trips for landing gear testing

Science.gov (United States)

McCarthy, Philip; Feltham, Graham; Ekmekci, Alis

2013-11-01

Dynamic similarity during scaled model testing is difficult to maintain. Forced boundary layer transition via a surface protuberance is a common method used to address this issue, however few guidelines exist for the effective tripping of complex geometries, such as aircraft landing gears. To address this shortcoming, preliminary wind tunnel tests were performed at Re = 500,000. Surface transition visualisation and pressure measurements show that zigzag type trips of a given size and location are effective at promoting transition, thus preventing the formation of laminar separation bubbles and increasing the effective Reynolds number from the critical regime to the supercritical regime. Extension of these experiments to include three additional tripping methods (wires, roughness strips, CADCUT dots) in a range of sizes, at Reynolds number of 200,000 and below, have been performed in a recirculating water channel. Analysis of surface pressure measurements and time resolved PIV for each trip device, size and location has established a set of recommendations for successful use of tripping for future, low Reynolds number landing gear testing.

Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Ahn, Sang June; Ha, Kwi Seok; Chang, Won Pyo; Kang, Seok Hun; Lee, Kwi Lim; Choi, Chi Woong; Lee, Seung Won; Yoo, Jin; Jeong, Jae Ho; Jeong, Tae Kyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-08-15

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

Evaluation of a Sodium–Water Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

Directory of Open Access Journals (Sweden)

Sang June Ahn

2016-08-01

Full Text Available The prototype generation IV sodium-cooled fast reactor (PGSFR has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS and the safety of the primary heat-transfer system (PHTS. In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

An intelligent safety system concept for future CANDU reactors

International Nuclear Information System (INIS)

Hinds, H.W.

1980-01-01

A review of the current Regional Over-power Trip (ROPT) system employed on the Bruce NGS-A reactors confirmed the belief that future reactors should have an improved ROPT system. We are developing such an 'intelligent' safety system. It uses more of the available information on reactor status and employs modern computer technology. Fast triplicated safety computers compute maps of fuel channel power, based on readings from prompt-responding flux detectors. The coefficients for this calculation are downloaded periodically from a fourth supervisor computer. These coefficients are based on a detailed 3-D flux shape derived from physics data and other plant information. A demonstration of one of three safety channels of such a system is planned. (auth)

Geologic field-trip guide to Long Valley Caldera, California

Science.gov (United States)

Hildreth, Wes; Fierstein, Judy

2017-07-26

This guide to the geology of Long Valley Caldera is presented in four parts: (1) An overview of the volcanic geology; (2) a chronological summary of the principal geologic events; (3) a road log with directions and descriptions for 38 field-trip stops; and (4) a summary of the geophysical unrest since 1978 and discussion of its causes. The sequence of stops is arranged as a four-day excursion for the quadrennial General Assembly of the International Association of Volcanology and Chemistry of the Earth’s Interior (IAVCEI), centered in Portland, Oregon, in August 2017. Most stops, however, are written freestanding, with directions that allow each one to be visited independently, in any order selected.

Accommodation of the spinal cat to a tripping perturbation

Directory of Open Access Journals (Sweden)

Hui eZhong

2012-05-01

Full Text Available Adult cats with a complete spinal cord transection at T12-T13 can relearn over a period of days-to-weeks how to generate full weight-bearing stepping on a treadmill or standing ability if trained specifically for that task. In the present study, we assessed short-term (msec-min adaptations by repetitively imposing a mechanical perturbation on the hindlimb of chronic spinal cats by placing a rod in the path of the leg during the swing phase to trigger a tripping response. The kinematics and EMG were recorded during control (10 steps, trip (1 to 60 steps with various patterns and then release (without any tripping stimulus, 10 to 20 steps sequences. Our data show that the activation patterns and kinematics of the hindlimb in the step cycle immediately following the initial trip (mechanosensory stimulation of the dorsal surface of the paw was modified in a way that increased the probability of avoiding the obstacle in the subsequent step. This indicates that the spinal sensorimotor circuitry reprogrammed the trajectory of the swing following a perturbation prior to the initiation of the swing phase of the subsequent step, in effect attempting to avoid the re-occurrence of the perturbation. The average height of the release steps was elevated compared to control regardless of the pattern and the length of the trip sequences. In addition, the average impact force on the tripping rod tended to be lower with repeated exposure to the tripping stimulus. EMG recordings suggest that the semitendinosus, a primary knee flexor, was a major contributor to the adaptive tripping response. These results demonstrate that the lumbosacral locomotor circuitry can modulate the activation patterns of the hindlimb motor pools within the time frame of single step in a manner that tends to minimize repeated perturbations. Furthermore, these adaptations remained evident for a number of steps after removal of the mechanosensory stimulation.

Application of best estimate and uncertainty safety analysis methodology to loss of flow events at Ontario's Power Generation's Darlington Nuclear Generating Station

International Nuclear Information System (INIS)

Huget, R.G.; Lau, D.K.; Luxat, J.C.

2001-01-01

Ontario Power Generation (OPG) is currently developing a new safety analysis methodology based on best estimate and uncertainty (BEAU) analysis. The framework and elements of the new safety analysis methodology are defined. The evolution of safety analysis technology at OPG has been thoroughly documented. Over the years, the use of conservative limiting assumptions in OPG safety analyses has led to gradual erosion of predicted safety margins. The main purpose of the new methodology is to provide a more realistic quantification of safety margins within a probabilistic framework, using best estimate results, with an integrated accounting of the underlying uncertainties. Another objective of the new methodology is to provide a cost-effective means for on-going safety analysis support of OPG's nuclear generating stations. Discovery issues and plant aging effects require that the safety analyses be periodically revised and, in the past, the cost of reanalysis at OPG has been significant. As OPG enters the new competitive marketplace for electricity, there is a strong need to conduct safety analysis in a less cumbersome manner. This paper presents the results of the first licensing application of the new methodology in support of planned design modifications to the shutdown systems (SDSs) at Darlington Nuclear Generating Station (NGS). The design modifications restore dual trip parameter coverage over the full range of reactor power for certain postulated loss-of-flow (LOF) events. The application of BEAU analysis to the single heat transport pump trip event provides a realistic estimation of the safety margins for the primary and backup trip parameters. These margins are significantly larger than those predicted by conventional limit of the operating envelope (LOE) analysis techniques. (author)

Siting of research reactors

International Nuclear Information System (INIS)

1987-01-01

The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

MCMII and the TriP chip

Energy Technology Data Exchange (ETDEWEB)

Juan Estrada et al.

2003-12-19

We describe the development of the electronics that will be used to read out the Fiber Tracker and Preshower detectors in Run IIb. This electronics is needed for operation at 132ns bunch crossing, and may provide a measurement of the z coordinate of the Fiber Tracker hits when operating at 396ns bunch crossing. Specifically, we describe the design and preliminary tests of the Trip chip, MCM IIa, MCM IIb and MCM IIc. This document also serves as a user manual for the Trip chip and the MCM.

Progress of the DUPIC fuel compatibility analysis (I) - reactor physics

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Jeong, Chang Joon; Roh, Gyu Hong; Rhee, Bo Wook; Park, Jee Won

2003-12-01

Since 1992, the direct use of spent pressurized water reactor fuel in CANada Deuterium Uranium (CANDU) reactors (DUPIC) has been studied as an alternative to the once-through fuel cycle. The DUPIC fuel cycle study is focused on the technical feasibility analysis, the fabrication of DUPIC fuels for irradiation tests and the demonstration of the DUPIC fuel performance. The feasibility analysis was conducted for the compatibility of the DUPIC fuel with existing CANDU-6 reactors from the viewpoints of reactor physics, reactor safety, fuel cycle economics, etc. This study has summarized the intermediate results of the DUPIC fuel compatibility analysis, which includes the CANDU reactor physics design requirements, DUPIC fuel core physics design method, performance of the DUPIC fuel core, regional overpower trip setpoint, and the CANDU primary shielding. The physics analysis showed that the CANDU-6 reactor can accommodate the DUPIC fuel without deteriorating the physics design requirements by adjusting the fuel management scheme if the fissile content of the DUPIC fuel is tightly controlled.

WIPP site and vicinity geological field trip

International Nuclear Information System (INIS)

Chaturvedi, L.

1980-10-01

The Environmental Evaluation Group (EEG) is conducting an assessment of the radiological health risks to people from the Waste Isolation Pilot Plant (WIPP). As a part of this work, EEG is making an effort to improve the understanding of those geological issues concerning the WIPP site which may affect the radiological consequences of the proposed repository. One of the important geological issues to be resolved is the timing and the nature of the dissolution processes which may have affected the WIPP site. EEG organized a two-day conference of geological scientists, titled Geotechnical Considerations for Radiological Hazard Assessment of WIPP on January 17-18, 1980. During this conference, it was realized that a field trip to the site would further clarify the different views on the geological processes active at the site. The field trip of June 16-18, 1980 was organized for this purpose. This report provides a summary of the field trip activities along with the participants post field trip comments. Important field stops are briefly described, followed by a more detailed discussion of critical geological issues. The report concludes with EEG's summary and recommendations to the US Department of Energy for further information needed to more adequately resolve concerns for the geologic and hydrologic integrity of the site

The moderating role of shopping trip type in store satisfaction formation

NARCIS (Netherlands)

Hunneman, Auke; Verhoef, Pieter; Sloot, Laurentius

Consumers may weigh store attributes differently depending on the type of shopping trip. For example, fill-in shoppers likely value convenience, due to the ad-hoc nature and urgency of such trips. However, no study has yet explored the effects of shopping trip types on satisfaction formation. This

SAME-DAY TRIPS: A CHANCE OF URBAN DESTINATION DEVELOPMENT

Directory of Open Access Journals (Sweden)

Dario Simicevic

2011-12-01

Full Text Available The global economic crisis, the decline of standard and climatic factors influence the allocation of tourism trends at the global level. Certain types of tourist movements start up and develop; they have been present, but not sufficiently studied by authors. They also include a short trip or visit to a particular destination. Considering their characteristics, they do not require a lot of money and they make an increasingly important segment of the tourism market. Therefore, the importance of same-day trips should not be neglected on today's tourism market. Although in practice this part of the tourist offers and demand has not often been attached enough importance, same day trip can achieve a very significant inflow of funds and encourage the development of many potential tourist destinations. For all the reasons mentioned above, and because of its importance, the organization of same day-trips should be the fundamental basis and essential focus for tourism development. Taking into consideration that inbound tourist agencies show special interest for same-day trips, we have tried to give a starting point for further research in this part of the tourism market.

Comparative study of the Peach Bottom turbine trip experiment using two different coupled codes approaches

International Nuclear Information System (INIS)

Bambara, M.; Bousbia-Salah, A.; D'Auria, F.

2005-01-01

Full text of publication follows: In the last years a great concern about the neutron-3D/thermal-hydraulic codes coupling took place. Owing to the improved computational technology, 'best estimate' analyses are today a common tool to assess safety features, and they are necessary if an asymmetric behaviour in the core region exists, or if strong interactions between the core neutronics and reactor thermal-hydraulic occur. In order to validate the coupled codes performances, several international programmes were issued. Among these activities, the OECD/NEA BWR Turbine Trip (TT) was chosen for further sensitivity analyses. It consists of a turbine trip (TT) experiment carried out at the Peach Bottom 2 BWR. In this paper, the results of two different coupled codes systems are summarized and compared. The BWR TT simulations were carried out coupling the thermal-hydraulic system code RELAP5/mode 3.2 to the 3D neutron kinetics code Parcs/2.3, and also the system code ATHLET to the neutronics code QUABOX-CUBBOX. An exhaustive overview of the main features is given, and those aspects, which need further developments and experiences, are pointed out. (authors)

Probabilities of inherent shutdown of unprotected events in innovative liquid metal reactors

International Nuclear Information System (INIS)

Mueller, C.J.; Wade, D.C.

1988-01-01

The uncertainty in predicting the effectiveness of inherent shutdown in innovative liquid metal cooled reactors with metallic fuel results from three broad contributing areas of uncertainty: (1) the inability to exactly predict the frequency of ATWS events with potential to challenge the safety systems and require inherent shutdown; (2) the approximation of representing all such events by a selected set of ''generic scenarios''; and (3) the inability to exactly calculate the core response to the selected generic scenarios. This paper discusses the work being done to address each of these contributing areas, identifies the design and research approaches being used at Argonne National Laboratory to reducing the key contributions to uncertainties in inherent shutdown, and presents results. The conditional probabilities (given ATWS initiation) of achieving temperatures capable of defeating inherent shutdown are shown to range from /approximately/0.1% to negligible for current designs

Fail-safe design criteria for computer-based reactor protection systems

International Nuclear Information System (INIS)

Keats, A.B.

1980-01-01

The increasing quantity and complexity of the instrumentation required in nuclear power plants provides a strong incentive for using on-line computers as the basis of the control and protection systems. On-line computers using multiplexed sampled data are already well established but their application to nuclear reactor protection systems requires special measures to satisfy the very high reliability which is demanded in the interests of safety and availability. Some existing codes of practice relating to segregation of replicated subsysttems continue to be applicable and lead to division of the computer functions into two distinct parts. The first computer, referred to as the Trip Algorithm Computer may also control the multiplexer. Voting on each group of status inputs yielded by the trip algorithm computers is performed by the Vote Algorithm Computer. The conceptual disparities between hardwired reactor-protection systems and those employing computers also rise to a need for some new criteria. An important objective of these criteria, minimising the need for a failure-mode-and-effect-analysis of the computer software, but is achieved almost entirely by 'hardware' properties of the system: the systematic use of hardwired test inputs which cause excursions of the trip algorithms into the tripped state in a uniquely ordered but easily recognisable sequence, and the use of hardwired 'pattern recognition logic' which generates a dynamic 'healthy' stimulus for the shutdown actuators only in response to the unique sequence generated by the hardwired input signal pattern. The adoption of the proposed design criteria ensure not only failure-to-safety in the hardware but the elimination, or at least minimisation, of the dependence on the correct functioning of the computer software for the safety system. (auth)

Analysis of loss of offsite power events reported in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Volkanovski, Andrija, E-mail: Andrija.VOLKANOVSKI@ec.europa.eu [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Ballesteros Avila, Antonio; Peinador Veira, Miguel [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Kančev, Duško [Kernkraftwerk Goesgen-Daeniken AG, CH-4658 Daeniken (Switzerland); Maqua, Michael [Gesellschaft für Anlagen-und-Reaktorsicherheit (GRS) gGmbH, Schwertnergasse 1, 50667 Köln (Germany); Stephan, Jean-Luc [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), BP 17 – 92262 Fontenay-aux-Roses Cedex (France)

2016-10-15

Highlights: • Loss of offsite power events were identified in four databases. • Engineering analysis of relevant events was done. • The dominant root cause for LOOP are human failures. • Improved maintenance procedures can decrease the number of LOOP events. - Abstract: This paper presents the results of analysis of the loss of offsite power events (LOOP) in four databases of operational events. The screened databases include: the Gesellschaft für Anlagen und Reaktorsicherheit mbH (GRS) and Institut de Radioprotection et de Sûreté Nucléaire (IRSN) databases, the IAEA International Reporting System for Operating Experience (IRS) and the U.S. Licensee Event Reports (LER). In total 228 relevant loss of offsite power events were identified in the IRSN database, 190 in the GRS database, 120 in U.S. LER and 52 in IRS database. Identified events were classified in predefined categories. Obtained results show that the largest percentage of LOOP events is registered during On power operational mode and lasted for two minutes or more. The plant centered events is the main contributor to LOOP events identified in IRSN, GRS and IAEA IRS database. The switchyard centered events are the main contributor in events registered in the NRC LER database. The main type of failed equipment is switchyard failures in IRSN and IAEA IRS, main or secondary lines in NRC LER and busbar failures in GRS database. The dominant root cause for the LOOP events are human failures during test, inspection and maintenance followed by human failures due to the insufficient or wrong procedures. The largest number of LOOP events resulted in reactor trip followed by EDG start. The actions that can result in reduction of the number of LOOP events and minimize consequences on plant safety are identified and presented.

Reactivity monitoring for safety purposes on the UK prototype fast reactor

International Nuclear Information System (INIS)

Lord, D.J.; Wilkes, D.J.

1987-01-01

The small size and high rating of the liquid metal cooled fast breeder reactor (LMFBR) make the provision of safety related instrumentation for individual subassemblies both difficult and expensive. Global monitoring of the core is thus very attractive. Reactivity monitoring is an important part of such global monitoring. Reactivity monitoring on a short timescale (a few seconds) is used on the UK Prototype Fast Reactor (PFR) as a trip parameter and long-term reactivity monitoring is being developed as a means of providing early warning of slowly developing faults. Results are presented from PFR to demonstrate the capabilities of reactivity monitoring in an operational fast reactor power station. (author)

Simulation test of PIUS-type reactor with large scale experimental apparatus

International Nuclear Information System (INIS)

Tamaki, M.; Tsuji, Y.; Ito, T.; Tasaka, K.; Kukita, Yutaka

1995-01-01

A large scale experimental apparatus for simulating the PIUS-type reactor has been constructed keeping the volumetric scaling ratio to the realistic reactor model. Fundamental experiments such as a steady state operation and a pump trip simulation were performed. Experimental results were compared with those obtained by the small scale apparatus in JAERI. We have already reported the effectiveness of the feedback control for the primary loop pump speed (PI control) for the stable operation. In this paper this feedback system is modified and the PID control is introduced. This new system worked well for the operation of the PIUS-type reactor even in a rapid transient condition. (author)

Trace analysis of loss of feedwater flow event in Lungmen ABWR

International Nuclear Information System (INIS)

Wang Jongrong; Lin Haotzu; Wang Weichen; Yang Shuming; Shih Chunkuan

2009-01-01

TRACE (TRAC/RELAP Advanced Computational Engine) model of Lungmen Nuclear Power Plant was used to analyze the Loss of Feedwater Flow transient as defined in Lungmen FSAR Chapter 15. The results were compared with those from FSAR and RETRAN02. Lungmen TRACE model will have two models: In model A, vessel is divided into 11 axial levels, 4 radial rings and 1 azimuthal sectors; In model B, vessel is divided into 11 axial levels, 4 radial rings, and 6 azimuthal sectors. The above models include feedwater control system, narrow range water level control system, and wide range water level control system. The loss of feedwater flow (LOFW) transient began with the trip of two operating feedwater pumps either from the pump mechanical/electric failure, or the operator human error, or high water level signal. Feedwater flow was assumed to descend to 0 in 5 seconds and led to the decrease of reactor water level. At L3 low water level setpoint, the system actuated reactor scram signal and RIP trip signal for RIPs not connected to the M/G set. At L2 low-low water level setpoint, the system would trip the other six RIPs. This paper compares those important thermal parameters at steady state, such as the dome pressure and temperature of reactor vessel, steam flow, feedwater flow, core flow, and RIP flow, etc.. It also compares system parameters under transient conditions, such as core thermal power, core flow, steam flow, feedwater flow, Narrow Range Water Level (NRWL), Wide Range Water Level (WRWL) and RIP flow, etc.. It was concluded that the steady state and transient results of TRACE calculations are in good agreement with those from RETRAN02. In summary, our studies concluded that Lungmen TRACE model is correct and accurate enough for future safety analysis applications. (author)

Managing the effect of TRIPS on availability of priority vaccines.

Science.gov (United States)

Milstien, Julie; Kaddar, Miloud

2006-05-01

The stated purpose of intellectual property protection is to stimulate innovation. The Agreement on Trade-Related Aspects of Intellectual Property Rights (TRIPS) requires all Members of the World Trade Organization (WTO) to enact national laws conferring minimum standards of intellectual property protection by certain deadlines. Critics of the Agreement fear that such action is inconsistent with ensuring access to medicines in the developing world. A WHO convened meeting on intellectual property rights and vaccines in developing countries, on which this paper is based, found no evidence that TRIPS has stimulated innovation in developing market vaccine development (where markets are weak) or that protection of intellectual property rights has had a negative effect on access to vaccines. However, access to future vaccines in the developing world could be threatened by compliance with TRIPS. The management of such threats requires adherence of all countries to the Doha Declaration on TRIPS, and the protections guaranteed by the Agreement itself, vigilance on TRIPS-plus elements of free trade agreements, developing frameworks for licensing and technology transfer, and promoting innovative vaccine development in developing countries. The role of international organizations in defining best practices, dissemination of information, and monitoring TRIPS impact will be crucial to ensuring optimal access to priority new vaccines for the developing world.

Unveiling E-Bike Potential for Commuting Trips from GPS Traces

Directory of Open Access Journals (Sweden)

Angel J. Lopez

2017-06-01

Full Text Available Common goals of sustainable mobility approaches are to reduce the need for travel, to facilitate modal shifts, to decrease trip distances and to improve energy efficiency in the transportation systems. Among these issues, modal shift plays an important role for the adoption of vehicles with fewer or zero emissions. Nowadays, the electric bike (e-bike is becoming a valid alternative to cars in urban areas. However, to promote modal shift, a better understanding of the mobility behaviour of e-bike users is required. In this paper, we investigate the mobility habits of e-bikers using GPS data collected in Belgium from 2014 to 2015. By analysing more than 10,000 trips, we provide insights about e-bike trip features such as: distance, duration and speed. In addition, we offer a deep look into which routes are preferred by bike owners in terms of their physical characteristics and how weather influences e-bike usage. Results show that trips with higher travel distances are performed during working days and are correlated with higher average speeds. Usage patterns extracted from our data set also indicate that e-bikes are preferred for commuting (home-work and business (work related trips rather than for recreational trips.

Analysis of the accident at Fukushima Daiichi nuclear power plant in an A BWR reactor; Analisis del accidente de la planta nucleoelectrica de Fukushima Daiichi en un reactor tipo ABWR

Energy Technology Data Exchange (ETDEWEB)

Escorcia O, D. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Salazar S, E., E-mail: daniel.escorcia.ortiz@gmail.com [UNAM, Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

2016-09-15

The present work aims to recreate the accident occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, making use of an academic simulator of forced circulation of the A BWR reactor provided by the IAEA to know the scope of this simulator. The simulator was developed and distributed by the IAEA for academic purposes and contains the characteristics and general elements of this reactor to be able to simulate transients and failures of different types, allowing also to observe the general behavior of the reactor, as well as several phenomena and present systems in the same. Is an educational tool of great value, but it does not have a scope that allows the training of plant operators. To recreate the conditions of the Fukushima accident in the simulator, we first have to know what events led to this accident, as well as the actions taken by operators and managers to reduce the consequences of this accident; and the sequence of events that occurred during the course of the accident. Differences in the nuclear power plant behavior are observed and interpreted throughout the simulation, since the Fukushima plant technology and the simulator technology are not the same, although they have several elements in common. The Fukushima plant had an event that by far exceeded the design basis, which triggered in an accident that occurred in the first place by a total loss of power supply, followed by the loss of cooling systems, causing a level too high in temperature, melting the core and damaging the containment accordingly, allowing the escape of hydrogen and radioactive material. As a result of the simulation, was determined that the scope of the IAEA academic simulator reaches the entrance of the emergency equipment, so is able to simulate almost all the events occurred at the time of the earthquake and the arrival of the tsunami in the nuclear power plant of Fukushima Daiichi. However, due to its characteristics, is not able to simulate later

Some notes on the big trip

International Nuclear Information System (INIS)

Gonzalez-Diaz, Pedro F.

2006-01-01

The big trip is a cosmological process thought to occur in the future by which the entire universe would be engulfed inside a gigantic wormhole and might travel through it along space and time. In this Letter we discuss different arguments that have been raised against the viability of that process, reaching the conclusions that the process can actually occur by accretion of phantom energy onto the wormholes and that it is stable and might occur in the global context of a multiverse model. We finally argue that the big trip does not contradict any holographic bounds on entropy and information

Some notes on the big trip

Energy Technology Data Exchange (ETDEWEB)

Gonzalez-Diaz, Pedro F. [Colina de los Chopos, Centro de Fisica ' Miguel A. Catalan' , Instituto de Matematicas y Fisica Fundamental, Consejo Superior de Investigaciones Cientificas, Serrano 121, 28006 Madrid (Spain)]. E-mail: pedrogonzalez@mi.madritel.es

2006-03-30

The big trip is a cosmological process thought to occur in the future by which the entire universe would be engulfed inside a gigantic wormhole and might travel through it along space and time. In this Letter we discuss different arguments that have been raised against the viability of that process, reaching the conclusions that the process can actually occur by accretion of phantom energy onto the wormholes and that it is stable and might occur in the global context of a multiverse model. We finally argue that the big trip does not contradict any holographic bounds on entropy and information.

Trip attraction rates of shopping centers in Northern New Castle County, Delaware.

Science.gov (United States)

2004-07-01

This report presents the trip attraction rates of the shopping centers in Northern New : Castle County in Delaware. The study aims to provide an alternative to ITE Trip : Generation Manual (1997) for computing the trip attraction of shopping centers ...

Pressurizer pump reliability analysis high flux isotope reactor

International Nuclear Information System (INIS)

Merryman, L.; Christie, B.

1993-01-01

During a prolonged outage from November 1986 to May 1990, numerous changes were made at the High Flux Isotope Reactor (HFIR). Some of these changes involved the pressurizer pumps. An analysis was performed to calculate the impact of these changes on the pressurizer system availability. The analysis showed that the availability of the pressurizer system dropped from essentially 100% to approximately 96%. The primary reason for the decrease in availability comes because off-site power grid disturbances sometimes result in a reactor trip with the present pressurizer pump configuration. Changes are being made to the present pressurizer pump configuration to regain some of the lost availability

Safety Evaluation of Kartini Reactor Based on Instrumentation System Design

International Nuclear Information System (INIS)

Tjipta Suhaemi; Djen Djen Dj; Itjeu K; Johnny S; Setyono

2003-01-01

The safety of Kartini reactor has been evaluated based on instrumentation system aspect. The Kartini reactor is designed by BATAN. Design power of the reactor is 250 kW, but it is currently operated at 100 kW. Instrumentation and control system function is to monitor and control the reactor operation. Instrumentation and control system consists of safety system, start-up and automatic power control, and process information system. The linear power channel and logarithmic power channel are used for measuring power. There are 3 types of control rod for controlling the power, i.e. safety rod, shim rod, and regulating rod. The trip and interlock system are used for safety. There are instrumentation equipment used for measuring radiation exposure, flow rate, temperature and conductivity of fluid The system of Kartini reactor has been developed by introducing a process information system, start-up system, and automatic power control. It is concluded that the instrumentation of Kartini reactor has followed the requirement and standard of IAEA. (author)

Hybrid Intelligent Warning System for Boiler tube Leak Trips

Directory of Open Access Journals (Sweden)

Singh Deshvin

2017-01-01

Full Text Available Repeated boiler tube leak trips in coal fired power plants can increase operating cost significantly. An early detection and diagnosis of boiler trips is essential for continuous safe operations in the plant. In this study two artificial intelligent monitoring systems specialized in boiler tube leak trips have been proposed. The first intelligent warning system (IWS-1 represents the use of pure artificial neural network system whereas the second intelligent warning system (IWS-2 represents merging of genetic algorithms and artificial neural networks as a hybrid intelligent system. The Extreme Learning Machine (ELM methodology was also adopted in IWS-1 and compared with traditional training algorithms. Genetic algorithm (GA was adopted in IWS-2 to optimize the ANN topology and the boiler parameters. An integrated data preparation framework was established for 3 real cases of boiler tube leak trip based on a thermal power plant in Malaysia. Both the IWSs were developed using MATLAB coding for training and validation. The hybrid IWS-2 performed better than IWS-1.The developed system was validated to be able to predict trips before the plant monitoring system. The proposed artificial intelligent system could be adopted as a reliable monitoring system of the thermal power plant boilers.

Pre-Trip Notification Database (PTNS)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — The PTNS contains pre-trip notification data from vessels participating in the Northeast Multispecies groundfish fishery from 2010 to present and the Longfin squid...

A Novel Trip Coverage Index for Transit Accessibility Assessment Using Mobile Phone Data

Directory of Open Access Journals (Sweden)

Zhengyi Cai

2017-01-01

Full Text Available Transit accessibility is an important measure on the service performance of transit systems. To assess whether the public transit service is well accessible for trips of specific origins, destinations, and origin-destination (OD pairs, a novel measure, the Trip Coverage Index (TCI, is proposed in this paper. TCI considers both the transit trip coverage and spatial distribution of individual travel demands. Massive trips between cellular base stations are estimated by using over four-million mobile phone users. An easy-to-implement method is also developed to extract the transit information and driving routes for millions of requests. Then the trip coverage of each OD pair is calculated. For demonstrative purposes, TCI is applied to the transit network of Hangzhou, China. The results show that TCI represents the better transit trip coverage and provides a more powerful assessment tool of transit quality of service. Since the calculation is based on trips of all modes, but not only the transit trips, TCI offers an overall accessibility for the transit system performance. It enables decision makers to assess transit accessibility in a finer-grained manner on the individual trip level and can be well transformed to measure transit services of other cities.

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)

CNMI Commercial Purchases (Trip Ticket)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — The Commonwealth of Northern Mariana Islands (CNMI), Division of Fish and Wildlife (DFW) collects 'Trip Ticket' or purchase invoice data from vendors that buy fish...

Transcription regulator TRIP-Br2 mediates ER stress-induced brown adipocytes dysfunction.

Science.gov (United States)

Qiang, Guifen; Whang Kong, Hyerim; Gil, Victoria; Liew, Chong Wee

2017-01-09

In contrast to white adipose tissue, brown adipose tissue (BAT) is known to play critical roles for both basal and inducible energy expenditure. Obesity is associated with reduction of BAT function; however, it is not well understood how obesity promotes BAT dysfunction, especially at the molecular level. Here we show that the transcription regulator TRIP-Br2 mediates ER stress-induced inhibition of lipolysis and thermogenesis in BAT. Using in vitro, ex vivo, and in vivo approaches, we demonstrate that obesity-induced inflammation upregulates brown adipocytes TRIP-Br2 expression via the ER stress pathway and amelioration of ER stress in mice completely abolishes high fat diet-induced upregulation of TRIP-Br2 in BAT. We find that increased TRIP-Br2 significantly inhibits brown adipocytes thermogenesis. Finally, we show that ablation of TRIP-Br2 ameliorates ER stress-induced inhibition on lipolysis, fatty acid oxidation, oxidative metabolism, and thermogenesis in brown adipocytes. Taken together, our current study demonstrates a role for TRIP-Br2 in ER stress-induced BAT dysfunction, and inhibiting TRIP-Br2 could be a potential approach for counteracting obesity-induced BAT dysfunction.

Trip Generations at “Polyclinic” Land Use Type in Johor Bahru, Malaysia

OpenAIRE

Ahmed, Ishtiaque; Abdulrahman, Suleiman; Hainin, Mohd Rosli; Hassan, Sitti Asmah

2014-01-01

Transportation planners need to estimate the trip generations of different land use types in the travel demand forecasting process. The Trip Generation Manual of Malaysia, similar to the Trip Generation Manual of the Institute of Transportation Engineers, USA, provides the trip generation rate at “Polyclinics” as a function of the Gross Floor Area. However, the data for this rate have no line of best fit resulting in the lack of confidence in the prediction. This study considered ten location...

Influence of the CVCS Modelling on Results of the Loss of Offsite Power (LOOP) Safety Analysis for NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Bajs, T.; Debrecin, N.

2006-01-01

A Loss of Offsite Power (LOOP) transient scenario is based on a complete loss of non-emergency AC power that results in the loss of all power to the plant auxiliaries, i.e., the Reactor Coolant Pumps (RCPs), condensate pumps, etc. An actual LOOP event would cause a loss of all feedwater, a loss of forced Reactor Coolant System (RCS) flow and a reactor trip within less than 2 seconds as a result of either loss of power to the rod cluster assembly gripper coils or any RCS flow trips. For safety analysis purposes the LOOP event is conservatively modelled as a Loss of Normal Feedwater (LONF) transient with a subsequent loss of offsite power as a result of a reactor trip. The reactor trip followed by RCP trip are delayed until a low-low Steam Generator (SG) level signal is reached. This is a more conservative scenario than the LOOP event because the least amount of SG secondary side water mass available for heat removal and the increased amount of the stored energy in the primary circuit at the time of the loss of RCS flow result. The standard LOOP safety analysis is aimed to demonstrate the natural circulation capability of the RCS to remove residual and decay heat from the core aided by Auxiliary Feedwater in the secondary system. In addition to this goal the presented work is aimed to resolve the potential safety issue resulting from the influence of the Chemical and Volume Control System (CVCS) operation during LOOP event for NPP Krsko. The potential safety concern for the LOOP analysis is that the loss of instrument air system may occur thus leading to the CVCS charging and letdown flow imbalance. A net RCS inventory addition may result with water solid pressurizer condition. Water discharge through the pressurizer relief and safety valves could lead to overpressurization of the Pressurizer Relief Tank (PRT) and rupture of the PRT rupture disks. Additional concern is that pressurizer relief and safety valves may fail to properly reseat when exposed to water relief

ORTAP: a nuclear steam supply system simulation for the dynamic analysis of high temperature gas cooled reactor transients

International Nuclear Information System (INIS)

Cleveland, J.C.; Hedrick, R.A.; Ball, S.J.; Delene, J.G.

1977-01-01

ORTAP was developed to predict the dynamic behavior of the high temperature gas cooled reactor (HTGR) Nuclear Steam Supply System for normal operational transients and postulated accident conditions. It was developed for the Nuclear Regulatory Commission (NRC) as an independent means of obtaining conservative predictions of the transient response of HTGRs over a wide range of conditions. The approach has been to build sufficient detail into the component models so that the coupling between the primary and secondary systems can be accurately represented and so that transients which cover a wide range of conditions can be simulated. System components which are modeled in ORTAP include the reactor core, a typical reheater and steam generator module, a typical helium circulator and circulator turbine and the turbine generator plant. The major plant control systems are also modeled. Normal operational transients which can be analyzed with ORTAP include reactor start-up and shutdown, normal and rapid load changes. Upset transients which can be analyzed with ORTAP include reactor trip, turbine trip and sudden reduction in feedwater flow. ORTAP has also been used to predict plant response to emergency or faulted conditions such as primary system depressurization, loss of primary coolant flow and uncontrolled removal of control poison from the reactor core

Teachers as Secondary Players: Involvement in Field Trips to Natural Environments

Science.gov (United States)

Alon, Nirit Lavie; Tal, Tali

2017-08-01

This study focused on field trips to natural environments where the teacher plays a secondary role alongside a professional guide. We investigated teachers' and field trip guides' views of the teacher's role, the teacher's actual function on the field trip, and the relationship between them. We observed field trips, interviewed teachers and guides, and administered questionnaires. We found different levels of teacher involvement, ranging from mainly supervising and giving technical help, to high involvement especially in the cognitive domain and sometimes in the social domain. Analysis of students' self-reported outcomes showed that the more students believe their teachers are involved, the higher the self-reported learning outcomes.

Make My Trip Count 2015

Data.gov (United States)

Allegheny County / City of Pittsburgh / Western PA Regional Data Center — The Make My Trip Count (MMTC) commuter survey, conducted in September and October 2015 by GBA, the Pittsburgh 2030 District, and 10 other regional transportation...

Flat Plate Boundary Layer Stimulation Using Trip Wires and Hama Strips

Science.gov (United States)

Peguero, Charles; Henoch, Charles; Hrubes, James; Fredette, Albert; Roberts, Raymond; Huyer, Stephen

2017-11-01

Water tunnel experiments on a flat plate at zero angle of attack were performed to investigate the effect of single roughness elements, i.e., trip wires and Hama strips, on the transition to turbulence. Boundary layer trips are traditionally used in scale model testing to force a boundary layer to transition from laminar to turbulent flow at a single location to aid in scaling of flow characteristics. Several investigations of trip wire effects exist in the literature, but there is a dearth of information regarding the influence of Hama strips on the flat plate boundary layer. The intent of this investigation is to better understand the effects of boundary layer trips, particularly Hama strips, and to investigate the pressure-induced drag of both styles of boundary layer trips. Untripped and tripped boundary layers along a flat plate at a range of flow speeds were characterized with multiple diagnostic measurements in the NUWC/Newport 12-inch water tunnel. A wide range of Hama strip and wire trip thicknesses were used. Measurements included dye flow visualization, direct skin friction and parasitic drag force, boundary layer profiles using LDV, wall shear stress fluctuations using hot film anemometry, and streamwise pressure gradients. Test results will be compared to the CFD and boundary layer model results as well as the existing body of work. Conclusions, resulting in guidance for application of Hama strips in model scale experiments and non-dimensional predictions of pressure drag will be presented.

Performance Monitoring for Nuclear Safety Related Instrumentation at PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2015-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on performance monitoring for nuclear safety related instrumentation in TRIGA PUSPATI Reactor (RTP) of based on various parameter of reactor safety instrument channel such as log power, linear power, Fuel temperature, coolant temperature will take into consideration. Methodology of performance on estimation and monitoring is to evaluate and analysis of reactor parameters which is important of reactor safety and control. And also to estimate power measurement, differential of log and linear power and fuel temperature during reactor start-up, operation and shutdown .This study also focus on neutron power fluctuation from fission chamber during reactor start-up and operation. This work will present result of performance monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that performance of nuclear safety related instrumentation will improved the reactor control and safety parameter during reactor start-up, operation and shutdown. (author)

Recommendations for Planning and Managing International Short-term Pharmacy Service Trips.

Science.gov (United States)

Johnson, Kalin L; Alsharif, Naser Z; Rovers, John; Connor, Sharon; White, Nicole D; Hogue, Michael D

2017-03-25

International pharmacy service trips by schools and colleges of pharmacy allow students to provide health care to medically underserved areas. A literature review (2000-2016) in databases and Internet searches with specific keywords or terms was performed to assess current practices to establish and maintain successful pharmacy service trips. Educational documents such as syllabi were obtained from pharmacy programs and examined. A preliminary draft was developed and authors worked on sections of interest and expertise. Considerations and current recommendations are provided for the key aspects of the home institution and the host country requirements for pharmacy service trips based on findings from a literature search and the authors' collective, extensive experience. Evaluation of the trip and ethical considerations are also discussed. This article serves as a resource for schools and colleges of pharmacy that are interested in the development of new pharmacy service trips and provides key considerations for continuous quality improvement of current or future activities.

Analysis of the events on the operating of the wrong compartment of NPPs

International Nuclear Information System (INIS)

Zheng Lixin; Zhou Hong; Zhang Hao; Che Shuwei; Zhang Jiajun

2013-01-01

In this paper, an operational event that unit trip caused by the operating of the wrong compartment, due to the personnel error is introduced. Through in-depth research on this kind of events the causes of the events are found, some suggestions are put forward. It can provide a reference for preventing the similar events from recurring to other NPPs. (authors)

Trip-oriented travel time prediction (TOTTP) with historical vehicle trajectories

Science.gov (United States)

Xu, Tao; Li, Xiang; Claramunt, Christophe

2018-06-01

Accurate travel time prediction is undoubtedly of importance to both traffic managers and travelers. In highly-urbanized areas, trip-oriented travel time prediction (TOTTP) is valuable to travelers rather than traffic managers as the former usually expect to know the travel time of a trip which may cross over multiple road sections. There are two obstacles to the development of TOTTP, including traffic complexity and traffic data coverage.With large scale historical vehicle trajectory data and meteorology data, this research develops a BPNN-based approach through integrating multiple factors affecting trip travel time into a BPNN model to predict trip-oriented travel time for OD pairs in urban network. Results of experiments demonstrate that it helps discover the dominate trends of travel time changes daily and weekly, and the impact of weather conditions is non-trivial.

Reactor safety method

International Nuclear Information System (INIS)

Vachon, L.J.

1980-01-01

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature

Safety analysis of the critical facility for AHWR and 500 MWe PHWR

International Nuclear Information System (INIS)

Pushpam, Neelima Prasad; Arvind Kumar; Srivenkatesan, R.

2002-01-01

Full text: The initiating event for the design basis reactivity accident is the uncontrolled moderator pump up at criticality. This uncontrolled pump up transient is considered to be the enveloping scenario and has been analysed for the reference core with AHWR fuel using point kinetics model. It is the most reactive core among the three with a small b value (core average b= 5.96 mk). The maximum pump rate in critical facility is limited to 300 litres/min which corresponds to the maximum rate of reactivity addition about 0.1 mk/sec. On any reactor trip 6 shut-off rods are inserted into the core along with partial moderator dumps. The reactor is provided with independent safety and regulating channels (SC and RC) to monitor reactor neutronic power and initiate trip at different power levels. After the reactor trips five of the six fast acting shut-off rods (maximum worth rod is unavailable) fall under gravity and at the same time moderator dump is initiated. We have considered shut off rods and moderator dump as two independent shutdown systems. The analysis shows that even if the reactor trips at the high power at 550 watt ignoring the earlier trips, the fuel temperature does not rise beyond 50 degC and the total energy released is less than 20 kW. We also analysed the transients due to uncontrolled withdrawal of absorber rod. In this case also we found that the fuel temperature became ∼54 degC and the total energy release was about 25 kW. The fuel can withstand this temperature. This shows that reactor is safe

Identification of Initiating Events for PGSFR

International Nuclear Information System (INIS)

Kim, Jintae; Jae, Moosung

2016-01-01

The Sodium-cooled Fast Reactor (SFR) is by far the most advanced reactor of the six Generation IV reactors. The SFR uses liquid sodium as the reactor coolant, which has superior heat transport characteristics. It also allows high power density with low coolant volume fraction and operation at low pressure. In Korea, KAERI has been developing Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) that employs passive safety systems and inherent reactivity feedback effects. In order to prepare for the licensing, it is necessary to assess the safety of the reactor. Thus, the objective of this study is to conduct accident sequence analysis that can contribute to risk assessment. The analysis embraces identification of initiating events and accident sequences development. PGSFR is to test and demonstrate the performance of transuranic (TRU)-containing metal fuel required for a commercial SFR, and to demonstrate the TRU transmutation capability of a burner reactor as a part of an advanced fuel cycle system. Initiating events that can happen in PGSFR were identified through the MLD method. This method presents a model of a plant in terms of individual events and their combinations in a systematic and logical way. The 11 identified initiating events in this study include the events considered in the past analysis that was conducted for PRISM-150

Identification of Initiating Events for PGSFR

Energy Technology Data Exchange (ETDEWEB)

Kim, Jintae; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

2016-10-15

The Sodium-cooled Fast Reactor (SFR) is by far the most advanced reactor of the six Generation IV reactors. The SFR uses liquid sodium as the reactor coolant, which has superior heat transport characteristics. It also allows high power density with low coolant volume fraction and operation at low pressure. In Korea, KAERI has been developing Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) that employs passive safety systems and inherent reactivity feedback effects. In order to prepare for the licensing, it is necessary to assess the safety of the reactor. Thus, the objective of this study is to conduct accident sequence analysis that can contribute to risk assessment. The analysis embraces identification of initiating events and accident sequences development. PGSFR is to test and demonstrate the performance of transuranic (TRU)-containing metal fuel required for a commercial SFR, and to demonstrate the TRU transmutation capability of a burner reactor as a part of an advanced fuel cycle system. Initiating events that can happen in PGSFR were identified through the MLD method. This method presents a model of a plant in terms of individual events and their combinations in a systematic and logical way. The 11 identified initiating events in this study include the events considered in the past analysis that was conducted for PRISM-150.

Thermal-hydraulic transient characteristics of ship-propulsion reactor investigated through safety analysis

International Nuclear Information System (INIS)

Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

1986-01-01

Thermal-hydraulic behaviors in the reactor of Nuclear Ship ''Mutsu'' were investigated through safety evaluation of operational transients by using RETRAN and COBRA-IV codes. The results were compared to the transient behaviors of typical commercial PWR and the characteristics of transient thermal-hydraulic behaviors in ship-loaded reactor were figured out. ''Mutsu'' reactor has larger thermal margin than commercial PWR because it is designed to be used as ship-propulsion power source in the load-following operation mode. This margin makes transient behavior in general milder than in commercial PWR but high opening pressure set point of main-steam safety valves leads poor heat-sink condition after reactor trip. The effects of other small-sized components are also investigated. The findings in the paper will be helpful in the design of future advanced reactor for nuclear ship. (author)

Addressing legal and political barriers to global pharmaceutical access: options for remedying the impact of the Agreement on Trade-Related Aspects of Intellectual Property Rights (TRIPS) and the imposition of TRIPS-plus standards.

Science.gov (United States)

Cohen-Kohler, Jillian Clare; Forman, Lisa; Lipkus, Nathaniel

2008-07-01

Despite myriad programs aimed at increasing access to essential medicines in the developing world, the global drug gap persists. This paper focuses on the major legal and political constraints preventing implementation of coordinated global policy solutions - particularly, the Agreement on Trade-Related Aspects of Intellectual Property Rights (TRIPS) and bilateral and regional free trade agreements. We argue that several policy and research routes should be taken to mitigate the restrictive impact of TRIPS and TRIPS-plus rules, including greater use of TRIPS flexibilities, advancement of human rights, and an ethical framework for essential medicines distribution, and a broader campaign that debates the legitimacy of TRIPS and TRIPS-plus standards themselves.

Evaluation to Mitigate Secondary System Peak Pressure for Loss of Condenser Vacuum Event

Energy Technology Data Exchange (ETDEWEB)

Choi, Bong Oh; Park, Jong Cheol; Park, Min Soo; Lee, Gyu Cheon; Kim, Shin Whan [KEPCO E and C, Inc., Daejeon (Korea, Republic of)

2015-10-15

In this paper, countermeasures to compensate the increased secondary pressure are introduced and evaluated. From the standpoint of the secondary system pressurization, consideration of the PPCS may result in a conservative secondary system peak pressure. The control systems are generally credited for the safety analysis if the analysis produces conservative results. However, in most of all non-loss of coolant accident (non-LOCA) events, the control system helps to mitigate a transient state. Accordingly, the safety analysis of non-LOCA assumes the control systems are in the manual mode of operation. The loss of condenser vacuum event (LOCV) is a typical anticipated operational occurrence (AOO) which results in an increase in primary and secondary system pressure. The pressurizer (PZR) pressure control system (PPCS) will function to reduce the primary system pressure increase during the transient. Therefore, it is assumed to be in manual mode and credit is not taken for its functioning. However, crediting the function of PPCS has been found to be more conservative with regard to the secondary system pressure. This is due to the delay of the reactor trip on high pressurizer pressure (HPP) and results in an increase in secondary pressure.

Nuclear reactor power supply system

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector prevents a parameter signal which differs from the other parameter signals of the set by more than twice the allowable variation from passing to the control system. Test signals are periodically impressed by a test unit on a selected pair of a selection unit and control channels. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test. (author)

Savannah River Site reactor hardware design modification study

International Nuclear Information System (INIS)

Fisher, J.E.

1990-03-01

A study was undertaken to assess the merits of proposed design modifications to the SRS reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. For the elevated piping design, system recovery was predicted for breaks in the plenum inlet or pump suction piping; response to the pump discharge break location did not show improvement compared to the present system configuration. The rotovalve closure design improved system response to plenum inlet or pump discharge breaks; recovery was not predicted for pump suction breaks. The pump suction valve closure design demonstrated system recovery for all break locations downstream of the valve. A combination of features is recommended to ensure liquid inventory recovery for all break locations. The elevated piping design performance during pump discharge breaks would be improved with addition of a dc pump trip in the affected loop. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 12 refs., 10 figs., 2 tabs

Trip time prediction in mass transit companies. A machine learning approach

OpenAIRE

João M. Moreira; Alípio Jorge; Jorge Freire de Sousa; Carlos Soares

2005-01-01

In this paper we discuss how trip time prediction can be useful foroperational optimization in mass transit companies and which machine learningtechniques can be used to improve results. Firstly, we analyze which departmentsneed trip time prediction and when. Secondly, we review related work and thirdlywe present the analysis of trip time over a particular path. We proceed by presentingexperimental results conducted on real data with the forecasting techniques wefound most adequate, and concl...

Type B investigation of the iridium contamination event at the High Flux Isotope Reactor on September 7, 1993

International Nuclear Information System (INIS)

1994-03-01

On the title date, at ORNL, area radiation alarms sounded during a routine transfer of a shielding cask (containing 60 Ci 192 Ir) from the HFIR pool side to a transport truck. Small amounts of Ir were released from the cask onto the reactor bay floor. The floor was cleaned, and the cask was shipped to a hot cell at Building 3047 on Oct. 3, 1993. The event was caused by rupture of one of the Ir target rods after it was loaded into the cask for normal transport operations; the rupture was the result of steam generation in the target rod soon after it was placed in the cask (water had entered the target rod through a tiny defect in a weld while it was in the reactor under pressure). While the target rods were in the reactor and reactor pool, there was sufficient cooling to prevent steam generation; when the target rod was loaded into the dry transport cask, the temperature increased enough to result in boiling of the trapped water and produced high enough pressure to result in rupture. The escaping steam ejected some of the Ir pellets. The event was reported as Occurrence Report Number ORO--MMES-X10HFIR-1993-0030, dated Sept. 8, 1993. Analysis indicated that the following conditions were probable causes: less than adequate welding procedures, practices, or techniques, material controls, or inspection methods, or combination thereof, could have led to weld defects, affecting the integrity of target rod IR-75; less than adequate secondary containment in the cask allowed Ir pellets to escape

Determination of Biology Department Students' Past Field Trip Experiences and Examination of Their Self-Efficacy Beliefs in Planning and Organising Educational Field Trips

Science.gov (United States)

Bozdogan, Aykut Emre

2015-01-01

The purpose of this study is to determine the past field trip experiences of pre-service teachers who are graduates of Faculty of Sciences, Department of Biology and who had pedagogical formation training certificate and to examine their self-efficacy beliefs in planning and organizing field trips with regard to different variables. The study was…

The Development and Evaluation of Inherent RPCS for the APR1400

Energy Technology Data Exchange (ETDEWEB)

Choi, Jae D.; Ryu, Seok H.; Jung, Dong C.; Kim, Joon S.; Baek, Byung C.; Sung, Song K.; You, Guk J. [KNF, Daejeon (Korea, Republic of); Kim, Han G. [KHNP, Daejeon (Korea, Republic of); Chi, Sung G. [KOPEC, Daejeon (Korea, Republic of)

2008-10-15

The APR1400 RPCS (Reactor Power Cutback System) is designed to rapidly reduce the core power to eliminate the need for a reactor trip following a large load rejection or a loss of two main feedwater pumps at high power. GDC (General Design Criteria) 25 says 'Protection system requirements for reactivity control malfunctions. The protection system shall be designed to assure that SAFDL (Specified Acceptable Fuel Design Limits) are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.' In order to comply GDC 25, CPCS (Core Protection Calculator System) will apply a big power penalty({approx} 1.3) to determine the minimum DNBR (Departure from Nucleate Boiling Ratio), the maximum LPD(Local Power Density) and reactor may immediately trip by CPCS low DNBR and high LPD during high power operation for 12-finger CEA drop. The purpose of this study is to develop the inherent RPCS to avoid unwanted reactor trips due to a 12-finger CEA drop. In order to accomplish this purpose, the CPCS should be modified to send RPCS actuation signal and not to apply power penalty due to 12-finger CEA drop for a shot period. During this period which is determined by assessment of safety, the SAFDL will not be violated without any CPCS trip function. The system and CPCS performance is evaluated to verify that CPCS trip does not occurred during 12-finger CEA drop event.

Lessons learned from the SONGS Unit 1 water hammer event

International Nuclear Information System (INIS)

Chiu, C.

1987-01-01

On November 21, 1985, a water hammer event occurred in horizontal feedwater line B at San Onofre Nuclear Generation Site (SONGS) Unit 1. The SONGS Unit 1 is a three-loop pressurized water reactor designed by Westinghouse Electric Corp. The event was initiated by a differential current trip on the bus of auxiliary transformer C. The root cause of the event was a simultaneous failure of five check valves in the feedwater system. Two of them are located downstream of the feedwater pump, and three of them are located further downstream and on the lines to the steam generators. The failure mechanism was determined to be flow-induced vibration, which caused repeated impact between the disk stud and the disk stop. The water hammer occurred in feedwater line B during the refilling of feedwater lines A, B, and C with auxiliary feedwater. The thermal-hydraulic process to initiate the water hammer and the reason that the water hammer only happened in line B have been fully investigated and explained. A root cause analysis after the event was prompted to answer the following two questions: (1) why did these five check valves fail at that time and not in the preceding 15 yr? (2) why did only these five check valves fail? The scope of the root cause analysis involves an investigation of the valve vibration characteristics, plant operation history, and the maintenance history of the valves. The paper answers these two questions, after a brief study of the vibration characteristics of a check valve

Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

International Nuclear Information System (INIS)

Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

2003-01-01

The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

Analysis of the accident at Fukushima Daiichi nuclear power plant in an A BWR reactor

International Nuclear Information System (INIS)

Escorcia O, D.; Salazar S, E.

2016-09-01

The present work aims to recreate the accident occurred at the Fukushima Daiichi nuclear power plant in Japan on March 11, 2011, making use of an academic simulator of forced circulation of the A BWR reactor provided by the IAEA to know the scope of this simulator. The simulator was developed and distributed by the IAEA for academic purposes and contains the characteristics and general elements of this reactor to be able to simulate transients and failures of different types, allowing also to observe the general behavior of the reactor, as well as several phenomena and present systems in the same. Is an educational tool of great value, but it does not have a scope that allows the training of plant operators. To recreate the conditions of the Fukushima accident in the simulator, we first have to know what events led to this accident, as well as the actions taken by operators and managers to reduce the consequences of this accident; and the sequence of events that occurred during the course of the accident. Differences in the nuclear power plant behavior are observed and interpreted throughout the simulation, since the Fukushima plant technology and the simulator technology are not the same, although they have several elements in common. The Fukushima plant had an event that by far exceeded the design basis, which triggered in an accident that occurred in the first place by a total loss of power supply, followed by the loss of cooling systems, causing a level too high in temperature, melting the core and damaging the containment accordingly, allowing the escape of hydrogen and radioactive material. As a result of the simulation, was determined that the scope of the IAEA academic simulator reaches the entrance of the emergency equipment, so is able to simulate almost all the events occurred at the time of the earthquake and the arrival of the tsunami in the nuclear power plant of Fukushima Daiichi. However, due to its characteristics, is not able to simulate later

Reactivity changes in hybrid thermal-fast reactor systems during fast core flooding

International Nuclear Information System (INIS)

Pesic, M.

1994-09-01

A new space-dependent kinetic model in adiabatic approximation with local feedback reactivity parameters for reactivity determination in the coupled systems is proposed in this thesis. It is applied in the accident calculation of the 'HERBE' fast-thermal reactor system and compared to usual point kinetics model with core-averaged parameters. Advantages of the new model - more realistic picture of the reactor kinetics and dynamics during local large reactivity perturbation, under the same heat transfer conditions, are underlined. Calculated reactivity parameters of the new model are verified in the experiments performed at the 'HERBE' coupled core. The model has shown that the 'HERBE' safety system can shutdown reactor safely and fast even in the case of highly set power trip and even under conditions of big partial failure of the reactor safety system (author)

An abnormal event advisory expert system prototype for reactor operators

International Nuclear Information System (INIS)

Hance, D.C.

1989-01-01

Nuclear plant operators must respond correctly during abnormal conditions in the presence of dynamic and potentially overwhelming volumes of information. For this reason, considerable effort has been directed toward the development of nuclear plant operator aids using artificial intelligence techniques. The objective of such systems is to diagnose abnormal conditions within the plant, possibly predict consequences, and advise the operators of corrective actions in a timely manner. The objective of the work is the development of a prototype expert system to diagnose abnormal events at a nuclear power plant and advise plant operators of the event and applicable procedures in an on-line mode. The major difference between this effort and previous work is the use of plant operating procedures as a knowledge source and as an integral part of the advice provided by the expert system. The acceptance by utilities of expert systems as operator aids requires that such systems be compatible with the regulatory environment and provide economic benefits. For this reason, commercially viable operator aid systems developed in the near future must complement existing plant procedures rather than reach beyond them in a revolutionary manner. A knowledge source is the resource providing facts and relationships that are coded into the expert system program. In this case, the primary source of knowledge is a set of selected abnormal operating procedures for a modern Westinghouse pressurized water reactor

Does ignoring multidestination trips in the travel cost method cause a systematic bias?

NARCIS (Netherlands)

Kuosmanen, T.K.; Nillesen, E.E.M.; Wesseler, J.H.H.

2004-01-01

The present paper demonstrates that treating multidestination trips (MDT) as single-destination trips does not involve any systematic upward or downward bias in consumer surplus (CS) estimates because the direct negative effect of a price increase (treating MDT as a single-destination trip) is

Recurring events, and the Possible Need to Reinforce Operating Experience Feedback Programs

International Nuclear Information System (INIS)

Ross, Denwood

1999-09-01

A nuclear power plant is designed for a spectrum of incidents and accidents, ranging from a reactor trip without other complications to more serious events such as pipe ruptures. Certain portions of the plant are designed for even more significant events such as severe accidents. Several thousand reactor years of experience have been recorded and many postulated events have in fact occurred. In some instances the same or similar event has occurred more than once within a single country or among several nations. Such cases are referred to as recurring events. One way to reduce the likelihood, or severity (or both) of recurrence is to maintain and utilize a system for reporting of events, both at the national and the international levels. The international system is referred as the Incident Reporting System. Events to be reported to IRS include: - The event itself is serious or important in terms of safety due to an actual or potential reduction in the plant's defense in depth; - The event reveals important lessons learned that will help the international community to prevent its recurrence as a safety significant event under aggravated conditions or to avoid the occurrence of a serious or important event in terms of safety; - The event is a repetition of a similar event previously reported to IRS, but highlights new important lessons learned for the international community. National systems for reporting of events vary in scope; there is guidance on systems for feedback of experience from events in nuclear power plants. Further, the Nuclear Safety Convention, Article 19 - Operation - provides (section vii) that each Contracting Party shall take the appropriate steps to ensure that 'programmes to collect and analyse operating experience are established, the results obtained and the conclusions drawn are acted upon and that existing mechanisms are used to share important experience with international bodies and with other operating organizations and regulatory bodies

Reactor power control device

International Nuclear Information System (INIS)

Imaruoka, Hiromitsu.

1994-01-01

A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

Safety philosophy in upgrading the EBR-II plant protection system

International Nuclear Information System (INIS)

Sackett, J.I.

1976-01-01

The EBR-II plant protection system (PPS) has been substantially modified, upgrading its performance to more fully comply with modern safety philosophy and criteria. The upgrading effort required that the total reactor system be evaluated for possible faults and that a PPS be designed to accommodate them. The result was deletion of a number of existing trip functions and upgrading of others. Particular attention was given to loss of primary pumping power and reactivity insertion events. The design and performance criteria for the PPS has been more firmly established, understanding of the PPS function has been improved and the reactor has been subjected to fewer spurious trips, improving operational reliability

Response to severe changes of load on the reactor system of nuclear ship Mutsu

International Nuclear Information System (INIS)

Ishida, Toshihisa; Kusunoki, Tsuyoshi; Ochiai, Masa-aki; Tanaka, Yoshimi; Yao, Toshiaki; Inoue, Kimio.

1993-01-01

The response of the nuclear power system of N.S. Mutsu to severe changes of load have been studied from records taken during the power-raising tests performed on the ship in 1990. The records examined were those involving the most severe load changes foreseen for marine reactors: (a) sharp load increase with total steam flow raised from 25 to 70 % rated full flow in 13s, (b) crash astern maneuver with the position of propulsion turbine command handle changed-taking several seconds-from cruising ahead to STOP, and after about 50s, further changed-taking 30s-to bring the astern propulsion turbine to full speed-to consume approximately 60 % rated total steam flow and (c) turbine trip with the ahead turbine intentionally tripped when operating at roughly 100 % rated total steam flow. The foregoing records from load changes-of severity beyond what is foreseen for land-based reactors-proved that the Mutsu reactor is capable of responding smoothly and securely to such severe load changes. These load changes occasioned relatively large mismatches between reactor power supply and steam flow demand, but with notable freedom from any conspicuous overshooting or hunting of the reactor power. This performance can be attributed to (a) correct functioning of the automatic power control system, (b) effective contribution of the self-regulating reactor control property deriving from the large negative feedback between moderator temperature and reactivity, and (c) the ample inventories of coolant in the primary and secondary loops. The responses to load change are discussed covering those relevant to (a) reactor power, (b) primary loop pressure, and (c) steam generator pressure, with particular reference to the differences seen in response to mild and to severe load changes. (author)

The AAA+ ATPase TRIP13 remodels HORMA domains through N-terminal engagement and unfolding

Energy Technology Data Exchange (ETDEWEB)

Ye, Qiaozhen; Kim, Dong Hyun; Dereli, Ihsan; Rosenberg, Scott C.; Hagemann, Goetz; Herzog, Franz; Tóth, Attila; Cleveland, Don W.; Corbett, Kevin D.

2017-06-28

Proteins of the conserved HORMA domain family, including the spindle assembly checkpoint protein MAD2 and the meiotic HORMADs, assemble into signaling complexes by binding short peptides termed “closure motifs”. The AAA+ ATPase TRIP13 regulates both MAD2 and meiotic HORMADs by disassembling these HORMA domain–closure motif complexes, but its mechanisms of substrate recognition and remodeling are unknown. Here, we combine X-ray crystallography and crosslinking mass spectrometry to outline how TRIP13 recognizes MAD2 with the help of the adapter protein p31comet. We show that p31comet binding to the TRIP13 N-terminal domain positions the disordered MAD2 N-terminus for engagement by the TRIP13 “pore loops”, which then unfold MAD2 in the presence of ATP. N-terminal truncation of MAD2 renders it refractory to TRIP13 action in vitro, and in cells causes spindle assembly checkpoint defects consistent with loss of TRIP13 function. Similar truncation of HORMAD1 in mouse spermatocytes compromises its TRIP13-mediated removal from meiotic chromosomes, highlighting a conserved mechanism for recognition and disassembly of HORMA domain–closure motif complexes by TRIP13.

Evaluation of postulated LOF [loss-of-flow] events in PRISM and SAFR

International Nuclear Information System (INIS)

Chan, B.C.; Van Tuyle, G.J.; Slovik, G.C.; Aronson, A.L.

1987-01-01

The PRISM and SAFR designs, as currently proposed by DOE, are designed for ''inherent'', as opposed to ''engineered'', safety. Brookhaven National Laboratory is supporting the initial NRC review of these advanced LMR concepts. A loss-of-flow (LOF) accident coupled with a failure of the reactor shutdown system is one of the major safety concerns in the advanced liquid metal reactor (LMR) evaluation effort. The analysis discussed here covers: (1) primary pipe break without pump trip, (2) primary coolant pump seizure, and (3) primary coolant pump coastdown. The analytical modelling and the calculated thermal and hydraulic behavior are described in detail

TRIP-Br2 promotes oncogenesis in nude mice and is frequently overexpressed in multiple human tumors.

Science.gov (United States)

Cheong, Jit Kong; Gunaratnam, Lakshman; Zang, Zhi Jiang; Yang, Christopher M; Sun, Xiaoming; Nasr, Susan L; Sim, Khe Guan; Peh, Bee Keow; Rashid, Suhaimi Bin Abdul; Bonventre, Joseph V; Salto-Tellez, Manuel; Hsu, Stephen I

2009-01-20

Members of the TRIP-Br/SERTAD family of mammalian transcriptional coregulators have recently been implicated in E2F-mediated cell cycle progression and tumorigenesis. We, herein, focus on the detailed functional characterization of the least understood member of the TRIP-Br/SERTAD protein family, TRIP-Br2 (SERTAD2). Oncogenic potential of TRIP-Br2 was demonstrated by (1) inoculation of NIH3T3 fibroblasts, which were engineered to stably overexpress ectopic TRIP-Br2, into athymic nude mice for tumor induction and (2) comprehensive immunohistochemical high-throughput screening of TRIP-Br2 protein expression in multiple human tumor cell lines and human tumor tissue microarrays (TMAs). Clinicopathologic analysis was conducted to assess the potential of TRIP-Br2 as a novel prognostic marker of human cancer. RNA interference of TRIP-Br2 expression in HCT-116 colorectal carcinoma cells was performed to determine the potential of TRIP-Br2 as a novel chemotherapeutic drug target. Overexpression of TRIP-Br2 is sufficient to transform murine fibroblasts and promotes tumorigenesis in nude mice. The transformed phenotype is characterized by deregulation of the E2F/DP-transcriptional pathway through upregulation of the key E2F-responsive genes CYCLIN E, CYCLIN A2, CDC6 and DHFR. TRIP-Br2 is frequently overexpressed in both cancer cell lines and multiple human tumors. Clinicopathologic correlation indicates that overexpression of TRIP-Br2 in hepatocellular carcinoma is associated with a worse clinical outcome by Kaplan-Meier survival analysis. Small interfering RNA-mediated (siRNA) knockdown of TRIP-Br2 was sufficient to inhibit cell-autonomous growth of HCT-116 cells in vitro. This study identifies TRIP-Br2 as a bona-fide protooncogene and supports the potential for TRIP-Br2 as a novel prognostic marker and a chemotherapeutic drug target in human cancer.

Medical and pharmacy student concerns about participating on international service-learning trips

OpenAIRE

Chuang, Chih; Khatri, Siddique H.; Gill, Manpal S.; Trehan, Naveen; Masineni, Silpa; Chikkam, Vineela; Farah, Guillaume G.; Khan, Amber; Levine, Diane L.

2015-01-01

Background International Service Learning Trips (ISLT) provide health professional students the opportunity to provide healthcare, under the direction of trained faculty, to underserved populations in developing countries. Despite recent increases in international service learning trips, there is scant literature addressing concerns students have prior to attending such trips. This study focuses on identifying concerns before and after attending an ISLT and their impact on students. Methods A...

Methods for analysis of passenger trip performance in a complex networked transportation system

Science.gov (United States)

Wang, Danyi

2007-12-01

The purpose of the Air Transportation System (ATS) is to provide safe and efficient transportation service of passengers and cargo. The on-time performance of a passenger's trip is a critical performance measurement of the Quality of Service (QOS) provided by any Air Transportation System. QOS has been correlated with airline profitability, productivity, customer loyalty and customer satisfaction (Heskett et al. 1994). Btatu and Barnhart have shown that official government and airline on-time performance metrics (i.e. flight-centric measures of air transportation) fail to accurately reflect the passenger experience (Btatu and Barnhart, 2005). Flight-based metrics do not include the trip delays accrued by passengers who were re-booked due to cancelled flights or missed connections. Also, flight-based metrics do not quantify the magnitude of the delay (only the likelihood) and thus fails to provide the consumer with a useful assessment of the impact of a delay. Passenger-centric metrics have not been developed because of the unavailability of airline proprietary data, which is also protected by anti-trust collusion concerns and civil liberty privacy restrictions. Moveover, the growth of the ATS is trending out of the historical range. The objectives of this research were to (1) estimate ATS-wide passenger trip delay using publicly accessible flight data, and (2) investigate passenger trip dynamics out of the range of historical data by building a passenger flow simulation model to predict impact on passenger trip time given anticipated changes in the future. The first objective enables researchers to conduct historical analysis on passenger on-time performance without proprietary itinerary data, and the second objective enables researchers to conduct experiments outside the range of historic data. The estimated passenger trip delay was for 1,030 routes between the 35 busiest airports in the United States in 2006. The major findings of this research are listed as

The analysis with the code TANK of a postulated reactivity-insertion transient in a 10-MW MAPLE research reactor

International Nuclear Information System (INIS)

Ellis, R.J.

1990-10-01

This report discusses the analysis of a postulated loss-of-regulation (LOR) accident in a metal-fuelled MAPLE Research Reactor. The selected transient scenario involves a slow LOR from low reactor power; the control rods are assumed to withdraw slowly until a trip at 12 MW halts the withdrawal. The simulation was performed using the space-time reactor kinetics computer code TANK, and modelling the reactor in detail in two dimensions and in two neutron-energy groups. Emphasis in this report is placed on the modelling techniques used in TANK and the physics considerations of the analysis

Development of ABWR inertia-increased reactor internal pump and thicker sleeve nozzle

International Nuclear Information System (INIS)

Takahashi, Shirou; Shiina, Kouji; Matsumura, Seiichi

2002-01-01

The conventional reactor internal pumps (RIPs) in the ABWR have an inertia moment coming from the shafts and Motor-Generator sets, enabling the RIPs to continue running for a few seconds, when a trip of all RIPs event occurs. It is possible to simplify the RIPs' power supply system without affecting the core flow supply when the above event occurs by eliminating M-G sets, if the rotating inertia is increased. This inertia increase due to an additional flywheel, which leads to gains in weight and length, requires the larger diameter nozzle with the thicker sleeve. However, too large a nozzle diameter may change the hydraulic performance. In authors' previous study, the optimum nozzle diameter (492 mm) was selected through 1/5-scale test. In this study, the 492 mm nozzle and the inertia-increased RIP were verified through the full-scale tests. The rotating inertia time constant on coastdown characteristics (behavior of the RIP speed in the event of power loss) for the inertia-increased RIP doubled compared with the current RIP. The casing and the shaft vibration were also confirmed to satisfy the design criteria. Moreover, hydraulic performance and heat increase in the motor casing due to the flywheel were evaluated. The inertia increased RIP with the 492 mm nozzle maintained good performance. (author)

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

Predictors of trips to food destinations

Directory of Open Access Journals (Sweden)

Kerr Jacqueline

2012-05-01

Full Text Available Abstract Background Food environment studies have focused on ethnic and income disparities in food access. Few studies have investigated distance travelled for food and did not aim to inform the geographic scales at which to study the relationship between food environments and obesity. Further, studies have not considered neighborhood design as a predictor of food purchasing behavior. Methods Atlanta residents (N = 4800 who completed a travel diary and reported purchasing or consuming food at one of five food locations were included in the analyses. A total of 11,995 food-related trips were reported. Using mixed modeling to adjust for clustering of trips by participants and households, person-level variables (e.g. demographics, neighborhood-level urban form measures, created in GIS, and trip characteristics (e.g. time of day, origin and destination were investigated as correlates of distance travelled for food and frequency of grocery store and fast food outlet trips. Results Mean travel distance for food ranged from 4.5 miles for coffee shops to 6.3 miles for superstores. Type of store, urban form, type of tour, day of the week and ethnicity were all significantly related to distance travelled for food. Origin and destination environment, type of tour, day of week, age, gender, income, ethnicity, vehicle access and obesity status were all significantly related to visiting a grocery store. Home neighborhood environment, day of week, type of tour, gender, income, education level, age, and obesity status were all significantly related to likelihood of visiting a fastfood outlet. Conclusions The present study demonstrated that people travel sizeable distances for food and this distance is related to urban. Results suggest that researchers need to employ different methods to characterize food environments than have been used to assess urban form in studies of physical activity. Food is most often purchased while traveling from locations other

Simulation and verification of transient events in large wind power installations

Energy Technology Data Exchange (ETDEWEB)

Soerensen, P.; Hansen, A.D.; Christensen, P.; Meritz, M.; Bech, J.; Bak-Jensen, B.; Nielsen, H.

2003-10-01

Models for wind power installations excited by transient events have been developed and verified. A number of cases have been investigated, including comparisons of simulations of a three-phase short circuit, validation with measurements of tripping of single wind turbine, islanding of a group of two wind turbines, and voltage steps caused by tripping of wind turbines and by manual transformer tap-changing. A Benchmark model is also presented, enabling the reader to test own simulation results against results obtained with models developed in EMTDC and DIgSILENT. (au)

Field Trips and the Law.

Science.gov (United States)

Troy, Thomas D.; Schwaab, Karl E.

1981-01-01

Legal aspects of field trips are addressed, with special attention on planning and implementation aspects which warrant legal consideration. Suggestions are based on information obtained from studies which reviewed and analyzed court cases, with recommendations geared to lessen the likelihood that negligence suits will result if students sustain…

Assessment of the turbine trip transient in Cofrentes NPP with TRAC-BF1

International Nuclear Information System (INIS)

Castrillo, F.; Gomez, A.; Gallego, I.

1993-06-01

This report presents the results of the assessment of TRAC-BF1 (G1-J1) code with the model of C. N. Cofrentes for simulation of the transient originated by the manual trip of the main turbine. C. N. Cofrentes is a General Electric designed BWR/6 plant, with a nominal core thermal power of 2894 Mwt, in commercial operation since 1985, owned and operated by Hidroelectrica Espanola, S. A. The plant incorporates all the characteristics of BWR/6 reactors, with two turbine driven FW pumps. As a result of this assessment a model of C. N. Cofrentes has been developed for TRAC-BF1 that fairly reproduces operational transient behavior of the plant. A special purpose code was generated to obtain reactivity coefficients, as required by TRAC-BF1, from the 3D simulator

Elastic tripping analysis of corroded stiffeners in stiffened plate with irregular surfaces

Energy Technology Data Exchange (ETDEWEB)

Rahbarranji, Ahmad [AmirKabir University of Technology, Tehran (Iran, Islamic Republic of)

2014-09-15

Tripping of stiffeners is one of the buckling modes of stiffened panels which could rapidly lead to its catastrophic failure. Loss of thickness in the web and flange of stiffeners due to corrosion reduces elastic buckling strength. It is common practice to assume a uniform thickness reduction for corroded surfaces. To estimate the remaining strength of a corroded structure, a much higher level of accuracy is required since corroded surfaces are irregular. Finite element method is employed to analyze elastic tripping stress of corroded stiffeners with irregular surfaces. Comparing the results with elastic tripping stress of un-corroded stiffener, a reduction factor is introduced. It is found that for flat-bars and angle-bars the reduction factor increases by increasing corrosion loss; however, for tee-bars remains almost unchanged. Surface roughness has no significant effect on reduction of tripping Euler stress of angle-bars and flat-bars; however, it has an effect on reduction of tripping Euler stress of small flat-bars. For high values of corrosion loss, reduction of tripping Euler stress is higher in flat-bars than angle-bars. Corrosion at the mid-length or ends of flat-bars is more detrimental than full length. Corrosion at the ends of angle-bars is more detrimental than full length and mid-length.

TRIP-Br2 promotes oncogenesis in nude mice and is frequently overexpressed in multiple human tumors

Directory of Open Access Journals (Sweden)

Peh Bee

2009-01-01

Full Text Available Abstract Background Members of the TRIP-Br/SERTAD family of mammalian transcriptional coregulators have recently been implicated in E2F-mediated cell cycle progression and tumorigenesis. We, herein, focus on the detailed functional characterization of the least understood member of the TRIP-Br/SERTAD protein family, TRIP-Br2 (SERTAD2. Methods Oncogenic potential of TRIP-Br2 was demonstrated by (1 inoculation of NIH3T3 fibroblasts, which were engineered to stably overexpress ectopic TRIP-Br2, into athymic nude mice for tumor induction and (2 comprehensive immunohistochemical high-throughput screening of TRIP-Br2 protein expression in multiple human tumor cell lines and human tumor tissue microarrays (TMAs. Clinicopathologic analysis was conducted to assess the potential of TRIP-Br2 as a novel prognostic marker of human cancer. RNA interference of TRIP-Br2 expression in HCT-116 colorectal carcinoma cells was performed to determine the potential of TRIP-Br2 as a novel chemotherapeutic drug target. Results Overexpression of TRIP-Br2 is sufficient to transform murine fibroblasts and promotes tumorigenesis in nude mice. The transformed phenotype is characterized by deregulation of the E2F/DP-transcriptional pathway through upregulation of the key E2F-responsive genes CYCLIN E, CYCLIN A2, CDC6 and DHFR. TRIP-Br2 is frequently overexpressed in both cancer cell lines and multiple human tumors. Clinicopathologic correlation indicates that overexpression of TRIP-Br2 in hepatocellular carcinoma is associated with a worse clinical outcome by Kaplan-Meier survival analysis. Small interfering RNA-mediated (siRNA knockdown of TRIP-Br2 was sufficient to inhibit cell-autonomous growth of HCT-116 cells in vitro. Conclusion This study identifies TRIP-Br2 as a bona-fide protooncogene and supports the potential for TRIP-Br2 as a novel prognostic marker and a chemotherapeutic drug target in human cancer.

Small break LOCA analysis for RCP trip strategy for YGN 3 and 4 emergency procedure guidelines

International Nuclear Information System (INIS)

Suh, Jong Tae; Bae, Kyoo Hwan

1995-01-01

A continued operation of RCPs during a certain small break LOCA may increase unnecessary inventory loss from the RCS causing a severe core uncovery which might lead to a fuel failure. After TMI-2 accident, the CEOG developed RCP trip strategy called 'Trip-Two/Leave-Two' (T2/L2) in response to NRC requests and incorporated it in the generic EPG for CE plants. The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis determines the RCP trip setpoint and demonstrates the safe operational aspects of RCP trip strategy during a small break LOCA for YGN 3 and 4. The trip setpoint of the first two RCPs for YGN 3 and 4 is calculated to be 1775 psia in pressurizer pressure based on the limiting small break LOCA with 0.15 ft 2 break size in the hot leg. The analysis results show that YGN 3 and 4 can maintain the core coolability even if the operator fails to trip the second two RCPs or trips at worst time. Also, the YGN 3 and 4 RCP trip strategy demonstrates that both the 10 CFR 50.46 requirements on PCT and the ANSI standards 58.8 requirements on operator action time can be satisfied with enough margin. Therefore, it is concluded that the T2/L2 RCP trip strategy with a trip setpoint of 1775 psia for YGN 3 and 4 can provide improved operator guidance for the RCP operation during accidents. 11 figs., 4 tabs., 9 refs. (Author)

Influence of Field Trip on the Development of Students Interest ...

African Journals Online (AJOL)

Result of the study showed that; field trip increased students' interest towards studying fine and applied art theory and practicals. Male interest towards studying fine and applied art after embarking on field trip is slightly higher than their female counterpart but the difference is not significant at 0.05 alpha level under 56 ...

Nuevos atrayentes de trips ayudan a los agricultores en el control de plagas

NARCIS (Netherlands)

Tol, van R.W.H.M.; Kogel, de W.J.; Teulon, D.

2007-01-01

Los trips constituyen una plaga importante que afecta a muchos cultivos diferentes. El año pasado se probaron con éxito, en situaciones prácticas, aromas atrayentes de trips de las flores y trips de la cebolla. El producto, que estará a disposición de los cultivadores en junio, resultó efectivo en

The SMS-GPS-Trip-Method

DEFF Research Database (Denmark)

Reinau, Kristian Hegner; Harder, Henrik; Weber, Michael

2015-01-01

This article presents a new method for collecting travel behavior data, based on a combination of GPS tracking and SMS technology, coined the SMS–GPS-Trip method. The state-of-the-art method for collecting data for activity based traffic models is a combination of travel diaries and GPS tracking...

Assessment of vehicle trip production rates in Ilorin (Nigeria) | Jimoh ...

African Journals Online (AJOL)

Occupation, age, gender, income lev-el, vehicle ownership, trip length and fare structure affected the total trip generation, with an average production rate of 3.5, in the range of 2.79 - 4.29. The lower rate was characteristic of school children (5 - 15 years), while the highest rate was attributed to affluent and elderly persons ...

PUMP: analog-hybrid reactor coolant hydraulic transient model

International Nuclear Information System (INIS)

Grandia, M.R.

1976-03-01

The PUMP hybrid computer code simulates flow and pressure distribution; it is used to determine real time response to starting and tripping all combinations of PWR reactor coolant pumps in a closed, pressurized, four-pump, two-loop primary system. The simulation includes the description of flow, pressure, speed, and torque relationships derived through pump affinity laws and from vendor-supplied pump zone maps to describe pump dynamic characteristics. The program affords great flexibility in the type of transients that can be simulated

Large Pelagic Logbook Trip Survey (Vessels)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains catch and effort for fishing trips that are taken by vessels with a Federal permit issued for the swordfish and sharks under the Highly...

An Evaluation of Telecommuting As a Trip Reduction Measure

OpenAIRE

Kitamura, Ryuichi; Mokhtarian, Patricia L.; Pendyala, Ram M.

1991-01-01

Telecommuting, which is the performance of work at home or at a center close to home using telecommunications, has attracted growing interest among planners and researchers as a strategy for reducing traveldemand. This paper investigates the potential of telecommuting as a trip reduction measure, using data obtained from a telecommuting pilot project involving State of California government employees. In this pilot project, a three-day trip diary was administered, before and after te...

The evaluation of research reactor TRIGA MARK II safety

International Nuclear Information System (INIS)

Jordan, R.; Kozuh, M.; Mavko, B.

1994-01-01

In the paper the Probabilistic Safety Analysis (PSA) of a research reactor is described. Five different initiating events were selected and analyzed with the use of event trees. Seven reactor systems were modeled with fault trees. Three groups of radiation releases were introduced - Success, Reactor-Hall, Environment - and their frequencies were estimated. The importance factors of initiating events, human errors and basic events were calculated regarding the consequence groups. (author)

Transient thermal-hydraulic simulations of direct cycle gas cooled reactors

International Nuclear Information System (INIS)

Tauveron, Nicolas; Saez, Manuel; Marchand, Muriel; Chataing, Thierry; Geffraye, Genevieve; Bassi, Christophe

2005-01-01

This work concerns the design and safety analysis of gas cooled reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbo-machine trip, 10 in. cold duct break, 10 in. break in cold duct combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation. The turbo-machine contribution is discussed and can offer a help or an alternative to 'active' heat extraction systems

A Bayesian additive model for understanding public transport usage in special events.

Science.gov (United States)

Rodrigues, Filipe; Borysov, Stanislav; Ribeiro, Bernardete; Pereira, Francisco

2016-12-02

Public special events, like sports games, concerts and festivals are well known to create disruptions in transportation systems, often catching the operators by surprise. Although these are usually planned well in advance, their impact is difficult to predict, even when organisers and transportation operators coordinate. The problem highly increases when several events happen concurrently. To solve these problems, costly processes, heavily reliant on manual search and personal experience, are usual practice in large cities like Singapore, London or Tokyo. This paper presents a Bayesian additive model with Gaussian process components that combines smart card records from public transport with context information about events that is continuously mined from the Web. We develop an efficient approximate inference algorithm using expectation propagation, which allows us to predict the total number of public transportation trips to the special event areas, thereby contributing to a more adaptive transportation system. Furthermore, for multiple concurrent event scenarios, the proposed algorithm is able to disaggregate gross trip counts into their most likely components related to specific events and routine behavior. Using real data from Singapore, we show that the presented model outperforms the best baseline model by up to 26% in R2 and also has explanatory power for its individual components.

Safety analysis of a high temperature supercritical pressure light water cooled and moderated reactor

International Nuclear Information System (INIS)

Ishiwatari, Y.; Oka, Y.; Koshizuka, S.

2002-01-01

A safety analysis code for a high temperature supercritical pressure light water cooled reactor (SCLWR-H) with water rods cooled by descending flow, SPRAT-DOWN, is developed. The hottest channel, a water rod, down comer, upper and lower plenums, feed pumps, etc. are modeled as junction of nodes. Partial of the feed water flows downward from the upper dome of the reactor pressure vessel to the water rods. The accidents analyzed here are total loss of feed water flow, feed water pump seizure, and control rods ejection. All the accidents satisfy the criteria. The accident event at which the maximum cladding temperature is the highest is total loss of feedwater flow. The transients analyzed here are loss of feed water heating, inadvertent start-up of an auxiliary water supply system, partial loss of feed water flow, loss of offsite power, loss of load, and abnormal withdrawal of control rods. All the transients satisfied the criteria. The transient event for which the maximum cladding temperature is the highest is control rod withdrawal at normal operation. The behavior of loss of load transient is different from that of BWR. The power does not increase because loss of flow occurs and the density change is small. The sensitivities of the system behavior to various parameters during transients and accidents are analyzed. The parameters having strong influence are the capacity of the auxiliary water supply system, the coast down time of the main feed water pumps, and the time delay of the main feed water pumps trip. The control rod reactivity also has strong influence. (authors)

How travellers’ schedule their trips under uncertain travel times

DEFF Research Database (Denmark)

Hjorth, Katrine

Travel times play an important role when people decide where, when and how much to travel. But travel times are not always predictable from the traveller’s point of view: They may vary from day to day due to demand fluctuations, weather conditions, accidents and other unforeseen events that cause...... road capacity to decrease. We refer to this uncertainty as travel time variability (TTV). TTV is likely to affect how travellers schedule their trips, since it affects their probability of arriving late at their destination. We would like to account for TTV in traffic models and cost-benefit analyses......, but in practice there are limits to the kinds of behaviour that can be accommodated in such applications. For that reason, we are not solely interested in explaining travellers’ behaviour, but also in whether this behaviour can be approximated by behavioural models that are simple enough to be applied in traffic...

Pig herd monitoring and undesirable tripping and stepping prevention

DEFF Research Database (Denmark)

Gronskyte, Ruta; Clemmensen, Line Katrine Harder; Hviid, Marchen Sonja

2015-01-01

Humane handling and slaughter of livestock are of major concern in modern societies. Monitoring animal wellbeing in slaughterhouses is critical in preventing unnecessary stress and physical damage to livestock, which can also affect the meat quality. The goal of this study is to monitor pig herds...... at the slaughterhouse and identify undesirable events such as pigs tripping or stepping on each other. In this paper, we monitor pig behavior in color videos recorded during unloading from transportation trucks. We monitor the movement of a pig herd where the pigs enter and leave a surveyed area. The method is based...... on optical flow, which is not well explored for monitoring all types of animals, but is the method of choice for human crowd monitoring. We recommend using modified angular histograms to summarize the optical flow vectors. We show that the classification rate based on support vector machines is 93% of all...

Student Self-Reported Learning Outcomes of Field Trips: The pedagogical impact

Science.gov (United States)

Lavie Alon, Nirit; Tal, Tali

2015-05-01

In this study, we used the classification and regression trees (CART) method to draw relationships between student self-reported learning outcomes in 26 field trips to natural environments and various characteristics of the field trip that include variables associated with preparation and pedagogy. We wished to examine the extent to which the preparation for the field trip, its connection to the school curriculum, and the pedagogies used, affect students' self-reported outcomes in three domains: cognitive, affective, and behavioral; and the extent the students' socioeconomic group and the guide's affiliation affect students' reported learning outcomes. Given that most of the field trips were guide-centered, the most important variable that affected the three domains of outcomes was the guide's storytelling. Other variables that showed relationships with self-reported outcomes were physical activity and making connections to everyday life-all of which we defined as pedagogical variables. We found no significant differences in student self-reported outcomes with respect to their socioeconomic group and the guide's organizational affiliation.

Mode, load, and specific climate impact from passenger trips.

Science.gov (United States)

Borken-Kleefeld, Jens; Fuglestvedt, Jan; Berntsen, Terje

2013-07-16

The climate impact from a long-distance trip can easily vary by a factor of 10 per passenger depending on mode choice, vehicle efficiency, and occupancy. In this paper we compare the specific climate impact of long-distance car travel with coach, train, or air trips. We account for both, CO2 emissions and short-lived climate forcers. This particularly affects the ranking of aircraft's climate impact relative to other modes. We calculate the specific impact for the Global Warming Potential and the Global Temperature Change Potential, considering time horizons between 20 and 100 years, and compare with results accounting only for CO2 emissions. The car's fuel efficiency and occupancy are central whether the impact from a trip is as high as from air travel or as low as from train travel. These results can be used for carbon-offsetting schemes, mode choice and transportation planning for climate mitigation.

Are short daily trips compensated by higher leisure mobility?

DEFF Research Database (Denmark)

Næss, Petter

2006-01-01

Studies in several cities have shown that inner-city residents travel shorter distances and use cars less for local transport than suburbanites do. However, according to some authors, a low daily amount of travel is likely to be compensated through more extensive leisure mobility at weekends...... and on holidays. On the basis of a study of residential location and travel in the Copenhagen metropolitan area, this paper addresses the phenomenon of compensatory travel. For travel within ‘weekend trip distance’ from the residence, inner-city living appears to have a certain compensatory effect in the form...... of a higher frequency of medium-distance leisure trips. Probably, this reflects a shortage of nature in the immediate surroundings of the dwelling as well as less leisure time tied to gardening and house maintenance. These compensatory trips imply a slight reduction of the transport-reducing effect of inner...

Determination of optimum pressurizer level for kori unit 1

Energy Technology Data Exchange (ETDEWEB)

Song, Dong Soo; Lee, Chang Sup; Yong, Lee Jae; Kim, Yo Han; Lee, Dong Hyuk [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

1998-12-31

To determine the optimum pressurizer water level during normal operation for Kori unit 1, performance and safety analysis are performed. The methodology is developed by evaluating {sup d}ecrease in secondary heat removal{sup e}vents such as Loss of Normal Feedwater accident. To demonstrate optimum pressurizer level setpoint, RETRAN-03 code is used for performance analysis. Analysis results of RETRAN following reactor trip are compared with the actual plant data to justify RETRAN code modelling. The results of performance and safety analyses show that the newly established level setpoints not only improve the performance of pressurizer during transient including reactor trip but also meet the design bases of the pressurizer volume and pressure. 6 refs., 5 figs. (Author)

Determination of optimum pressurizer level for kori unit 1

Energy Technology Data Exchange (ETDEWEB)

Song, Dong Soo; Lee, Chang Sup; Lee Jae Yong; Kim, Yo Han; Lee, Dong Hyuk [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

1997-12-31

To determine the optimum pressurizer water level during normal operation for Kori unit 1, performance and safety analysis are performed. The methodology is developed by evaluating {sup d}ecrease in secondary heat removal{sup e}vents such as Loss of Normal Feedwater accident. To demonstrate optimum pressurizer level setpoint, RETRAN-03 code is used for performance analysis. Analysis results of RETRAN following reactor trip are compared with the actual plant data to justify RETRAN code modelling. The results of performance and safety analyses show that the newly established level setpoints not only improve the performance of pressurizer during transient including reactor trip but also meet the design bases of the pressurizer volume and pressure. 6 refs., 5 figs. (Author)

Impacts of energy consumption and emissions on the trip cost without late arrival at the equilibrium state

Science.gov (United States)

Tang, Tie-Qiao; Wang, Tao; Chen, Liang; Shang, Hua-Yan

2017-08-01

In this paper, we apply a car-following model, fuel consumption model, emission model and electricity consumption model to explore the influences of energy consumption and emissions on each commuter's trip costs without late arrival at the equilibrium state. The numerical results show that the energy consumption and emissions have significant impacts on each commuter's trip cost without late arrival at the equilibrium state. The fuel cost and emission cost prominently enhance each commuter's trip cost and the trip cost increases with the number of vehicles, which shows that considering the fuel cost and emission cost in the trip cost will destroy the equilibrium state. However, the electricity cost slightly enhances each commuter's trip cost, but the trip cost is still approximately a constant, which indicates that considering the electricity cost in the trip cost does not destroy the equilibrium state.

THE NETWORK OF CITY PUBLIC TRANSPORT AS THE BASE FOR TRIP LENGTH DISTRIBUTION DETERMINING

Directory of Open Access Journals (Sweden)

P. Horbachov

2015-07-01

Full Text Available The up-to-date methods of modelling the demand for public transport services require an objective estimation and improvement. Such an improvement can be achieved by taking into account the trip length distribution during trip matrix calculation that requires determining the reasons of regularities occurance in city population trip lengths.

Enhanced Westinghouse WWER-1000 fuel design for Ukraine reactors

International Nuclear Information System (INIS)

Dye, M.; Shah, H.

2015-01-01

Westinghouse has completed design, development, and region quantity delivery of an enhanced Westinghouse fuel assembly for WWER-1000 reactors to support continued safe reactor operations. The enhanced design builds on the successful performance of an earlier generation design which has operated in the South Ukraine 3 reactor for multiple cycles without any fuel rod failures. Incorporated design enhancements include a thicker spacer grid outer strap, an enhanced spacer grid outer strap profile to limit the risk for, and impact of, mechanical interaction/interference with coresident fuel, an all Alloy 718 grid structure for improved stability and strength, and improvements to the top and bottom nozzles. Capable of meeting increased lateral loads generated from using a higher axial trip limit for the refueling machine crane, the design was verified by extensive mechanical and thermalhydraulic testing, which included a newly developed fuel assembly-to-fuel assembly handling test rig to assess performance during bounding core loading and unloading conditions. Through these extensive design enhancements and comprehensive testing program, the enhanced WWER-1000 design provides additional performance, handling, and reliability margins for safe reactor operation. (authors)

Trip Travel Time Forecasting Based on Selective Forgetting Extreme Learning Machine

Directory of Open Access Journals (Sweden)

Zhiming Gui

2014-01-01

Full Text Available Travel time estimation on road networks is a valuable traffic metric. In this paper, we propose a machine learning based method for trip travel time estimation in road networks. The method uses the historical trip information extracted from taxis trace data as the training data. An optimized online sequential extreme machine, selective forgetting extreme learning machine, is adopted to make the prediction. Its selective forgetting learning ability enables the prediction algorithm to adapt to trip conditions changes well. Experimental results using real-life taxis trace data show that the forecasting model provides an effective and practical way for the travel time forecasting.

Anticipated and abnormal transients in nuclear power plants

International Nuclear Information System (INIS)

Karam, R.A.

1987-01-01

This book contains the proceedings of an international conference on Anticipated and Abnormal Transients in Nuclear Power Plants. Included are the following papers: Comparative evaluation of recent water hammer events in light water reactors, Rick reduction through enhanced human performance, Assessment of the performance of an emergency boration system for anticipated transients without trip faults, Emergency procedure planning to mitigate event progression

Turn-key SRF accelerators to drive subcritical reactors

International Nuclear Information System (INIS)

Johnson, Rolland P.

2011-01-01

Large particle accelerator projects, both accomplished and proposed, have been used to engage US industry through contracts and grants to develop efficient capabilities to design, develop, produce, and deliver entire accelerator systems or any needed subsystems. Staffed in many cases by experienced scientists and engineers from National Laboratories and Universities, existing companies could extend their portfolios to offer turn-key accelerators with parameters to match the needs of ADS. If the reactors were based on molten salt fuel such that trip rate requirements were relaxed, the developments needed for a multi-MW proton accelerator for ADS would be minimal. Turn-key SRF proton linacs for ADS operation can be ordered now to enable GW-level power generation from natural thorium, natural uranium, or nuclear waste from conventional reactors. (author)

Safety evaluation report by the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission for U.S. Energy Research and Development Administration Light Water Breeder Reactor. Special project No. 561

International Nuclear Information System (INIS)

1976-07-01

The Safety Evaluation Report is presented for the Light Water Breeder Reactor (LWBR). The LWBR core is to be installed in the Shippingport reactor at the Shippingport Atomic Power Station. The Safety Evaluation Report is the result of an NRC staff review of the LWBR Safety Analysis Report submitted by the Division of Naval Reactors, U. S. Energy Research and Development Administration. As a result of its review, the NRC staff has recommended that: (1) a diverse trip signal, such as containment high pressure, be included in a 2-out-of-3 logic for initiation of safety injection; (2) power be locked out from the pressurizer surge isolation valve during normal operation; and (3) a chlorine monitor be installed in the main control room

Selection of important initiating events for Level 1 probabilistic safety assessment study at Puspati TRIGA Reactor

International Nuclear Information System (INIS)

Maskin, M.; Charlie, F.; Hassan, A.; Prak Tom, P.; Ramli, Z.; Mohamed, F.

2016-01-01

Highlights: • Identifying possible important initiating events (IEs) for Level 1 probabilistic safety assessment performed on research nuclear reactor. • Methods in screening and grouping IEs are addressed. • Focusing only on internal IEs due to random failures of components. - Abstract: This paper attempts to present the results in identifying possible important initiating events (IEs) as comprehensive as possible to be applied in the development of Level-1 probabilistic safety assessment (PSA) study. This involves the approaches in listing and the methods in screening and grouping IEs, by focusing only on the internal IEs due to random failures of components and human errors with full power operational conditions and reactor core as the radioactivity source. Five approaches were applied in listing the IEs and each step of the methodology was described and commented. The criteria in screening and grouping the IEs were also presented. The results provided the information on how the Malaysian PSA team applied the approaches in selecting the most probable IEs as complete as possible in order to ensure the set of IEs was identified systematically and as representative as possible, hence providing confidence to the completeness of the PSA study. This study is perhaps one of the first to address classic comprehensive steps in identifying important IEs to be used in a Level-1 PSA study.

AN INTEGRATED MODELING FRAMEWORK FOR ENVIRONMENTALLY EFFICIENT CAR OWNERSHIP AND TRIP BALANCE

Directory of Open Access Journals (Sweden)

Tao FENG

2008-01-01

Full Text Available Urban transport emissions generated by automobile trips are greatly responsible for atmospheric pollution in both developed and developing countries. To match the long-term target of sustainable development, it seems to be important to specify the feasible level of car ownership and travel demand from environmental considerations. This research intends to propose an integrated modeling framework for optimal construction of a comprehensive transportation system by taking into consideration environmental constraints. The modeling system is actually a combination of multiple essential models and illustrated by using a bi-level programming approach. In the upper level, the maximization of both total car ownership and total number of trips by private and public travel modes is set as the objective function and as the constraints, the total emission levels at all the zones are set to not exceed the relating environmental capacities. Maximizing the total trips by private and public travel modes allows policy makers to take into account trip balance to meet both the mobility levels required by travelers and the environmentally friendly transportation system goals. The lower level problem is a combined trip distribution and assignment model incorporating traveler's route choice behavior. A logit-type aggregate modal split model is established to connect the two level problems. In terms of the solution method for the integrated model, a genetic algorithm is applied. A case study is conducted using road network data and person-trip (PT data collected in Dalian city, China. The analysis results showed that the amount of environmentally efficient car ownership and number of trips by different travel modes could be obtained simultaneously when considering the zonal control of environmental capacity within the framework of the proposed integrated model. The observed car ownership in zones could be increased or decreased towards the macroscopic optimization

Medical and pharmacy student concerns about participating on international service-learning trips.

Science.gov (United States)

Chuang, Chih; Khatri, Siddique H; Gill, Manpal S; Trehan, Naveen; Masineni, Silpa; Chikkam, Vineela; Farah, Guillaume G; Khan, Amber; Levine, Diane L

2015-12-23

International Service Learning Trips (ISLT) provide health professional students the opportunity to provide healthcare, under the direction of trained faculty, to underserved populations in developing countries. Despite recent increases in international service learning trips, there is scant literature addressing concerns students have prior to attending such trips. This study focuses on identifying concerns before and after attending an ISLT and their impact on students. A survey comprised of closed and open-ended questions was developed to elucidate student concerns prior to attending an ISLT and experiences which might influence concerns. A five-point Likert-scale (extremely concerned = 1, minimally concerned = 5) was used to rate apprehension and satisfaction. Paired t-test was used to compare pre- and post-trip concerns; Chi-Square test was used to compare groups. Thirty-five students (27 medical, 8 pharmacy) attended ISLTs in December 2013. All completed pre and post-trip surveys. Significant decreases were seen in concerns related to cultural barriers (4.14 vs 4.46, P = .047), disease/epidemics (3.34 vs 4.60, P travel (3.86 vs 4.51, P food (3.83 vs 4.60, P students described benefits of attending an ISLT. Students had multiple concerns prior to attending an ISLT. Most decreased upon return. Addressing concerns has the potential to decrease student apprehension. The results of this study highlight the benefits of providing ISLTs and supporting development of a curriculum incorporating trip-related concerns.

RETRAN-3D analysis of the base case and the four extreme cases of the OECD/NRC Peach Bottom 2 Turbine Trip benchmark

International Nuclear Information System (INIS)

Barten, Werner; Coddington, Paul; Ferroukhi, Hakim

2006-01-01

This paper presents the results of RETRAN-3D calculations of the base case and the four extreme cases of phase 3 of the Peach Bottom 2 OECD/NRC Turbine Trip benchmark for coupled thermal-hydraulic and neutronic codes. The PSI-RETRAN-3D model gives good agreement with the measured data of the base case. In addition to the base case, the analysis of the extreme cases provides a further understanding of the reactor behaviour, which is the result of the dynamic coupling of the whole system, i.e., the interaction between the steam line and vessel flows, the pressure, the Doppler, void and control reactivity and power. For the extreme cases without scram the bank of safety relief valves is able to mitigate the effects of the turbine trip for short times. The 3-D nature of the core power distribution has been investigated by analysing the power density of the different thermal-hydraulic channels. In all cases prior to the reactor scram the course of the power is similar in all the channels with differences of the order of a few percent showing that, by and large, the core acts in a coherent manner. At the time of maximum power, the axial power distribution in the different channels is increased at the core centre with respect to the distribution at time zero, by an amount, which is different for the different channels

Effects of low upper shelf fracture toughness on reactor vessel integrity during pressurized thermal shock events

International Nuclear Information System (INIS)

Bamford, W.H.; Heinecke, C.C.; Balkey, K.R.

1988-01-01

For the past decade, significant attention has been focused on the subject of nuclear rector vessel integrity during pressurized thermal shock (PTS) events. The issue of low upper shelf fracture toughness at operating temperatures has been a consideration for some reactor vessel materials since the early 1970's. Deterministic and probabilistic fracture mechanics sensitivity studies have been completed to evaluate the interaction between the PTS and lower upper shelf toughness issues that result from neutron embrittlement of the critical beltline region materials. This paper presents the results of these studies to show the interdependency of these fracture considerations in certain instances and to identify parameters that need to be carefully treated in reactor vessel integrity evaluations for these subjects. This issue is of great importance to those vessels which have low upper shelf toughness, both for demonstrating safety during the original design life and in life extension assessments

A Quasi-Practical Interstellar Rocket Trip

Science.gov (United States)

Edmonds, James D., Jr.

1974-01-01

Mathematically shows that in principle a spaceship could travel eight light years in ten earth years, with the passengers arriving 4.6 years older than when they left earth and having experienced an acceleration induced effective gravity of one g for the entire trip. (MLH)

Singular system analysis of the Local Power Range Monitor (LPRM) readings of a Boiling Water Reactor (BWR) in an unstable event

International Nuclear Information System (INIS)

Ginestar Peiro, D.; Verdu, G.; Miro, R.

2006-01-01

Singular system analysis is a successful technique to separate oscillating components from a given signal. A methodology is proposed to apply this technique to the signals obtained from the LPRMs of a boiling water reactor core and extract the contributions of the in-phase oscillation and the out-of-phase oscillations from the LPRM readings during an unstable event. This methodology has been validated with synthetic signals and simulations of in-phase and out-of-phase oscillations of the Leibstadt reactor. Finally, one case of Ringhals I Stability Benchmark has been analysed. (author)